dennis j. kubicki, u.s. dept. of energy, md tel€¦ · (log #13) 805- 2 - (1-2): accept in...

Report of the Committee on

Fire Protection for Nuclear Facilities

Wayne D. Holmes, Chair HSB Professional Loss Control, CT [I]

Richard B. Alpert, Triad Fire Protection Engr Corp., PA [SE] Marlo A. Antonetti, Gage-Babcock & Assoc., Inc., NY [SE] James B. Biggins, Marsh USA, Inc., IL [I] William G. Boyce, U.S. Dept. of Energy, DC [El David J. Buell, Black & Veatch Corp., KS [SE] Thomas C. Carlisle, Virginia Power, VA [U] Harry M. Corson, IV, Siemens Cerberus Division, NJ [M]

Rep. Nat'l Electrical Mfrs. Assn. Stanford E. Davis, PP & L, PA [U] Edgar G. Dressier, American Nuclear Insurers, PA [I] Franklin D. Garrett, Arizona Public Service Co., AZ [U]

Rep. Nuclear Energy Inst. Al'ie T. P. Go, Bechtel Corp., CA [SE] Dennis Wayne Henneke, Southern California Edison, CA [U] Robert Kalantari, EPM Inc., MA [SE] Robert P. Kassawara, Electric Power Research Inst., CA [U] Donald J. Kohu, Kohn Engr, PA [SE] Christopher A. Ksoblech, Wisconsin Electric Power Co., WI [U] Patrick M. Madden, U.S. Nuclear Regulatory Commission, DC [E] David P. Notley, Science Applications Int'l Corp., MD [SE] Clifford R. Sinopoli II, Peco Energy, PA [U]

Rep. Edison Electric Inst. Wayne R. Sohlman, Nuclear Electric Insurance Lmt., DE [I] WtUiam M. Sullivan, Contingency Mgrnt. Assoc., Inc., MA [SE] Raymond N. Tell, Los Alamos Nat'l Laboratory, NM [U] Steven F. Vieira, Tyco Int'l, Ltd., RI [M]

Rep. Nat'l Fire Sprinkler Assn. Michael J. Vitacco, Jr., Westinghouse Savannah River Co., SC [U]

Alternates

Amir Afzali, Pacific Gas and Electric Co., CA [U] (Alt. to D. W. Henneke)

Edward A. Conneli, U.S. Nuclear Regulatory Commission, DC [E] (Alt. to P. M. Madden)

Frank Czysz, Pennsylvania Power & Light, Inc., PA [U] (Alt. to S. E. Davis)

Thomas K. Furlong, Nuclear Service Organization, DE [I] (Ah. to W. B. Sohlman)

L. Paul Herman, HSB Professional Loss Control, IL [I] (Ah. to W. D. Holmes)

Neal W. Krantz, Nat'l Time & Signal Corp., MI [M] (Ah. to H. M. Corson)

Dennis J. Kubicki, U.S. Dept. of Energy, MD tEl (Ah. to W. G. Boyce)

Andrew R. Ratchf0rd, Retchford Diversified Services (RDS), CA [SE] (Voting Alt.)

Ronald Rispoli, Entergy Corp., AR [U] (Ah. to F. D. Garrett)

Thomas A. Storey, Science Applications Int'l Corp., VA [SE] (Ah. to D. P. Notley)

James R. Streit, Los Alamos Nat'l Laboratory, NM [U] (Ah. to R. N. Tell)

William B. Till, Jr., Westinghouse Savannah River Co., SC [U] (Alt. to M. J. Vitajcco)

Nonvoting

Leonard R. Hathaway, Alpharetta, GA [I] (Member Emeritus)

Donald J. Keigher, Los Alamos, NM (Member Emeritus)

James E. Lechner, Nebraska Public Power District, NE [U] Walter W. Maybee, Bothell, WA

(Member Emeritus) Nathan O. Siu, U.S. Nuclear Regulatory Commission, DC [E]

Staff Liaison: Richard P. Bielen

Committee Scope: This Committee shall have primary responsibility for documents on the safeguarding of life and 0roperty from fires in which radiation or other effects of nuclear energy might be a factor.

This list represent.s the member~'hip at the time the Committee was balloted on the text qf this edition. Since that time, changes in the membership may have occurred. A key to classifications is found at the front of this book.

This portion of the Technical Committee Report of the Committee on Fire Protection for Nuclear Facilities is presented for adoption.

This Report on Comments was prepared by the Technical Committee on Fire Protection for Nuclear Facilities and documents its action on the comments received on its Report on Proposals on NFPA 805, Peformance~Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, as published in the Report on Proposals for the 2000 November Meeting.

This Report on Comments has been submitted to letter ballot of the Technical Committee on Fire Protection for Nuclear Facilities which consists of 26 voting members. The results of the balloting, after circulation of any negative votes, can be found in the report.

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N F P A 8 0 5 m N o v e m b e r 2 0 0 0 R O C

(Log #78) 805- 1 - (Title): Reject SUBMITTER: T.A. O ' C o n n o r , A m e r G e n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Change title to read:

"Performance Based Standard for Nuclear Reactor Electric Genera t ing Plants" SUBSTANTIATION: T he cur ren t title excludes heavy water plants. A title change or in t roductory note regarding applicability to o the r types of reactor plants used for genera t ion of electricity would result in wider use o f the s tandard such as in o the r countr ies such as Canada. All nuc lear power electrical genera t ing facilities have fire tl;~reats that could be addressed us ing this s tandard but with the cu r ren t title, users and regulatory agencies o f these o the r types o f facilities will avoid use of the s tandard as "not applicable" based on the title alone. C O M M r r T E E ACTION: Reject. COMMITTEE STATEMENT: The d o c u m e n t was written a round light water nuclear reactors only. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:

AFFIRMATIVE: 24 N O T RETURNED: 2 Biggins, Vieira

(Log #13) 805- 2 - (1-2): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise last sen tence to read as follows:

Defense-in-depth shall be achieved when an adequate level o f each o f the following e lements is provided: SUBSTANTIATION: Clarify intent . COMMITTEE ACTION: Accept in Principle.

I Revise text to read as follows: "Defense-in-depth shall be achieved when an adequate tevet

ba lance o f each o f the following e lements is provided:" COMMITTEE STATEMENT: The change clarifies that a balance is desired. NUMBER OF C O M M I T r E E MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #94) 805- 3 - (1-2): Accept SUBM1TrER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

Protect ing the safety o f the public, the env i ronment , and plant personnel f rom a plant fire and its potential effect on safe reactor operat ions is be- p a r a m o u n t to this s tandard. S U B S T A N T I A T I O N : Editorial. COMMITTEE ACTION: Accept. NUMBER OF C O M M I T r E E MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITT E E ACTION:


NOTE: The ballot on this C o m m e n t did not receive the required two-thirds aff irmaive vote o f the Committee , but the C o m m e n t is still rejected.

(Log #74) 805- 4 - (1-3.4, 1-4.4, 1-5.4, 4-4): SUBMITTER: Ira Heatherly, Tennessee Valley Authori ty COMMENT O N PROPOSAL NO: 805d04

D I I N / ] D RECOMMENDATION : D e l e t e . . a a t ~ a m a g c , ~t:~iP.c~ !a:::=',:~::.~,~ (This entire topic should be deleted - Reference Sections 1-3.4, 1-4.4, 1-5.4 and 4-4.) S U B S T A N T I A T I O N : Issuing code based gu idance for economic loss is no t appropriate . Additionally, this topic only discusses the obvious (e.e., in this area criteria will be establ ished by the owner /ope ra to r ) . Inclusion o f this topic does not e n h a n c e the technical aspect o f this code. COMMITrEE ACTION: Reject. COMMITTEE STATEMENT: The Commi t t ee has debated this several t imes and the existing wording reflects the in tent o f the commi t tee that Plant D a m a g e / B u s i n e s s In te r rup t ion shou ld remain in the s tandard.

NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 15 NEGATIVE: 9 NOT RETURNED: 2 Biggins, Vieira

EXPLANATION OF'NEGATIVE: ALPERT: See c o m m e n t s by Sinopoli and Garrett. CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated

8 /9 /2000 . DAVIS: I concur with Mr. Heather ly 's commen t . GARRETT: Issuing code based requ i rements for prevent ing

economic loss is no t appropria te for this s tandard. Additionally, this topic only discusses the obvious (i.e., in this area the criteria will be establ ished by the owner /ope ra to r ) . Inclusion o f this topic does not e n h a n c e the technical aspect of this code.

HENNEKE: Economic loss should not be in this s tandard. Previous Nuclear fire protect ion requ i rements did not include economic loss, so the s tandard is providing new and addit ional r equ i rements for opera t ing plants.

KASSAWARA: Guidance on mat ters conce rn ing economic loss is no t appropria te in a s tandard which is primarily in tended to define safety related requirements .

KSOBIECH: The commit tee should accept this com m en t . This topic should be deleted from the s tandard. Property Damage and Business In ter rupt ion is an economic issue that should be hand led separately by the owner /ope ra to r and NOT regulated th rough the NFPA process. I know of no o ther industry for which NFPA provides fire protect ion s tandards that also includes s tandard requ i rements for property damage and business in terrupt ion. This may be a great topic for the NFPA Handbook , but should not be regulated th rough the NFPA standards process. In addit ion, there is no guidance in the s tandard or technical direction on how to apply a pe r fo rmance based approach to this issue and provides no value to the s tandard. In fact, I believe leaving this topic in the s tandard in its cu r ren t form quest ions the credibility of the r emain ing s tandard and its use and application.

RATCHFORD: Issuing code based requ i rements for prevent ing economic loss is no t appropria te for this s tandard. Additionally, this topic only discusses the obvious (i.e., in this area the criteria will be establ ished by the owner /ope ra to r ) . Inclusion of this topic does not e n h a n c e the technical aspect o f this code.

SINOPOLI: The inclusion o f Plant D a m a g e / B u s i n e s s In te r rupt ion has been the source of m u c h debate within the commit tee . Economic losses associated with p lan t damage and business in ter rupt ion are the responsibility o f the o w n e r / o p e r a t o r and their insurance provider. The Plant Damage /Bus in e s s In ter rupt ion requi rements in the body of the s tandard are vague and do no th ing to improve the cu r ren t fire protect ion programs at existing plants. The informat ion provided by the s tandard is primarily append ix informat ion and thus is no t enforceable. Each insurance provider has developed loss prevent ion s tandards that direct the o w n e r / o p e r a t o r as to what is required to mainta in insurability. The insurance companies will no t e l iminate these Loss Prevention Standards in favor of NFPA 805. Finally, the opera t ing sites for which this s tandard is written, are mature opera t ing sites with good fire protect ion loss histories. The void that the Plant Damage /Bus ines s In te r rupt ion requ i rements are in t ended to fill, simply does not exist.

(Log #14) 805- 5 - (1-4.1(c)): Accept SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

Capable of prevent ing fuel clad damage so that the primary c o n t a i n m e n t boundary is no t chal lenged. S U B S T A N T I A T I O N : None. COMMITTEE ACTION: Accept. COMMITTEE STATEMENT: The change should keep the heading , Fission Product Boundary. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 22 NEGATIVE: 2 N O T RETURNED: 2 Biggins, Vieira

EXPLANATION OF NEGATIVE: CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated

8 /9 /2000 . SINOPOLI: The wording proposed by this c o m m e n t and

accepted by the commi t tee is incorrect. The re is no cause and

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N F P A 805 ~ N o v e m b e r 2000 R O C

effect relat ionship between fuel c ladding integrity and primary con t a inmen t integrity. The change incorpora ted by this c o m m e n t el iminated reference to the integrity o f the primary coolant houndaly . A het ter approach would have been to state, Capable o f p~eventing fuel clad damage "and" rupture of pr imary coolant bounda~x. In addit ion, this action uses the word "challenge." It is difficult io quantify and enforce the implications o f "challenge." COMMENT O N AFFIRMATIVE:

GARRETT: This c o m m e n t is aff irmed based on the Industry 's under s t and ing that it was in tended to preclude the need for analyzing all primary c o n t a i n m e n t c o m p o n e n t s as required safe shutdown equ ipment . This is reflected in the per fo rmance criteria to prevent fllel damage as well, and is consistent with the cu r ren t regulat ion.

IL4.TCHFORD: This c o m m e n t is aff i rmed based on the lndustry 's unde r s t and ing that it was in tended to preclude the need fi~r analyzing all primary con t a i nmen t c o m p o n e n t s as required safe shutdown e q u i p m e n t . . T h i s is reflected in the pe r fo rmance criteria to prevent tirol damage as well, and is consis tent with the cu r ren t regulat ion.

(Log #95) 805- 6 - (14 .1(c ) ) : Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Add the following to the end o f the existing statement:

...so that the primary c o n t a i n m e n t is not chal lenged. SUBSTANTIATION: Provides a more direct tie to the exist ing regulat ion. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t t ee Action on C o m m e n t 805-5 (Log #14). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #15) 805- 7 - (1-4.2): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory Commis s ion COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

In the event o f a fire du r ing any operat ional mode or plant configurat ion, the c o n t a i n m e n t of radioactive materials shall be main ta ined . SUBSTANTIATION: Make consis tent with 1-4.1. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t t ee Action on Comrn'ent 805-8 (Log #96). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #96) 805- 8 - (1-4.2): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

1-4.2 Radioactive Release Objective. Either of the T h e following objectives shall be me t du r ing all opera t ional modes and plant configurat ions:

(2) Societal risks to life and heal th associated with f i re- induced releases o f radioactivity shall no t be a significant addi t ion to o the r societal risks.

(a) C o n t a i n m e n t integrity is caoable o f being ma in ta ined (b) The source term is capable of be ing l imited

SUBSTANTIATION: More direct f rom objectives to criteria. COMMITTEE ACTION: Accept. COMMITTEE STATEMENT: Editorial. Part (2) should have been deleted. This was clarified by the submit ter . T he original submittal conta ined a format t ing error.

NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #16) 805- 9 - (1-5.1(b)): Reject SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text to read as follows: •

Shall be capable o f control l ing coolant level such that subcool ing and level indication in the pressurizer is main ta ined for a PWR, and reactor coolant level is ma in ta ined above the top o f active fuel for a BWR. SUBSTANTIATION: Curren t wording does not provide an adequate level of nuc lear safety. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: See Commi t tee Sta tement on 805-10 (Log~98). NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:

AFFIRMATIVE: 23 NEGATIVE: I NOT RETURNED: 2 Biggins, Vieira

EXPLANATION OF NEGATIVE: CONNELL: The nuclear safety performance criteria for inventory

and pressure control in Section 1-5.1 of the draft standard deviate from the criteria specified in Section III.L of Appendix R to I0 CFR Part 50. The NRC staff has not completedits evaluation of the proposed criteria to determine its adequacy for ensuring public health and safety.

(Log #98) 805- 10- (1-5.1(b)): Accept in Principle SUBMITrER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise the sen tence to read as follows:

(b) Inventory and Pressure Control. Shall be capable of cont ro l l ing coolant level such that subcool ing is main ta ined for a PWR and shall be capable o f main ta in ing or rapidly :eatoring reactor ~ coolant level above the top o f active fuel for a BWR. SUBSTANTIATION: This is consis tent with the existing regulation. A fire should not result in a severe plant t ransient but ra ther should be treated as an ant icipated operat ional occurrence. COMMITTEE ACTION: Accept in Principle.

I evise existing 1-5.1(b) by adding at the end: "such that fuel clad damage as a result o f a fire is prevented ~

COMMITTEE STATEMENT: The nuclear safety objectives in part, requires "prevent ing fuel-clad damage". The commit tee ' s response to 805-9 (Log #16) fur ther s t reng thens the nuclear safety pe r fo rmance criteria by add ing the expectat ion, "such that fuel- clad damage as a result o f a fire is prevented". It was the commi t t ee ' s posit ion that pe r fo rmance criteria in conjunct ion with the requ i rements and guidance of the s tandard will provide an adequate level of safety. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: CONNELL: See my Explanat ion o f Negative Vote on C o m m e n t

805-9 (Log #16).

(Log #CC2) 805- 11 - (1-5.1(b)): Accept SUBMITTER: Technical Commi t t ee on Fire Protect ion for Nuclear Facilities COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

[ Inventory and Pressure Control. With fuel in the reactor vessel. ] head on and tensioned, inventorv and oressure control shall be [ capable o f control l ing coolant level such that subcool ing is | ma in ta ined for a PWR and shall be capable o f main ta in ing or [ rapidly restoring reactor water level above top o f active fuel for a I BWR, such that fuel clad damage as a result of a fire is prevented.

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NFPA 8 0 5 1 N o v e m b e r 2 0 0 0 ROC

SUBSTANTIATION: The criteria as current ly written is no t applicable to states when the head is off or the fuel is in the spent fuel pool. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #3) 805- 12- (1-5.1(e)): Accept in Principle SUBMITTER: Ronald Rispoli, Entergy Corp. COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

"Shall be capable o f providing the necessary informat ion to assure nuc lear safety..." SUBSTANTIATION: Use of the wbrd indicat ion implies that a device readout is provided for each process variable. However, the use o f boron sampl ing for shu tdown mon i t o r i ng will provide appropria te informat ion if source range mon i to r ing is lost. The use o f the word versus indication will better reflect the opt ions available for process moni tor ing . COMMITTEE ACTION: Accept in Principle.

I dd append ix material as follows: A-l-5.1(e) Indicat ion may be obta ined by various means such as

sampl ing/analys is , provided the requi red in format ion can be obta ined witbin the t ime f rame needed . COMMITTEE STATEMENT: The append ix material gives fur ther informat ion on the type of indication and informat ion needed . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 24 N O T RETURNED: 2 Biggin,s, Vieira

(Log #92) 805- 13- (1-5.2): Accept SUBMITTER: Chr i s topher P r agman , PECO Energy COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

I 1-5.2 Radioactive Release Per formance Criteria. Radiat ion I release to any 'unres t r ic ted area due to the direct effects o f fire or

fire suppress ion activities (but not involving fuel damage) shall be I as low as reasonably achievable and shall no t exceed 0.002 rem I (0.02 millisevert) in any one hour .

SUBSTANTIATION: Release criteria needs to specify that the criteria applies to member s of the public. This is consis tent with 10 CFR 20.1301. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE .MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: CONNELL: The commi t tee action for this c o m m e n t is incorrect.

The commi t tee action is listed as "Accept" and it should be "Accept in Principle" as the submi t te r ' s r e c o m m e n d a t i o n was not adopted as proposed.

(Log #CC9) 805- 14- (1-5.2): Accept SUBMITTER: Technical Commi t t ee on Fire Protect ion for Nuclear Facilities COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: 1) Revise 1-5.2 to read as follows: "Radiation release due to the direct effects o f fire suppress ion activities (but not involving fuel damage) shall be as low as reasonably achievable and shall no t exceed applicable i0 CFR Part 20 Limits."

2) Add reference to 10 CFR, Part 20 in the reference section. SUBSTANTIATION: This change references federal law for requi rements . COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #18) 805- 15- (1-6): Accept SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805404

I RECOMMENDATION: Add text to read as follows: PWR - Pressurized Water Reactor BWR - Boiling Water Reactor

S U B S T A N T I A T I O N : Abbreviat ions BWR/PWR are used in Section 1-5 but are not defined. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #CC8) 805- 16- (1-6): Accept SUBMITTER: Technical Commi t tee on Fire Protect ion for Nuclear Facilities COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Add definit ion o f Approved.

1-6 Definitions. Approved. Accentable to the authori ty having Jurisdiction. Add appendix material from the s tandard NF-P-A definit ion o f .

approved. A-l-6 Approved. The National Fire Protect ion Association does

not approve, inspect, or certify any installations, procedures , equ ipmen t , or materials; no r does it approve or evaluate testing laboratories. In de t e rmin ing the acceptability o f installations, procedures , equ ipmen t , or materials, the authori ty having jur isdic t ion may base acceptance on compl iance with NFPA or o the r appropria te s tandards. In the absence o f such s tandards, said authori ty may require evidence o f proper installation, procedure , or use. The authori ty having jur isdic t ion may also refer to the listings or labeling practices of an organizat ion that is conce rned with product evaluations and is thus in a posit ion to de te rmine compl iance with appropria te s tandards for the cu r ren t

I p roduc t ion o f listed items. , SUBSTANTIATION: NFPA 805 does not have a defini t ion for the word "approved". The word "approved" is used in Section 3-9.5 relative to suppress ion system isolation valves but is never defined. (Section 3-9.5 is deal ing with Automat ic and Manual Water Based Suppress ion Systems.) This topic is best covered by NFPA 13.

NFPA 13's defini t ion section def ines the word "Approved" as follows:

"Approved. Acceptable to the authori ty having jurisdict ion." COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #17) 805- 17 - (1-6 Analysis Area) : Accept SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT ON PROPOSAL NO: 804-404

I RECOMMENDATION: Delete the term "Analysis Area". SUBSTANTIATION: Te rm is confus ing and conflicts with "Fire Area". COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #88) 805- 18- (1-6 Combust ible , Fire Resistance Rating, Fuel Dam ag e ) : Accept in Principle in Part SUBMITTER: Chr i s topher P r agman , PECO Energy. COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise the following definit ions:

Combust ible . Capable of u n d e r g o i n g combus t ion per applicable test proto¢91~.

Fire Resistance Rating. The time, in minutes or hours , that materials or assemblies have withstood a fire exposure as

261


establ ished in accordance with an approved test p rocedure appropria te for the strilcture, bui ld ing material, or c o m p o n e n t u n d e r considerat ion.

Fuel Damage. Exceeding the reactor fuel design limits. SUBSTANTIATION: Definit ions lack consistency with their use elsewhere in the s tandard. COMMITTEE ACTION: Accept in Principle in Part.

1. Use the definit ion o f "Combustible" f rom NFPA 804 which reads as follows:

1-6 Combust ible* Capable o f u n d e r g o i n g combust ion . A-I-6 Combustible. Any material that, in the form in which it is

used and u n d e r the condi t ions ant icipated will ignite and burn . A material that does not mee t the defini t ion of noncombus t ib le or l inf i ted-combust ible .

2. Accept part 2. 3. Reject part 3.

COMMITTEE STATEMENT: T he s tandard addresses all fuels, no t ju s t reactor fuel. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMMITTEE ACTION:


VOTE O N COMMITTEE ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #99) 805- 22 - (1-6 Essential Personnel) : Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services CO MMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

Essential Personnel . Personnel that are required to perform funct ions ic, the event to mitieate the effects of a fire, including but are not l imited to fire brigade members , operat ions, heal th physics, security, and ma in tenance . SUBSTANTIATION: More accurately reflects the intent ion of the s tandard - - it is not the intent ion of the s tandard to address personnel v)ho may be involved in staffing the TSC or EOF. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #91) 805- 19- (1-6 Comple teness Uncertainty): Accept SUBMITTER: Chr i s topher P ragman , PECO Energy COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION" Revise defini t ion for Comple teness Uncer ta inty as follows:

Uncer ta inty in the predict ions o f a mode l due to mode l scope limitations This uncer ta inty reflects an unanalyzed contr ibut ion .or reduct ion of risk d u e to l imitations of the available analytical methods . SUBSTANTIATION: Uncer ta inty can unders ta te or overstate risk. Cur ren t definit ion implies that uncer ta inty can only unders ta te risk. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #90) 805- 2 0 - (1-6 Con ta imnen t ) : Accept SUBMITTER: Chr i s topher P r agman , PECO Energy COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise defini t ion for "Conta inment" as follows:

Structures, systems, and -o r componen t s provided to prevent or mitigate the release of radioactive materials in the c;'~nt c, f a .qrc. SUBSTANTIATION: Cnr ren t defini t ion creates conflict with 10 CFR. Fire is not the design basis accident which establishes con t a inmen t design criteria. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #65) 805- 2 3 - (1-6 Fire Brigade): Accept SUBMITTER: Stanford E. Davis, PA Power & Light Co. COMMENT O N PROPOSAL NO: 805-75 RECOMMENDATION: Change defini t ion for Fire Brigade to read:

"Industrial Fire Brigade, An organized g roup of employees within an industrial occupancy who are knowledgeable, trained, and skilled in at least basic fire-fighting operat ions, and whose fidl t ime occupat ion migh t or might not be the provision o f fire suppress ion and related activities for their employer." S U B S T A N T I A T I O N : The defini t ion for Fire Brigade should be the same as the NFPA preferred definit ion from NFPA 600, Industr ial Fire Brigades. COMMITTEE ACTION: Accept. COMMITTEE STATEMENT: Also change "Fire Brigade" to "Industrial Fire Brigade" everywhere it appears in the s tandard. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #36) 805- 24 - (1-6 Fire Zone): Accept SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute C O M M E N T O N PROPOSAL NO: 805-404

I RECOMMENDATION: Change "canalso" to "can also". S U B S T A N T I A T I O N : Typographial error. COMMITTEE ACTION: Accept. N U M B E R OF C O M M I T r E E MEMBERS ELIGIBLE TO VOTE: 26 VO TE O N COMMITTEE ACTION:

AFFIRMA'I'IVE: 24 N O T RETURNED: 2 Biggins, Vieiva

(Log #89) 805- 21 - (1-6 Electrical Raceway Fire Barrier System (ERFBS)): Accept SUBMITTER: Chr i s topher P ragman , PECO Energy COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Revise defini t ion of ERFBS (as follows:

Nonload-bear ing partition-type envelope system installed a round electrical c o m p o n e n t s and cabl ing that have withstood a fire exposure as establ ished in accordance with an approved test p rocedure and are rated by a test laboratory in hours of fire resistance and are used to mainta in specified nuclear safety funct ions fi-ee o f fire damage . S U B S T A N T I A T I O N : Funct ion should be ~ by some r equ i r emen t or analysis result. Correct terminology is "nuclear safety function." COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

(Log #87) 805- 25 - (1-6 High-Low Pressure Interface): Reject SUBMITTER: Chr i s topher P ragman , PECO Energy C O M M E N T ON PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Revise text as shown:

High-Low Pressure Interface. Reactor coolant boundary valves whose spur ious open ing could potentially rup ture downst ream piping on an interfacing system or cc-:!d cau~c a Ic,~ .vf inventc: 7 that could not be mitigated in sufficient time to achieve the nuclear safety pe r fo rmance criteria. SUBSTANTIATION: Make scope consis tent with NRC GL 81-12 and "Clarification to GA 81-12." COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The Commi t tee believes the s t a tement "or could cause a loss of inventory" is impor tan t and should remain . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

262

N F P A 8 0 5 - - N o v e m b e r 2 0 0 0 R O C

VOTE ON COMMITTEE ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

COMMENT ON AFFIRMATIVE: GARRETT: This c o m m e n t is aff i rmed based upon Industry 's

recognit ion that it is more conservative than the cu r ren t regulat ion, in that it recognizes any potentially unrecoverable condi t ions beyond the specific interfacing system rupture scenario described in the Generic Letters. It is consis tent with Industry practice in def in ing Hi /Low pressure interfaces as any potentially unrecoverable loss o f reactor vessel inventory.

~ T C H F O R D : This c o m m e n t is aff i rmed based u p o n Indust ry ' s recognit ion that it is more conservative than the cur ren t regulation, in that it recognizes any potentially unrecoverable condi t ions beyond the specific interfacing system rup ture scenario described in the Generic Letters. It is consis tent with Industry practice in def in ing Hi /Low pressure interfaces as any potentially unrecoverable loss o f reactor vessel inventory.

(Log #100) 805- 26 - (1-6 High-Low Pressure Interface): Accept in Principle S U B M r r r E R : Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add the following explanatory informat ion f rom Append ix B to Append ix A:

It should be noted that sour ious onen i n~ of reactor coolant v

boundary valves that could ootentiallv run ture downst ream nioin~ on an interfacin~ svstem or the loss o f inventorv that could not be mit iaated in sufficient t ime to achieve the nuclear safety oe r fo rmance criteria are classified as hiah-low nressure interfaces [e.~.. residual heat removal (PWR). nressurizer oower ooera ted relief valves (PORVs) at PWRs. and shu tdown cool ina system at boi l ing water reactors (BWR) t and are subiected to the" more restrictive criteria for circuit analysis and subseuuen t addit ional fire orotect ion features, as d o c u m e n t e d in Chan te r 3 of this

SUBSTANTIATION: More appropria te place for explanatory material . COMMITTEE ACTION: Accept in Principle.

Reword as follows: "It should be noted that spur ious open i ng o f reactor coolant

boundary valves that could potentially rup ture downst ream piping on an interfacing system or the loss o f inventory that could not be mit igated in sufficient t ime to achieve the nuclear safety pe r fo rmance criteria are classified as high-low.pressure interfaces [e.g., residual heat removal (PWR), pressurizer power opera ted relief valves (PORVs) at PWRs, and shu tdown cool ing system at boi l ing water reactors (BWR)] and are subjected to the more restrictive criteria for circuit analysis and subsequen t addit ional fire protect ion features."

Deleted: "as d o c u m e n t e d in Chapter 3 o f this s tandard." COMMITTEE STATEMENT: T he informat ion is no t in Chapte r 3. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #23) 805- 27 - (1-6 Impor tan t to Safety): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add definit ion to read as follows:

Impor tan t to Safety. Structures, systems and componen t s that provide reasonable assurance that the facility can be opera ted without u n d u e risk to the heal th and safety o f the public. S U B S T A N T I A T I O N : Term is used in Chapte r 3 and is no t defined. COMMITTEE ACTION: Accept in Principle.

I odify 3-3.1.3.4 to read as follows: "Plant administrative procedure shall control the use o f portable

electrical heaters in the plant. Portable fuel-fired heaters shall no t be permi t ted in p lant areas con ta in ing e q u i p m e n t impor tan t to nuc lear safety or where there is a potential for radiological releases result ing from a fire." COMMITTEE STATEMENT: Section 3-3.1.3.4 was modif ied to delete the term " impor tant to safety".

NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N C O M M I T r E E ACTION:


(Log #55) 805- 28 - (1-6 Noncombust ib le Material): Accept in Principle SUBMITTER: Franklin D. Garret t , APS/NEI COMMENT O N PROPOSAL NO: 805-238 RECOMMENDATION: Add definit ion to read as follows:

Noncombust ib le Material. A material that in the form in which it is used and u n d e r the condi t ions anticioated, will no t i~nite, burn . suuoor t c o m b u s t i o n , or release f lammable vaoors when subiected to fire or heat. Materials mee t in~ the aualification s tandards as

v

given in Section 3-2.2.4.1 o f this s tandard are de t e rmined to satisfv this definit ion. SUBSTANTIATION: The defini t ion as given is too restrictive. This is consis tent with the defini t ion given in NFPA 220, which is where the defini t ions for combust ib le and l imited combust ib le material come from. The reference to Section 3-2.2.4.1 is that conta ined in submittal 805-238 (Log #269). COMMITTEE ACTION: Accept in Principle.

I Delete the last sentence in the definit ion. COMMITTEE STATEMENT: You can not have a r equ i r emen t in a definit ion. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #101) 805- 29 - (1-6 Recovery Action (New) ) : Accept SLrBMrVrER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404

I RECOMMENDATION: Combine the defini t ions of Manual Action and Repair into new defini t ion "Recovery Action."

Recovery Action. Activities to achieve the nuclear safety pe r fo rmance criteria that take place outside of the main control room or outside of the primary control stat ion(s) for the e q u i p m e n t being operated inc luding the rep lacement or modif icat ion o f componen t s . SUBSTANTIATION: Moves away f rom old controversial terms - - more accurately reflects the type o f actions. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON C O M M r I ' r E E ACTION:


(Log #93) 805- 30 - (1-6 Safe and Stable Condi t ions) : Accept in Principle SUBMITTER: Chr is topher P r a g m a n , PECO Energy COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise defini t ion as follows:

Safe and Stable Condit ions. For fuel in the reactor vessel, head on and tensioned, safe and stable condi t ions are def ined as the ability to mainta in Keff <0.99, with a reactor coolant t empera ture at or below the requ i rements for hot shutdown for a boiling water reactor and hot s tandby for a pressurized water reactor. For all o the r configurat ions, safe and stable condi t ions are def ined as main ta in ing Keff <0.99 ar.~ a fuc! eGG!am tempera ture <200°F and fuel coolant t emoera ture as nrescr ibed bv olant technical soecif icat ions. SUBSTANTIATION: The code should not create pe r fo rmance criteria for the nuclear system (i.e., 200°F) that are not integrated with plant-specific refuel ing and fuel-handl ing l icensing criteria. They should all be consistent. Specifying a specific t empera tu re (200°F) appears to be arbitrary. COMMITTEE ACTION: Accept in Principle.

Revise defini t ion as follows: Safe and Stable Condit ions. For fuel in the reactor vessel, head

on and tensioned, safe and stable condi t ions are def ined as the ability to mainta in Keff <0.99, with a reactor coolant t empera ture at or below the requ i rements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all o the r configurat ions, safe and stable condi t ions are def ined as main ta in ing Keff <0.99 and fuel coolant t empera tu re below boil ing.

263

N F P A 805 - - N o v e m b e r 2000 R O C

COMMITTEE STATEMENT: The change makes it more generic. Boiling temperatures vary depending on location and pressure. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #66) 805- 31 - (1-6 Through Penetrat ion Fire Stop): Accept SUBMITTER: Stanford E. Davis, PA Power & Light Co. COMMENT O N PROPOSAL NO: 805-118 RECOMMENDATION: Change definition of Through Penetration Fire Stop to read:

'A tested, fire rated construction consisting of..." SUBSTANTIATION: The definition needs to specify that the installation of a Through Penetration Fire Stop is a fire rated tested configuration. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #CC7) 805- 32 - (2-2): Accept SUBMrrTER: Technical Commit tee on Fire Protection for Nuclear Facilities COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Change the order of the steps in Section 2-2 as follows:

a) Establish the fundamental fire protection program (See Change 3).

b) Identify fire areas and associated fire hazards c) Identif') the performance criteria that apply to each fire area

(Section 1-5). d) Identify systems, structures, and componen t s (SSCs) in each

fire area to which the performance criteria apply.

e) Select the Deterministic a n d / o r performance-based approach for the performance criteria (See Chapter 4).

Also r ecommend reorder ing sections 2-2.1 through 2-2.8 to reflect the revised order in items (a) through (j) o f Section 2-2: The revised order would be 2-2.1, 2-2.2, 2-2.5, 2-2.3, 2-2.4, 2-2.5.1, 2-2.5.2, 2-2.5.3, 2-2.6, 2-2.8, 2-2.7. SUBSTANTIATION: Editorial comment . The sequence of the steps in the ger/eral approach in Section 2-2 does not match the methodology in Figure 2-2. Recommend revising to more closely align with Figure 2-2. This may help make the process easier to follow and understand. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


VOTE ON COMMITTEE ACTION: AFFIRMATIVE: 17 NEGATIVE: 7 NOT RETURNED: 2 Biggins, Vieira

EXPLANATION OF NEGATIVE: ALPERT: See comments by Sinopoli and Garrett. CARLISLE: I concur with Mr. Sinopoli 's c~mment dated

8/9/2000. DAVIS: I concur with Mr. Rispoli's comment . GARRETT: Performance of a risk informed change evaluation

should only be required when selecting the performance based option of the standard. This is consistent with o ther risk-informed applications utilized in the US nuclear industry such as in the Inservice Inspection Program whereby risk is not considered when selecting the existing deterministic requirements. The current figure implies that this must be done regardless of the option selected, which is in conflict with Section 2-2.5.1, which states that compliance with dete~ninist ic requirements satisfies the s tandard 's performance criteria.

HENNEKE: Since the standard allows both deterministic and risk-informed processes, changes should be allowed that follow ei ther method. Once the baseline is established, changes should be allowed with ei ther method, and deterministic changes do not need risk-informed verification.

RATCHFORD: Performance of a risk informed change evaluation should only be required when selecting the performance based option of the standard. This is consistent with o ther risk-informed applications utilized in the US nuclear industry such as in the Inservice Inspection Program whereby risk is not considered when selecting the existing deterministic requirements. The current figure implies that this must be done regardless of the opt ion selected, which is in conflict with Section 2-2.5.1, which states that compliance with deterministic requirements satisfies the s tandard 's performance criteria.

SINOPOLI: Performance of a risk informed change evaluation should not be required when selecting a deterministic approach. This is consistent with o ther risk-informed applications utilized in the US nuclear industry, in which risk is not considered when selecting the existing deterministic requirements. Figure 2-2 leads one to perform a change evaluation regardless of the option selected. Section 2-2.5.1 states that compliance with deterministic requirements satisfies the standard 's performance criteria.

(Log #19) 805- 34 - (Figure 2-2): Accept SUBMITTER: Edward A. Connell , U.S. Nuclear Regulatory Commiss ion COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add "Existing" to Engineering Equivalency Evaluations box. SUBSTANTIATION: Make consistent with Section 2-2.5.2. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #5) 805- 33 - (Figure 2-2): Reject SUBMITTER: Ronald Rispoli, Entergy Corp. COMMENT O N PROPOSAL NO: 805~t04 RECOMMENDATION: Revise Figure 2-2 to only require Risk Informed Change Evaluation when choosing Performance Based approach. SUBSTANTIATION: Performance of such an analysis should only be required when selecting the performance based option of the standard. This is consistent with o ther Risk Informed applications utilized in the US nuclear industry such as in the Inservice Inspection arena whereby risk is not considered when selecting the existing deterministic requirements. The current figure implies that this must be done regardless of the opt ion selected which is in conflict with Section 2-2.5.1 which states that compliance with deterministic requirements satisfies the standards performance criteria. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: When implement ing any change, a risk-informed change evaluation will need to be performed. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

(Log #2O) 805- 35 - (Figure 2-2): Accept SUBMITTER: Edward A. Connell , U.S. Nuclear Regulatory. Commiss ion COMMENT O N PROPOSAL NO: 805-404

] RECOMMENDATION: Replace "mandatory" with "fundamental". SUBSTANTIATION: Make consistent with Chapter 3. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #102) 805- 36 - (Figure 2-2): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineering Services COMMENT ON PROPOSAL NO: 805-123 RECOMMENDATION: Revise Box 4 from top to read as follows:

Identify structures, systems, or components (SSCs) that apply to erite~fia-in each fire area to which the performance criteria applies.

264


Add "existing" to box with Engineering Equivalency Evaluations so that it reads:

" ~ Engineering Equivalency Evaluations.' S U B S T A N T I A T I O N : More accurately reflects text in Chapter 2. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRNL~,_TIVE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #39) 805- 37- (2-2.7): Accept SUBMITrER: Frederick A. Emerson, Nuclear Energy Institute COMMENT O N PROPOSAL NO: 8054-04

I 2_RI~COMMENDATION: Change "see Section 2-5" to "see Section

SUBSTANTIATION: Incorrect reference used. Correct reference is Section 2-6. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #40) 805- 38- (2-2•8): Accept SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT O N PROPOSAL NO: 805-404

I RECOMMENDATION: Add "See Section 2-5" in italics at the end of the sentence, similar to the format used in Section 2-2.7. S U B S T A N T I A T I O N : If a reference to the appropriate section (2- 5) is used in 2-2.7, it is appropriate to add a similar reference in 2- 2.8, for clarity. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #86) 805- 39 - (Figure 2-3): Reject SUBMITTER: Christopher Pragman, PECO Energy COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise Figure 2-3 as shown on the next page: S U B S T A N T I A T I O N : For a performance-based approach, whether the fire protection systems are designed to a specific code is of secondary importance to whether they can mitigate the fire threat under consideration. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The committee feels that compliance with the applicable installation standards is an important, part of the. requirements. , . The effectiveness of the system is d~scussed m prevmus ooxes. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: KASSAWARA: In a performance-based evaluation, once a feature

is effective in meeting performance criteria, its compliance to a code that in all likelihood is designed to meet different set of performance criteria is not relevant. Compliance to fire codes is important, but as it is used here Under Performance-Based approach is unnecessary and goes against the PB concept.

(Log #71 ) 805- 40 - (2-3.1.3): Reject SUBMITTER: Ira Heatherly, Tennessee Valley Authority C O M M E N T O N PROPOSAL NO: 805-404 RECOMMENDATION: Add text to read as follows:

When using fire modeling, a set of fire scenarios shall be defined for each plant area being modeled, unless o ther considerations clearly exclude the necessity for additional evaluations.

S U B S T A N T I A T I O N : Requiring sensitivity analyses on all areas is not justified due to the significant number of plant areas which are not subject to transient type combustible loads or ignition source concerns. These areas currently have substantial margin to hazard classification changes, have restricted access due to dose considerations, etc. Using this approach will negate the necessity for producing large volumes of analyses which are required to be maintained. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The code does not use the large volume of analysis discussed in the substantiation. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #CC3) 805- 41 - (2-3.1.4): Accept SUBMITTER: Technical Committee on Fire Protection for Nuclear Facilities C O M M E N T O N PROPOSAL NO: 805-404 RECOMMENDATIO N: Revise the section as follows:

2-3.1.4 Defining the Fire Scenario. "A fire scenario shall consider all operational conditions of the plant including 100 percent power, cold shutdown, refueling modes of operation and the following characteristics as necessary to ~upport the requirement~ e,f t,~.i; ;:az:dard meet the required performance criteria." S U B S T A N T I A T I O N : Clarification. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VO TE O N COMMITTEE ACTION:


(Log #1) 805- 42 - (2-3.1.5): Reject SUBMITrER: Rashid Abbas, Tennessee Valley Authority C O M M E N T O N PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Revise text to read as follows:

"(2). . . through opening fire stop and can be installed within a on either side of the barrier..."

Exception: "...with a fire rated seal unless the conduit extends g:'eater than 5 ft on each side of the fire barrier• In this case the conduit opening shall be provided with ~,e.nce.mb'a~t-b!e material a seal to prevent the passage of smoke and hot gasses. The location and materials for hot and ~as seals shall be based on aualified fire tests. ,w.~ ¢,~ a~pt!~ of the material packed to a depth of 2 in. ~a ! ! . . . . . . . . . . . . . . . . . . . v . . . . . . . . . . . . . . . . . . . . . ga; ;eal ;,~, thi~

S U B S T A N T I A T I O N : The internal conduit fire seals should be allowed to be installed within the barrier or on either side of the barrier.

Exception requires that conduits 4 in. diameter and less shall be sealed at the barrier with a fire seal. The exception should also allow installation of fire seals outside the barrier if installed on both sides. The non-combustible seal (smoke & gas seal) installation is only allowed when the conduits extend 5 ft beyond the barrier. This was an old prescriptive requirement. Current fire tests have shown that smoke and gas seals can be installed in conduits which extend at least 12 in. beyond the barrier. The wording in the exception should be changed to allow installation of smoke and gas seals in conduits extending beyond the barrier based on qualified fire tests. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The submitter 's recommendat ion is on an incorrect section and is not applicable to assumptions. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VO TE O N COMMITTEE ACTION:


265

NFPA 805 - - November 2000 ROC

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805- 43 - (2-3.1.5): Reject (Log #72) SUBMITTER: Ira Heatherly, Tennessee Valley Authori ty C O M M E N T ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete the following text:

S U B S T A N T I A T I O N : This r equ i r emen t appears to be in conflict with Paragraph 3-5.4 which requires fire p u m p s to be Seismic Category I Class IE and Paragraph 3-6.4. This would imply that you should postulate a DBE (i.e. seismic event) concu r r en t with the fire.

COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: See Commi t tee Action on C o m m e n t 805-44 (Log #103). NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMMITTEE ACTION:


266


(Log #103) 805- 44 - (2-3.1.5): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT ON PROPOSAL NO: 85-404 RECOMMENDATION: Revise section as follows:

2-3.1.5 Assumptions . T he fo!!awing aaaump:iona =ha!! bc . . . . : . ~ a . . . . . . . :_~ ,~.~ r. . . . . . . . -:~' T he followin~ assumot ions are nrovided to ner form a determinist ic analysis of ensu r ing the nuclear safety oer formance criteria are met. Per formance-based informat ion (i.e.. e u u i n m e n t out o f service. e a u i o m e n t failure unre la ted to the fire. concur ren t design basis even~) are integral parts o f a PSA and should be consider¢~t when ne r fo rmance based anDroaches are uti l ized:

,-,~°~ Thc ~rc accnzria dac= not havc to be pa=tu!ated cencu r r ca t v;ith :,cvcrc nat ' -ra ~. even*,~.

(1) I n d e n e n d e n t failures (i.e.. failures that are not a direct consequence o f fire damage) of systems, equ ipmen t , ins t rumenta t ion , controls, or power supplies do not occur before, dur ing, or followin~ the fire. Therefore . contrary to o the r nuclear power plant design basis events, a concur ren t single active failure is no t reouired to be postulated.

(2) No abnormal system transients, behavior, or design basis accidents precede the onse t of the fire. nor do any of these events. which are not a direct conseouence of fire damage , occur dur in~ or following the fire. SUBSTANTIATION: Reflects NUREG 0800. COMMITTEE ACTION: Accept in Principle.

Revise and move existing 2-3.1.5 to 2-3 and r enumber . Revise as follows: 2-3.1.5 Assumptions . T he followin~ assumpt ions are nrovided to

perform ~ determinist ic analysis o f ensu r ing the nuclear safety oerforrnance criteria are met: [Performance-based informat ion (i.e.. e q u i p m e n t out o f service, e q u i p m e n t failure unre la ted to the fire. concu r r en t design basis events) are integral harts o f a PSA and shall be conzidered when ne r fo rmance based aoDroaches are utilized:] (1) I n d e p e n d e n t failures (i.e.. failures that are not a direct

consequence of fire damage) of systems, equ ipmen t . ins t rumenta t ion , controls, or power supplies relied upon to achieve the nuclear safety pe r f romance critera do not occur before, dur ing, or following the fire. Therefore . contrary, to o the r nuclear nower plant ~l¢~ign basis events, a concur ren t single active failure is no t required ¢O be nostulated.

(2) No abnormal svstem transients, behavior, or design basis accidents precede the onset of the fire. no r do iany of these events. which are not a direct conseouence o f fire damage , occur du r ing or following the fire. COMMITTEE STATEMENT: The assu'mptions were moved up front earlier in the d o c u m e n t and revised for clarity. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #21) 805- 45 - (2-3.2): Accept S U B M r r r E R : Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Add "performance-based or" pr ior to "risk- informed" in last sentence. SUBSTANTIATION: Performance-based or risk in fo rmed al terat ions would be acceptable. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #110) 805- 46 - (2-3.2(4), 2-3.2.4, 4-1.2, 4-1.3.3, 4-1.3.3(b), 4-1.3.3(c), 4- 1.4.1.1, 4-1.4.1.3, 4-1.4.1.4, 4-1.4.1.6, A-4.1.4): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Replace the term "analysis area" everywhere with "fire area."

Delete the definit ion o f "analysis area."

SUBSTANTIATION: More consistent terminology with that in the industry. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITrEE ACTION:


(Log #6) 805- 47 - (2-3.2.2.1): Accept in Principle SLrBMITTER: James E. Lechner , Nebraska Public Power District COMMENT ON PROPOSAL NO: 805-1 RECOMMENDATION: Revise text to read as follows:

"...open circuits and shorts to ground, and ahc, rt circui*~..." SUBSTANTIATION: Te rm "short circuits" is duplicate of "hot shorts". This is consis tent with industry vernacular for circuit faults. COMMITTEE ACTION: Accept in Principle.

Revise 2-3.2.2.1 to read as follows: 2-3.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits

required for the nuclear safety funct ions shall be identified. This includes circuits that are required for operat ion, could prevent the operat ion, or result in the malopera t ion o f the e q u i p m e n t identified in 2.3.2.1. This evaluation shall consider f ire-induced failure modes such as hot shorts (external and internal) , open circuits and shorts to g round , to identify circuits that are required to suppor t the proper operat ion o f componen t s required to achieve the nuclear safety pe r fo rmance criteria, inc luding spur ious o p e r a u o n and signals. This will ensure that a comprehens ive populat ion of circuitry is evaluated. (See Appendix B for considerat ions in analyzing circuits). COMMITTEE STATEMENT: The commit tee revised the wording to reflect cur ren t industry guidance. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #85) 805- 48 - (2-3.2.2.1): Accept in Principle SUBMITTER: Chr i s topher P r agman , PECO Energy COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise the third sentence of 2-3.2.2.1 as follows:

This evaluation shall consider f i r eqnduced failure modes such as hot shorts (external and internal) , open circuits, shorts to g round , and short circuits, to identify circuits that are required to suppor t the proper operat ion of componen t s required to achieve the nuclear safety pe r fo rmance criteria, inc luding spur ious opera t ion and signals. SUBSTANTIATION: Clarify that the failure modes of interest are fire induced. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t tee Action on C o m m e n t 805-47 (Log #6). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #104) 805- 49 - (2-3.2.2.1): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT ON PROPOSAL NO: 85-404 RECOMMENDATION: Revise text as follows:

This evaluation shall consider failure modes such as hot shorts (external and internal) , open circuits, and shorts to g round , a n d shor t circuits, to identify circuits that are required to suppor t the proper operat ion of c o m p o n e n t s requi red to achieve the nuclear safety per formance criteria, inc luding spur ious operat ion and signals. SUBSTANTIATION: Accuracy reflects cu r ren t industry guidance. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: The proposed change is in the third sentence.

See Commi t tee Action on C o m m e n t 805-47 (Log #6). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


267


(Log #105) 805- 50 - (2-3.2.2.1): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Eng inee r ing Services COMMENT O N PROPOSAL NO: 85-404 RECOMMENDATION: Revise as follows:

See Appendix B for ~cceptable methoda -aaed to analyze considerat ions in analyzing circuits. SUBSTANTIATION: More accurately reflects what is in Appendix B. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t t ee Action on C o m m e n t 80547 (Log #6). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #107) 805- 51 - (2-3.2.2.2(a)): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Eng inee r ing Services COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

(See Appendix B for accc~ptablc mctbc.da uaed to. analyze considera t ions when analyzing c o m m o n power supply concerns . ) SUBSTANTIATION : More accurately reflects what is in Appendix B. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #106) 805- 52 - (2-3.2.2.2(b)): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 85404 RECOMMENDATION: Revise text as follows:

"The concern is that the effects of a fire can ex tend outside of the immedia te fire area due to f ire-induced electrical faults on inadequately protected cables or via inadeouatelv sealed fire area boundar ies . (See Appendix B for acceptable m e t h ~ a uae~ t~,

considerat ions when analyzing c o m m o n enclosure c o n c e r n s . ) "

SUBSTANTIATION: More accurately reflects cu r r en t industry guidance and what is in Appendix B. COMMITTEE ACTION: Accept. NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #108) 805- 53 - (2-3.2.3): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

"(See Appendix B for acceptable meth~d~ u~ed ta i~cntih/ considerat ions when identifying locations.)" SUBSTANTIATION: More accurately reflects what is in Appendix B. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #109) 805- 54 - (2-3.2.4): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

"(See Appendix B for acceptable :netha~a aaed to, pcffe.rn.~ considerat ions when pe r fo rming the fire area n'ac!car av~tety assessments . ) " SUBSTANTIATION : More accurately reflects what is in Append ix B. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #57) 805- 55 - (2-3.4.1): Accept SUBMITTER: Dennis H e n n e k e , Sou the rn California Edison COMMENT O N PROPOSAL NO: 805404

I RECOMMENDATION: Change "made available for review by the AHJ" to "performed". S U B S T A N T I A T I O N : The entire FHA is available to the AHJ. Section 2-3.3 discusses base risk evaluations and these are "performed" us ing acceptable methods . It is inconsis tent to require selected par t s and made available to the AHJ. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #41) 805- 56 - (2-5.3): Reject SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

"If the establ ished levels of availability, reliability or per formance are no t met , appropr ia te corrective actions to return to the establ ished levels shall be imp!emente~ . Manito,rlng o"ha!l hc con t inued to cn~'arc that the corrective actiona arc effective identif ied for the corrective ~action program." S U B S T A N T I A T I O N : This s tandard should not set priorities for the plant corrective action program. It is sufficient for this s tandard to identify potential inclusions in the corrective action program. COMMITTEE ACTION: Reject. C O M M r r r E E STATEMENT: The proposed change only indentif ies and does not require a corrective action to be taken. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #22) 805- 57 - (2-6.1.1): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add "mainta ined for the life o f the plant and" after "documenta t ion" in last sentence . SUBSTANTIATION: Add a r equ i r emen t for re tent ion of documen ta t i on . COMMITTEE ACTION: Accept in Principle.

I Add text to read as follows: "Shall be main ta ined for the life of the plant " after

"documenta t ion" in the last sentence. COMMITTEE STATEMENT: The revised wording makes this mandatory . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


NOTE: Since the ballot on this C o m m e n t did not conf irm the Commit tee Action, the co mmen t is being rejected.

(Log #82) 805- 58 - (Chapter 3 and Appendix A) : S U B M r r T E R : Clifford R. Sinopoli, II, PECO Nuclear, Peach Bot tom Atomic Power Station COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: The following is a comple te revision o f Chap te r 3 and the sections in Appendix A related to Chapter 3:

Chanter 3 Fundamental Fire Protect ion Progrdm and Design Elemen~

3-1" General . This chaDter conta ins the fundamen ta l e lements o f the fire orotect ion Program and soecifies the m i n i m u m design / r eou i rements for fire nrotect ion systems and f~atures. These fire orotect ion n rogram e lements and m i n i m u m design requ i rements shall no t be subject to the oer formance-based me tho d s uermi t ted elsewhere in this s tandard. Previously aonroved alternatives from the fundamen ta l orotect ion program attr ibutes o f this chan te r bv the AHI take Drecedence over tlae reou i rements ¢gnta ined h¢rein,

3-2 Fire Protect ion Plan.

268


3-2.1 Intent . A site-wide fire Drotection plan ~hall be e~Iablished. This nlan shall d o c u m e n t m a n a g e m e n t policy and p rogram direct ion and shall define the responsibiliti¢~ o f those individuals resnonsible for the n lan ' s imnlementa t ion . This section establishes the criteria for an in tegrated combina t ion of comnonen t s . Drocedures. and pe rsonne l to i m p l e m e n t all fire protect ion program activities,

3-2.2* Management Policy Direction and Responsibility, A DoIicv d o c u m e n t shall be orenared that def ines managemcn~ authori tv and resoonsibilities and establishes the general nolicv for the site fire orotect ion nrogram. T he policy ~hal laddress th¢

1) M a n a g e m e n t authority, and responsibility 2) Program i~dministration and control 3) Interface with o ther organizat ions 4) Applicable Authori ty I~aving Jurisdict ion (AHD for p rogram

e lemen t s ~ - - r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . P r o g r ~ a=d Den

!* Cenem~. Thia chapter cantaina the fundamen t a l c lcmcnt~

"'rcccd~r:cc sv'~r t!;c rc-uirc~,.c~ts cantair,,cd !:groin.

p &...,. ~ . . .~ . . . . . . . .~ 0...., = ~ . . , . . . . . . -~ov . . . . . . . . . . ,w

. . . . . . . . . . . . . . . a . . . . . . . .a peraonnc! to i m p l e m e n t all fire

¢'I q : k 11 .1 + . . . . . . . . + D ~ I I = . 1 1 ~ l - - ~ + + l . . . . ,..,11 D . . . . . . . ~ k l l l ~ . A

the aite fire pro tec t ion program.

m a n a g e m e n t - ~ " : + : . . . . : , r . ; ~ A : ~ , . . . . * L .~ .% . . . . , 4 . . . . . . . : l . : l ; + . ,

rcapon~iS!e for the dai!y adminis t ra t ion and cocrd ina t lon of the fire protect ion program and iu implementa t ion .

3 . . . . . . . . . 1 " . . . . ! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 " . . . . . . . . . .

~ ] . . . . . : ~ t E c . .~ .a . . . . . . . . . . e ++.~ ~ , ~ p , - o t c c t i a n p r o g r a m .

3-2.3 Procedures . Procedures shall be establ ished for implementa t ion of the fire protect ion program. In addit ion to procedures that might be required by o ther sections o f the s tandard, the p rocedures to accompl ish the following shall be established:

(a)* Inspection, testing, and ma i n t enance for fire protect ion systems and features credi ted by the fire protect ion program

(b)* Compensa tory actions i mp l emen t ed when fire protect ion systems and o ther systems credi ted by the fire protect ion program and this s tandard canno t per form their in tended funct ion and limits on impa i rmen t dura t ion

(c)* Reviews of fire protect ion p rog ram- re l a t ed per fo rmance and t rends

(d) Reviews of physical p lant modif icat ions and p rocedure changes for impact on the fire protect ion program

(b) Long-term ma in tenance and conf igurat ion o f the fire protect ion program

(c) Emergency response procedures for the plant fire br igade (ROP Log #19)

$-3 Prevention. A fire prevent ion program with the goal of prevent ing a fire from star t ing shall be established, documen ted , and imp lemen ted as part of the fire protect ion program. The two

' basic c o m p o n e n t s o f the fire prevent ion p rogram shall be 1) prevent ion o f fires and fire spread by controls on opera t ions activities, and 2) design controls that restrict the use o f combust ib le materials. The design control r equ i rements listed in the r ema inde r o f this section shall be provided as described.

3-3.1 Fire Prevention for Operational Activities. T he fire prevcnt ion program activities shall consist o f the necessary

e l emen t s to address the control o f ignition sources and the use of t rans ient combust ib le materials du r ing all aspects of f)lant operat ions . The fire prevent ion program shall focus on the h u m a n and programmat ic e lements necessary to prevent fires from star t ing or, should a fire start, to keep the fire as small as possible.

3-3.1.1 General Fire Prevent ion Activities. The fire prevention activities shall include but not be l imited to the following program elements :

(a) Tra in ing on fire safety informat ion for all employees and contractors including, as a m i n i m u m , familiarization with plant fire prevent ion procedures , fire reporting, and plant e nergency a l a rms

(b)* D o c u m e n t e d plant inspect ions inc luding provisions for corrective actions for condi t ions where unanalyzed fire hazards are identified

(c)* Administrat ive controls address ing the review of plant modif icat ions and ma in tenance to ensure that both fire hazards and the impact on plant fire protection systems and features is min imized

3-3.1.2" Control of Combustible Materials. Procedures for tile control of general housekeep ing practices inc luding and- the control o f t ransient combust ibles and combust ib le storage shall be developed and implemented . T!=e~e prc.ccdurca ahal! inc!ude l;ut

.... r . ~e, . . . . . . or . . . . . . . . . . . . a . . . . . . . . . . rc tardant a cation. Exc@tion: Cri55i=.g ti:nbcrs ~ 5:. ~i 6 in. vr L~rg~" .;hatl-eob~

....... • ~'t'+:+"'s t a 5tr i te rcte.rdar, t treated.

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . an area imn 'ed ia te ,

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7~" . . . . . . q u m q t i t i e a o f ~ : o r e ~

accordance with applicaMc NFPA atandard~. 3-3.1.3 Control of Ignition Sources. 3-3.1.3.1" A hot work safety procedure shall be developed,

imp lemen ted , and periodically upda ted as necessal T in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeg~tarding Con.~truction, Alteration, and Demolition Operatio..s.

3-3.1.3.2 Smoking and o ther possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant.

$-3.1.3.3 O p e n flames or combus t ion-genera ted smoke shall no t be permi t ted for leak or air flow testing.

$-3.1.3.4" Plant administrative procedure shall control tire use of portable electrical heaters in the plant. Portable tirol-fired heaters shall no t be permi t ted in plant areas conta in ing e q u i p m e n t impor tan t to safety or where there is a potential for radiological releases resul t ing from a fire.

$-3.2 Structural . Walls, floors, and componen t s required to main ta in structural integrity shall be of noncombus t ib le const ruct ion, ~ def ined in NFPA 220, Standard on 7~,pes q[ Building Construction.

$-3.3 Inter ior Finishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101 ®, L![e Sqfetv Code ®, requ i rements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requ i rements for Class I inter ior floor finishes.

$-3.4 Insula t ion Materials. The rma l insulat ion materials, radiat ion shie ld ing materials, ventilation duc t materials, and s o u n d p r o o f i n g materials shall be noncombus t ib l e or l imited combus t ib le .

$-3.5_* Electrical. Electrical des ign shall address i~nition sources and combust ib les associated with cable and raceways and at a m i n i m u m addrcs~ the following:

1) Wir ing above susnended ceilings 2) Trays and condui ts 3/ Cable const ruct ion

3.5.! Wiring . . . . . . . . . v . . . . . . . . . . . . . ~ . . . . . . . . . . . v . . . . . .

269

N F P A 805 - - N o v e m b e r 2000 ROC

, , L 3 / £ o . ; , . Z _ , ; 2 . T ; ~ , 2 2 U ~ o 2 . 5 7 `+ r . . . . . . . . . . . . . . . . . . . . . . . . . # . . . . . . . . . . . . . . . . . . .

3-3.6 Roofs. Metal roof deck construction shall be designed and installed so the roofing system will not sustain a self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building. Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Method o[Fire ]?sts o[Roof Coverings.

3-3.7_* Bulk Flammable Gas Storage. Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety. Storage requirements shall address the following are~_L

1~ Location of containers 2) Orientation of containers 3) Isolation of containers

~.7.! S to rage of f l a m m a S l c gaz 3Ea!l bc l oca t ed a u : d a o r z , o r in

. . . . . . . . . . . . . . . . ~ , , ~ . . . . . . . . . . . . . . . . . . . . gac, 3toragc containers ~ha!! be loca ted ~ o t h a t the I o a g a'xi~ i ; =or po i r . : cd at bu i ld ings .

3-3.7.3 Flammable gas storage cylinders not required for normal operation shall be isolated from the system.

3-3.8 Bulk Storage of Flammable and Combustible Liquids. Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing systems, e.quipment, or components important to nuclear safety. As a mlmmum, storage and use shall comply with NFPA 30, Flammable and Comb~mtible Liquids Cnde.

3-3.9* Transformers. Where provided, transformer oil collection basins and drain paths shall be periodically inspected to ensure that they are free of debris and capable of performing their design f imct ion .

3-3.10" Hot Pipes and Surfaces. Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact With hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation.

3-3.11 Eleclricai Equipment. Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment.

3-3.12" Reactor Coolant Pumps. For facilities with non-inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed aud installed such that leakage from the oil system is safely contained. The followin~ areas shall be addressed in system design: or of f n o r m a ! cond i t i on~ w-ch ~

~ u s

i ~ Leakage :halt be . . . . . . . . .

1) Collection and storage 2) Flame arrestors 3) Leakage ooints

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g.,.,,,p~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v.u~, . . . . . . . . . . * ; . . . . . . ; I I ; . . . . . . 4 +K~ ~ ; I . . . . . . . ; . . . . . 1.. . . . . . . . K

3 4 Fire Brigade.

3-4.1" On-Site Fire-Fighting Capability. ?d! of the fo!!owing

(a) - ~ } A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty. The following areas relating to the fire brigade shall be addressed: . . . . . . . . . . . . . . . . . . . . v.': . . . . . . e .o . .v . ; . . . 6 xtFpA : t a n d a r d :

. - - h i ; ~ . k l ~ .

(1) Determination of applicable NFPA Standard for the site fire

(2) Limitations on collateral duties (3) Fire brigade member knowledge of plant systems (4) Immediate notification of fire brigacle upon verification of

a fire. (5) Annual physical examinations

/O \ ~N.T]ff'DA IK /~ / ' I ~21 . . , . I ~ . . . 1 . . . . . ~ ' . ; . . . . F I . . j . _ .+ . . . . . . t { } . . . . , * J , . , l d . . . . . I Qo/ ' . , t . ,

. . . . . 1 l f f . , . h l , I . I , . . . . . . . .

L ' ; . . . . 1 ~ T~TWDA 1 ~ Q O ~ . . . , ~ / __ . J ~_ AAI .Ad~ . I I ) . , ) . . , , ~ . . . . . . . . . . I . J fG ; " • ~ , ,

k .~ - - .A . . . . I . . . . . k - - I I k . . . . . . . g ' ~ - . ; - - - - + + . - . : - - 1 . . . . A I . . . . . . I - -A - - ^ - -C

~ u p p r c ~ a n ~ , o n n u c l c a r ~ a f c t y p e . ' f o . ~ , . a n c c c r . t c r i a , Exccptic,:; tc / ~ e . . ¢ c . . : . . . . , , . . . = . = . . . . . . n :. . . . . . . . ~+~ . . . . i,+11 ~ , . . , . . . . . :,,...s

. ~+ . . n+ . . . . . . . :.--~,I d . , . ; . . . . . . . . I g~ . . n l~ ,~1~+ ; . . . . . . . . + : ~ - - 0 T h e

3-4.2* Pre-fire Plans. Current and detailed pre-fire plans shall be available to the fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1-5. (ROP Log #187) The ore-fire olans shall address the followin~ areas:

(1) Configuration. hazards, nuclear safety components, and f ire protection features of the ~rqa

(2) Review and uodate proces~ (3) Availabilitv of the ore-fire nlans (4) Coordination with other olant grouns during fire

emergencies

present.. ~.2.2 Pre f i re plan~ ~hal! bc rc; ' .e-':cd a n d u p d a t e d ",~ n e c c ~ a r / . {.2.~* Pro .Src piano a!'a]! be avai laSlc in the con t ro l re, ore and

3-4.3 Training and Drills. Fire brigade members and other plaut personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.

(a) Plant Fire Brigade 77aining. All of the following requirements shall apply:

(1) Plant fire brigade members shall receive training consistent with the requirements contained in NFPA 600, Standard on Ind'ustrml Fire Brigades, or NFPA 1500, Standard on Fire Departme'nt Occapational Safety and Health Program, as appropriate.

(2) Fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire.

(3) A written program shall detail the fire brigade training program•

(4) Written records that include, but are not limited to, initial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill attendance records, and leadership training for fire brigades shall be maintained for each fire brigade member.

(b) Training for Non-Fire B:~gade t'er~onnel. Plant personnel that respond with the fire brigade shall be trained as to their

270


responsibilities, potential hazards to be encountered, and interfacing with the fire brigade.

(c)* Drills. Drills shall be conducted quarterly for each shift. The following shall be addressed by the fire protection program. M1 ^c,t.~ c~,, • _~q..: . . . . . . . . . . i..~11 ~pp,y. ~ . . . . ~ .~..owlng , ~ ~ . . . . . . . . . . o . . ~ . . . . . .

1) Drill scenarios and locations 2) Drill records 3) Post-drill critiques

,~, . . . . . . tgad . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . [ . . . . . . . . . . ,

F.'~ A ~ . . ; * ; . ~ . k ~ l l I . ~ K ~ I A ~ . .M I ,,4 . . . . . . . + ~ A . C . . . . I . A . - ; I ] (=, , . . . . . . q~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . te . . . . . . . . . . . . 3-4.4 Fire-Fighting Equipment. Protective clothing, respiratory

protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shah be provided for the fire brigade. This equipment shah conform with the applicable NFPA standards.

3-4.5 Off-Site Fire Department Interface. 3-4.5.1 Mutual Aid Agreement. Off-site fire authorities shall be

offered a .plan for. their interface during fires and related emergencies on stte.

3-4.5.2* Site-Specific Training. Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited to participate in a drill at least annually.

3-4.5.3* Security and Radiation Protection. Plant security and radiation protection plans shall address off-site fire authority response.

3-4.6* Communications. An effective emergency communicat ions capability shall be provided for the fire brigade.

(ROP Log #127aa) 3-5 Water Supply. 3-5.1 A fire protection water supply of adequate re iability,

quantity, aud duration shall be provided by one of the two following methods:

(a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies.

(b) Calculate the fire flow rate for two hours. This fire flow rate shall be based on 500 ga l /min (1892.5 L /min ) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system(s) in the power block as determined in accordance with NFPA 13, Standard .fi~r the Installation of Sprinkler Sy.~tems, or NFPA 15, Standard for Water Spray Fixed System.~ for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service. (ROP Log #70)

3-5.2* The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water 7"anks for Privale ]'}re Protec[ion.

Exception No. 1: Water storage tanks shall not be required when fire p'umps are able. to take suction from a large body of water (s'uch as a lake), provided each .fire pump has its own s'uclion and both suctions and pumps are adequately separated.

Exception No. 2: Cooling tower basins are an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands oJ the fire semiee.

3-5.3* Fire pumps, designed and installed in accordance with NFPA 20 Standard for the InstaUation of Stationary Pomps Jor Mre Proteetion, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of tbe largest pump or pump power source.

3 5.1 At lea~t o.~m dicacl engine driven firc pump or two ,.'nnre . . . . . . . . . . . . g ' 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . V . . . . . V °

connected to redundant C!"~a IE emergency po.';er h'.:aca capah!c

o~.~n r. . . . . . . . :.~=a (Existing 3-5.4 deleted, basis for deletion is that 3-5.3 actually addresses this issue.)

(Existing 3-5.5 delet~ed, basis for deletion is that 3-5.3 addresses the loss of fire pump~.)

3 E.5 Each pump and itz dr 'ver and contrM~ aha!! bc acparated from the ; 'emairing fire pu:np~ and from the rcat of the v,-,,~'~-" b;,' rated fire 5arrier~.

3-5.46 Fire pumps shall be provided with automatic start and manual stop only.

3-5.5_7- Individual fire pump connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections.

~ . - - - , ~ . + ; . . . . . . + . . . . . . . , . . . . . k - - l l | . . . . . . . ; A o A : ~ A . . . . A ~ . . + --g" * k - - 4 ~ . . ~

(Existing 3-5.8 deleted, basis for deletion is that NFPA 20- 1999, Paragraph 2-19.4 states that the main fire pump shall not be used as a pressure maiuteuance DumP.)

room, ~,̂ ~ ~+~'~'- . . . . . . . . . a:itahlc conatant!;. . . . . . . ,,~'~a .,~' . . . . . . . . . . ; ~ , of operation

(Existing 3-5.9 deleted, basis for deletion is that NFPA 20-1999, Paragraphs 7-4.7 and 9.4.2 states that alarms be provided at a ooint of constant attendance.)

3-5.1416-- An underground yard fire main loop, designed and installed in accordance with NFPA 24, Standard for the Installation of Private Fire Selvice Mains and tkeir Appurtenances, shall be installed to furnish anticipated water requirements. The design and iustallation shall address the following attributes of the fire main system:

1) Impact of isolation of the fire main 2) Compatibility of threads with local fire departments 3) Construction of headers off the fire main 4) Isolation capability of suppression and standpipe systems 5) Control of valves 6) Hydraut spacing 7) Restriction of fire system usage for non-fire orotection

up___u_~ o se s. {NFPA 24 no longer exists. Replace this reference with another

document? 13 or 230?] = ~ ~ ~ . . . . ahall bc provided to iac, latc p^- ' : . . . . . c ,~. . . . . . -I ~..=

fire hoac atatinm provided for man'aal 5ack'-p. Sprinkler a)'.~temc and man-.:a! hoac oration atandpipca aha!l be connected to the plant fire protection ;;'afar main an that a aing!e acti;'c failure or a crack to t,hc water aupply piping to these ayatcma can hc iaolatad an as no to impair beth the prima&' and 5ackup fire aupprcaaion systcm~. /t,C',D, . . . . Log . . . . . +

• '~ . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . r . . . . . . . . ; . . . . r . . . . . . . . . . . . . . . . . . ;

d;7;,~;.t;::;.:: t ....... : ........... ,. . . ...,.~ p ...... ~ . . . . . . . . . . p.r=..:2.~ " t " ° t . . . . . . . . t~ a d c q : ; . a t c t r a : : ; : : ; g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

. . . . . . . . . . . . . . . d f . . . . . . . . . . . . . . . . . . . . . . p e r m . . . . . . . . . . . . .

Code ;;..°,,,,.:, P +i:;g . . . . . . . . . . . . . . . . . . . . . . . ~ : v . . . . . . including the first valve) o~vv.z.., ~-';-'~ *~'-,,.~ aprinklcr a;'atema. ;;'here

extenaian of the yard main ayatem. Each aprink!cr and atandplpe

(a) Eleet:~.ca! aupc:Maion ;;~.th audible and ;a.a-aal aigna!a in the

. . .L, : . . . . . . . . ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . Y . . . . . . . . . . . . . . Y= . . . . . . . . . .

the direct control of the o;; 'ncr/operator.

, , , / . p . . . . . . . . , ~ y~rd . . . . . . . . y . . . . . . . . . . . . . . . . . . . . . " ~ ' t ' t " . . . . . . . .

s p ~ ; C ; . . ~ l 1 . . NT'IK'DA O A K ' ~ , d - . . d r . . . . , , t , . , L . o , - I I - , . : ~ , 3f 0. .2. .- , . £ ' ; - . ,

~ . - - , ; . . . , hA . . ; . . . . . . . -# 7"1 , .2 . . A l ,a . , . , , .4 . . . . . . . . . . . . I ~ . 1 1 I~ . . . . . . ; *4- -*4 . +

• ~ A L ' ~ ) A 0 /11 - [ S a m e - n o t e a s a ~ v c regarding . . . . . . . . .

271

N F P A 805 ~ N o v e m b e r 2 0 0 0 R O C

. . . . : . . . . . . . . . . . . . . ~:" '~ L 7 t h ; ~ ' c h a ~ c / ' c : a ; e ~ . °"#'E;'k. . . . . . ~ . . . . .

, ~ . . . l . l d . . . . K T . . 1 . I 2 , ; - . • , , . . ~ 1 . . ~ 1 2 . . . . . . . # . . . . . . . a . ~ l . . . . . . ~ . . . . . . . 1 , . . 1 1 1 . . • . . . . . . 1 t t . 2

r . . . . . . . . . . . . . . . . . . v r v ~y~tcm:; arc dc.; ;'d a:;d ;aai::taigwd ta d;:!iv'.~.: , i . . . . . . . ; . : _ . . , ' c . . . . . 2 .. . . . . . I . . . . . . t.,,., n . . . . .n . . . . . . . . 2~ f_ . . , t . . d u r a t i c l g '

~7~.~.,;~. r.r.. ~" [~rt ];rctcclicT; zJa&T .~l~.ragc car; ~. l ............. y r~"~/

:.pc'c~%d darati~:; a~ d;tcrmi~wd i:: this ~ectic:~. 3-6 Standpipe and Hose Stations. 3-6.1" ~ P _ p o w e r block buildings, Cla~s !!I atandpipc and

t. . . . . . . . . . . . shall be provided with standoioes and hose systems installed in accordance with NFPA 14, Standard. for the [nstallation of Standpipe and Hose Systems. The design of the standpipe and hose systems shall address the following:

1) Adequate pressure and flow at each hose station, including orotection affainst excessive pressure

2) Nozzle selection 3) Availability of hose stations following seismic events

n . . . . . . . . . . .~ .... ~- bility- . . . . . . . . . . . . . . . . . . . . pressure for all hose stations, This capa inc!udca the pro;'iaion of hose station pressure rcduccrs where ncccasa,~, for the safeg' of plant fire ,..-:_~.a . . . . k . . . . . a ~ o : , ~

e . a T h e p r c , p c r t y p e o f h c s e n e . z - " ' e t o b e a u p p l i c d te, e a c h

in areas ,,a. . . . . ,. . . . . . : - ~ . . . . . . . . cause unacceptable damage . . . . . . . . . . . . . . . . e, . . . . . . . . . . . can c,r present an electrical hazard t o fire fighting pcrsc.nnel. Listed ~'-~" " "-, safe .q×ed fag nc.zzlea shall be . . . . a,a~.a ~ * 1 . . . . : ~ - - . . . . k - - . - - k ; ~ l ~ . , ~ 1 . . . . . l . - - - - I r k . . . . . 4 . . . . ; ~ , A l l h . . . . . . . . 1 ~ . . k . I I k . . . .

t o ,Cu l l c l o a c a .

5 A Pro;•.sic, n• shall hc made to supply water a t Icc~t to

nuclcar safety functiona in thc cvent of a safe shutdown carthquake

I ~ . . . . . . . . . . . . a l ~ " f i ~1 .~2~ .~ • . . . _ • . . . . . . h - I I L . . . . . 4 . , " P I , 2 . • . ~ _ . . 2 . 2 . . . . I

.......... , ~._. c_,,,: . . . . . . . . ~,,a,,.n ............... ~y~tc~= ~ha!l be capab!g af

"1"1 . . • . . . . . 2 . 2 . . . . f . . . . . . . i . l . l d ~ l , 2 . . . . ~ 1 , ; ~ • . . . . . . 2 ~ : . . . . 1 . . . . t . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . s . . . . . . t . . . . . . . . . . . . . . . . system shall 5e rc p!a:;::ad a.r,d 5c capa!;!c of. 5ci:;.g i~.p!c:::c~.:tM i:;. :: oo"-'"'-°l'v ma::r;,e~ f . . l l ~ ~ . , g . , ~ ( ' ¢ ' ~

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . n fi[~ ~. . . . . . . . . . . . . . . . . . . 1~1~,;

w . . . . . . . . . . . . . . . , the fire tic..;- shall not degrade the essential water system requiremcat .

3-7.1 IT~re Extinguishers. Where provided, fire extinguishers of the appropriate number, size, and type shall be ~ pr~9~4ed in accordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions.

3-8 Fire Alarm and Detection Systems. 3-8.1 Fire Alarm. Alarm initiating devices shall be installed in

accordance with NFPA 72, National Fire Alarm Code*. Alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervismy signals, and trouble signals to the control room or other constandy attended location from which required notifications and response can be initiated. Personnel assigned to the proprietmy alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted:

(1) Actuation of any fire detection device (2) Actuation of any fixed fire suppression system (3) Actuation of any manual fire alarm station (4) Starting of any fire pump (5) Actuation of any fire protection supervisory device (6) Indication of alarm system trouble condition (ROP P319(a) Committee Proposal deleted 3-6.1.2) 3-8.1.1 Means shall be provided to allow a person observing a

fire at any location in the plant to quickly and reliably

communicate to the control room or other suitable constantly attended location.

3-8.1.2 Means shall be provided to promptly notify the following of any fire emergency in such as way as to allow them to determine an appropriate course of action:

(1) General site population in all occupied areas. (2) Members of the fire brigade and other groups support ing

fire emergency response. (3) Off-site fire emergency response agencies. Two independent

means shall be available (e.g., telephone and radio) for notification of off-site emergency services.

3-8.2 Detection. If automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes.

(ROP Log #296 and #74 deleted 3-6.2,2.1) (ROP Log #127ad deleted 3-6.2.2.2) 3-9 Automatic and Manual Water-Based Fire Suppression

Systems. 3-9.1" If an automatic or manual water-based fire suppression

system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including:

(1) NFPA 13, Standard for the Installation of Sprinkler Systems (2) NFPA 15, Standard for Water ,Spray Fixed Systems for Fire

Protection (3) NFPA 750, Standard on. Water Mist Fire. Protection Systems (4) NFPA 16, Standard for the Installation of bbam-Water Sprinkler

and Foam-Water ,Spray ,Systems 3-9.2 Each system shall be equipped with a water flow alarm. 3-9.3 All alarms from fire suppression systems shall annunciate

in the control room or other suitable constantly attended location. 3-9.4 Diesel-driven fire pumps shall be protected by automatic

sprinklers.

~.t~ . . . . . . . . . . . a . . . . . . nr.,_,.,_ (3-9.5 deleted, basis for deletign alreadv addressed bv 3-5.)

9.5 All -'ah'ca controlling watcr baaed fire a'apprcaaion systems • ~'1 ~''~:'-̂ '~ ~ . . . . . . . . . . ~ ...~L'~ performance or dctcrministic rcqaircmcnts of Chapter A shall be s'dpe:'Aacd ",,a described in ~ 5.!~. /.3-9.6 deleted, basis for deletion, already addressed by 3-5.)

3-10 Gaseous Fire Suppression Systems, 3-10.1 If an automatic total flooding and local application

gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes:

(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems (2) NFPA 12A, Standard on llalon 1301 Fire Extinguishing System.~ (3) NFPA 2001, Staodard on Clean Agent k~re Extinffaishing

Systems 3-10.2" The design and installation of a gaseous fire suppression

system shall address the following: 1) Annunciation . . . . . . . . . r . . . . . . . . . . 8 . . . . . . . . . .

sappressian •)'stems shall annanclate and alarm in d'c control

2) Ventilation system impact on the gaseous system 3) Primary and hackuD system independence 4) Control of system disarming capabilities 5) Limitations and controls on tbe use of total flooding carbon

dioxide systems. 6) Thermal shock 7) Corrosive decomDosition orodncts

~-~ . . . . . . . . . . u . . . . . p . . . . . . . t . . . . c . . . . . a n d c c ' , n f i n c m c n t o f

radioacti;'e :antaminanL~. !gA* Ia an;,' area required to be . . . . . . . . "~ by bo~.h ~-; . . . . . . . "~

. . . . . 1 o ~ . . . . . . l ] . r e s u p p r c 3 3 i G n a y s t e m g , a ~ : ~ g ~ . . . . : . ,~ c~:~ . . . . o r a

. . . . . . . . . . . ; . . . . . . + . . . . I ~ . l l k . . . . . . . . * 4 ~ - - a , , _ a - - ~

~ ' ~ ' v J - " . . . . . . . . . . . . . . . . . . v~ . ~, . . . . . . . . ~ - . . . . . . . . . . . . . . . . . . . 8 . . . . .

272

N F P A 8 0 5 ~ N o v e m b e r 2 0 0 0 R O C

~!9.E Pooitive mechanica l meanz zha[! hc provided to !e.ck out

sp,ae~

--3-11 Passive Fire Protection Features. This section shall be used to de te rmine the design and installation requ i rements for passive protect ion features. Passive fire protect ion features include wall, ceiling, and floor assemblies, fire doors, fire dampers , and th rough fire barrier penet ra t ion seals. Passive fire protect ion features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical c o m p o n e n t s and e q u i p m e n t f rom the effects of fire.

3-11.1 Building Separation. Each major bui lding within the power block shall be separated from the o thers by barriers having a designated fire resistance rat ing o f 3 hours or by open space of at least 50 ft or meets the requ i rements of NFPA 80A.

Ezc,~tlc=,,: !~,q:c'r,c a pc','fcr~ancc b~sed ar:a!y~i~ dc"~r:~i:;c~ the . . . . . . " l " " ~Y " J " '®~ . . . . . . 6 ~ ' t " . . . . . . . . . . . . . , . . . . . . . 1 . . . . . . . . . . . . . . . . j . . . . 1 . . . . . . . . . . . . . .

(Basis for deletion, NFPA 80A provides for an evaluation m e t h o d of de t e rmin ing separation. This except ion is unnecessary. )

3-11.2 Fire Barriers. Fire barriers required by Chapte r 4 shall include a specific fire-resistance rating. Fire barriers shall be des igned and installed to mee t the specific fire-resistance rat ing us ing assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Test~ ~?f Fire Endurance of Building Construction and Materials, and ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Mate.,rials.

3-11.3 ' Fire Barrier Penetrat ions. Penetra t ions in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire-resistance rating consis tent with the des ignated fire-resistance rating o f the barrier as de t e rmined by the pe r fo rmance requ i rements established in Chapte r 4. (See 3- 11.3.1 for penetration seals for through penetration fire stops.) Passive fire protect ion devices such as doors and damper s shall conform with the following NFPA standards, as applicable:

(1) NFPA 80, Standard for Fire Doors and Windows (2) NFPA 90A, Standard for the Installation of Air-Conditioning and

Ventilating Systems (3) NFPA 101, Life Safety Code Exception: Where fire area boundaries are not wall-to-wall, floor,to-

ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall be required to assess the adequacy of fire barrier forming the fire boundary to determine i f the barrier will withstand the fire effecL~ of the hazards in the mea. Openings in fire barriers shall be permitted to be protected by other nteans as acceptable to the AItJ.

3-11.3.1" T h r o u g h pene t ra t ion fire stops for penet ra t ions such as pipes, condui ts , bus ducts, cables, wires, pneumat i c tubes and ducts, and similar bui ld ing service e q u i p m e n t that pass th rough fire barriers shall be protected as follows:

(1) The annu la r space between the penet ra t ing item and the th rough open ing in the fire barrier shall be filled with a qualified fire-resistive pene t ra t ion seal assembly capable o f main ta in ing the fire resistance o f the fire barrier. T he assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.

(2) Condui ts shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier t h rough open ing fire stop and can be installed on e i ther side of the barrier iv a location that is as close to the barrier as possible.

Exception: Openings inside conduit 4 in. or- les~ in diameter shall be sealed at the fire barrier with a fire-rated internal seal unless the conduit extends greater than 5 f i on each side of the fire barrier. In this case the conduit opening shall be provided with noncombt~stible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. shall constitute an acceptable smoke and/tot gas seal in this applir-ation.

3-11.4" Electrical Raceway Fire Barrier Systems (ERFBS). ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area. ERFBS shall be tested in accordance with and shall mee t the acceptance criteria o f NRC Generic Let ter 86-10, Supp lemen t I, "Fire Endurance Test Acceptance Criteria For Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area." The ERFBS needs to adequately address the design requ i rements and limitations of suppor ts and in tervening i tems and their impact on the fire barrier

system rating. The fire barrier system's ability to mainta in the required nuclear safety circuits free of fire damage for a specific thermal exposure , barrier design, raceway size and type, cable size, fill, and type shall be demons t ra ted .

Exception No. l: When the temperature,~ inside the.fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, Supplement 1, fnnctionality of the cable, at these elevated temperatures shaU be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement l, Attachment 1, "Attachment Methods fro" Demonstrating Fnnctionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure. " (Note. NRC Concern) Except ion No. 2: EHFBS system6 emplo'ced prior to the issuance of Generic Letter 86-10, Supplement 1, are acceptable providing that the system saccessfnlly met the limiting end point temperature requirements as specked by the AHJ at the time of acceptance. (Note NRC Concern)

Chapter 3 At~pendix A-3-1 (No change to existing text as oubl ished in ROP) A-3-2.2 The policy d o c u m e n t that defines the m a n a g e m e n t

authori ty and responsibility should be consistent with o ther uppe r tier p lant policy documents . The policy d o c u m e n t should contain the following informat ion:

1) Designate the Senior M a n a g e m e n t Position responsible for fire protect ion with the immedia te authori ty and responsibility for the fire protection program. (_Insert Add- th~ existing text from infe, rma~ic, a ".'n A-3-2.2.1 here.)

2~ Designate the posit ion responsible for the daily adminis t ra t ion and coordinat ion o f the fire protect ion program and its implementa t ion . (Insert ~ existing text from infe.rmat:c,~ in A-3-2.2.2 here.)

3) Define the fire protect ion interfaces with o the r organizat ions and assign responsibilities for the coordinat ion of activities. In addit ion, the various plant positions having authori ty for imp lemen t ing the various parts of the fire p ro tecuon program should be identified. (Insert ~ existing text from :,nfe, rm..atic, n in A-3-2.2.3 here.)

4_2 The appropriate AHJ for the various areas of the fire protect ion program should be identified. ( Insert ~ hffc, rm.atie, n fre, m the existing from A-3-2.2.3 here.)

A-3-2.3 remains u n c h a n g e d A-3-2.$(ia) remains u n c h a n g e d A-3-2.3(b] remains u n c h a n g e d A-3-2.3(c) remains u n c h a n g e d A-3-3.1.1(b] remains u n c h a n g e d A-3-3,1.1(c) remains u n c h a n g e d A-3-3.1.2 (Keen first naragravh as is. Add the following as a new

oara~raoh. ) A combust ib le control p rogram should include controls on the

(a) Wood used within the bower block should be listed pressure- impregnated or coated with a listed f ire-retardant annlicat ion. /Add the existing text f rom A-3-3.1.2(a~ here.) Qribbing t imbers 6 in. bv 6 in. or larger are not reouired to be fire- re tardant treated.

(b) Plastic shee t ing materials used in the nower block should be f ire-retardant woes that have oassed NFPA 701. Standard Methods of Fire Tests for Flame Prot)a~ation of Textiles and Films. large-scale tests. Qr equivaient " -

(c) Waste. debris, scran, hacking materials, or o the r combust ib les shou ld be removed f rom an area immediate ly fol lowin, the comple t ion of work or at the end of the shift. whichever comes first,

(d) Gqmbustible storage or s taging areas should be des ignated and limil;s ~hould be established on the types and quanti t ies of ~mred materials. (Add existin~ A-3-3.1.2(d) here.)

(e) Controls on use and storage of f lammable and combust ib le liouids should be in accordance with NFPA 30. Flammable and Combustible Lianids Code. or o the r avolicable standards. (Add exist ing A-3-3.1.21e) here.)

(f) Controls on the use and ~torage o f f lammable gases should be in accordance with annlicable NFPA standards. (Add exis t ine A-3-3.1.2(f) here.)

A-3-3.1.3.1 remains u n c h a n g e d A-3-3.1.3,4 remains u n c h a n g e d

273


A-3-3,5 The following ~rovides additional information about each of the areas that are re0uired to be addressed bv the

1) Wiring above susoended ceiling should be kent to a minimum. Where installed, electrical wiring should be listed for plenum use. routed in armor metal Jacketed. routed in metallic conduit, or routed in cable trays with solid metal too and bottom covers.

2) Only metal tray and metal conduits should be used for electrical raceways. Thin wall metallic tubing should not be used for power instrumentation or control cables. Flexible metallic conduits sho~tld only be used in short lengths to connect components .

3) Electric; cable construction should comply with a flame propagation test as ~/cceutable to the AHI. l~lectric cable insulation should be of a true that has been tested using a ret;qgnized flame soread test. An examnle of such a test is IEEE 817, Standard Test Procedure for Flame-Retardant Coatings Applie0 to Insulated Cables in Cable Trays. and IEEE 1202. Standard for Flame Testing of Cables for Use in Cable Travs and Industrial and Commercial Occuoancies.

A-3-:~,7 Th¢ following orovides additional information about Cinch of the areas that are reouired to be addressed bv the

1) Storage of flammable gas should be located outdoors, or in senarate detached buildings, so that a fire or exnlosion will not a~lverselv imoact systems, eouioment, or comoonents imnortant to nuclear safe~v. NFPA 50A. Standard for Gaseous tt'~dro~en S~stems at Consumer Sites. should be followed for hvdrogen storage.

~) Outdoor high-nressure flammable gas storage containers ~hquld be located so that the long axis is not ooi"nted at buildings..

3) Flammable gas storage cylinders not reouired for normal operation should be isolated i'rom the svstem.

A-3-3.9 remains unchanged A-3-3.10 remains unchanged A-~3.12 The following items need to be considered for the

Reactor Coolant Pumo Lube oil collection system (a) The oil collection svstem for each reactor coolant numn

should be cat)able of collecting lubricating oil from all notential pressurized and nonnressurized leakage sites in each reactor coolant oumo oil svstem.

(b) Leakage should be collected and drained to a vented closed Container that can hold the inventorv of the reactor coolant oumo lubricating oil system.

(c) A flame arrestor is reouired in the vent if the flash noint chi~racteristics of the oil oresents the hazard of a fire flashback.

(d) Leakage points on a reactor coolant tmmo motor to be orotected should include, but not be limited to. the lift oumo and piping, overflow lines, oil cooler, oil fill and drain lines and olugs. flanged connections on oil lines, and the oil reservoirs, where such fei~t~.;res exist on the reactor coolant Dumos.

(e) The collection basin drain line to the collection tank should be: large enough to accommodate the largest ootential oil leak ~ueh that oil leakage does not overflow the basin.

(Insert the existing words from A-3-3.12 here) The intent of the reactor coolant numn lube oil collection svstem

is not to confine oil inisting adjacent to the RCP motor. It is recognized that oil misting and subseouent condensation of oil is an inherent characteristic of large motors of this true.

A-3~LI An effective fire brigac/e organization neects to address each of the following issues:

( I t Based on the organization of the fire brigade a determination of what NFPA standard is aoulicable. Consider the following NFPA standards:

(at NFPA 600. Standard on Industrial Fire Brigades (interior structural fire fighting)

(b) NFPA 1500. Standard on Fire Detmrtnurnt Occubational Safety and llealth Programs

(¢) NFPA 1582. Standard on Medical Require, roents f~r Fire

(2) Plant nrocedures and nractice should include measures to ensure that fire brigade members have no other assigned normal plant duties that would orevent immediate response to a fire or other emergency as reouired. Immediate resnonse is considered

tO be achieved if nominal actions are taken to nut associated eauinmerit in a safe condition.

(3)" It is critical that the fire brigade understand the uotential effects that fire fighting activities could have on nlant operation. During every shift, the brigade leader and at least two brigade members should have sufficient training and knowledge of nuclear safetv systems to understand the effects o f fire and f iresuuoressants on r~uciear safety performance criteria. The t3~¢ of an operations advisor, dedicated to fire brigade suonort is an alternative. This advisor needs to oarticiuate in this role during fire brigade training and drills to be effective.

(4) (Insert existing A-34.1(d) test heret (~5) Successful completion of an annual physical examination for

¢~ch fire brigade member is critical oart of the fire brigade nrogram, the annual ohvsical examination should determine if the

v . .

fire briga~t¢ member can nerform the strenuo~l~ activity required during manual fire-fighting ooerations. The uhvsical examination ~hquld also be used to determine the ability of each member to ~1~¢ resoiratorv orotection eau ioment . The NFPA standards referenced in "A~-3-4.1 can be used in the develonment of an aooronriate uhvsical examination.

A-3-4,2 (Keen existing text here. Add the following text; (1) The nlans should detail the fire area configuration and fire

hazards to be encountered in the fire area. along with anv nuclear safety comt)onents and fire nrotection systems and feal;~.Ires that are nresent. (Insert the text from the existing A-3-4.2.1 here)

(2) Pre-fire nlans should be reviewed and uDdated as necessatpy, (~onsideration'of plant modifications on the nre-lalans ~hould be included as oart of the configuration management orocess within the ulant design change nrocess.

(3) Pre-fire olans should be available in the control room and made available to the nlant fire brigade. (Insert text from the existing A-3-4.2.3 here)

(4) Pre-fire nlans should address coordination with other olant grouos during fire emergencies. (Insert the text from the existing A-3-4".2.4 heret -

A-3-4.3(ct Fire brigade drills should be develooed tO test and challenge fire brigade resnonse, including brigade performance a, a team. nrooer use of equipment, effective use of nre-fire plans, and coordination with other grouos. These drills should evaluate the fire brigade's abilities to react, resound, and demonstrate nrooer fire-fighting technioues to control and extinguish the fire and smoke conditions being simulated bv the drill scenario. Fire brigade drills should be conducted in variou~ plant arenas. esoeciallv in those areas identified to be essential to nlant oneration and to contain significant fire hazards. Drill~ should focus on those areas that are identified as the most risk significant,

Acceutable fire brigade drills should be orovided using realistic nlant conditions to maintain fire brigade proficiency. Fire brigade drills should include the following: [Insert all existing information from A-3-4.3(ct here starting with (at]

Drill records should be maintained detailing the drill ,cenariO, fire brigade member resoonse and oerformance• and ability of the fire brigade to oerform as a team. These records should inclu(te documentat ion of the critioue held after each drill.

A-3-4.5.2 remains unchanged A-3-4.5.3 remains unchanged A-3-4.4 remains unchanged A-3-5.2 remains unchanged A-3-5.3 remains I, lnchangeO

(1) Means should be orovided to isolate oortions of the vard fire main loon for maintenance or repair without simultaneouslv shutting off the supply to both fixed fire suppression systems arid fire hose stations nrovided for manual backun. The arrangement of the connections of sorinkler svstems and manual hose station standnines to the nlant fire protection water main should be designed so that a single active failure or a crack to the water sunDlv nioing to these systems can be isolated so as not to imt~air both the nrimarv and backuo fire sunoression svstems. (ROP Log #__22m

(2) Threads comoatible with those used bv local fire debartments should be orovided on all hydrants, hose couplings, and standnine risers. An accentable, alternative is to orovide the

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. . . . . . . . . ~ . . . . . . . . . . . . ~ . . . . . . . . . . ~ ~ t ~ ,


local fire departments with adaoters that allow interconn¢ction between nlant eouinment and the fire denar tment e~3uioment if adequate training and orocedures are provided,

(3) Headers fed from each end are oermitted inside buildings tO supply both sprinkler and standpipe systems, provided steel dining and fittings meeting the reuuirements of ANSI B31.1. Code lot Power Pihing. are used for the headers (uo to and including the first valve} suonlving the sorinkler systems where such headers are Dart of the seismically analyzed hose standpipe system. Where nrovided, such headers are considered an extension of the yard main system.

(4) Each sorinkler and standoioe svstem should be eouiooed with an outside screw and voke (OS&Y/ gate vah,e or other aDoroved shutoff valve.

(5) All fire protection water supply and fire suppression svstem control valves should be under a nefiodic insoection orogram and sunervised by one of I~h¢ following methods;

(a) Electrical sunervision with audible and visual signals in the main control room Or ol~her suitable constantly attended location•

(b) Locking valves in their normal position with access to keys limited to authorized personnel.

(c) Sealing valves in their normal positions, This omion is limited to situations where valves are located within fenced areas or under the direct control of the owner /ooera tor .

(Insert exisljng A-3-5.14 test here} (6) Hydrants should be installed aot3roximatel'v every 250 ft (76

m/ apart on the yard main system. A hOSe house equipped with hose and combination nozzle and other auxiliary eQuinment soecified in NFPA 24. Standard h*r the Installation of Private Fire Service Mains and Their Ai?#urtena*tres. should be orovided at intervals of not more than 1000 ft (305 m} along the yard main system. An acceptable alternative is the use of mobile m e a n s o f t3roviding hose and a~ssociated equipment• such as hose carts or trucks, in lieu of hose houses. If mobile means is provided, the uuantitv of hose should be eauivalent to the eouioment sunolied by three hose houses.

(7) It is important for the availability and reliability of the fire t3rotection water supply system dedicated for fire protection use only. (Insert existing A-3-5.16 test here) However, where the fire nrotection wal:er supply system is used tO provide backuo to nuclear safety systems an analysis should be performed to determine if the fire protection water supply System can deliver the combined fire and nuclear safety flow demands for the duration soecified by the annlicable analysis. If the analysis finds that both demands cannot be satisfied simultaneously, this should be addressed in the aDoronriate Dlaut orocedure.

Fire protection water storage can be used by plant systems serving o0mr functions provided the storage has a dedicated capacity caoable of providing the maximum fire protection demand for tile specified duration as determined in this section.

A-3-6.1 The standniDe and hose systems installed in dower block structures should be Class III as defined in NFPA 14.

1) The standoioe and hose systems needs to address tile flow rate and nozzle oressure at each standoiue to nrovide adeuuate pressure and flow for an effective hose stream given the hose length and nozzle type. Nuclear .D°wer plants often have high capacity fire DumPs and below grade deviations, so hose station oressure reducers may be necessmv for the safety of olant fire brigade members and off-site fire deoar tment nersonnel.

2) The Droner tvoe of hose nozzle to be sl,nolied to each nower block area needs to be considered on a case by case basis considering both special hazards and fire hazards in the area. The usual combination snrav/straight stream nozzle should not be used in areas where the straight stream can cause unaccentable damage or present an electrical hazard to fire-fighting personneh Listed electrically safe fixed fog nozzles are necessary at locations where high-voltage shock hazards exist. All hose nozzles should have shutoff capability and be able to control water flow from fllll open to full closed.

3) Provisions to suoolv water at least to standnioes and hose stations for manual fire suooression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSEI should be need-to be-addressed. Where SUODIv to the standoioe and hose stations

cannot be made directly from the fire protection water supply system other provisions to restore a water supply and distribution svstem for manual fire fighting ournoses should be addressed. This orovisional manual fire-fighting standpipe-hose station system would be caoable of providing fire fighting protection tO the various plant locations imnortant to support ing and maintaining the nuclear safety fnnction. The provisions for establishing this provisional svstem need to be pre-planned and be caoable of being imnlemented in a timely manner following a SSE in order to be

• v

effective. Where the seismic reouired hose stations are CROSS- connected to essential seismic non-fire protection water supply svstems, the fire flow must not degrade the essential water system reouirement.

A-3-9.1 remains unchanged. A-3-10.g Gaseous based fire sunnression system installed in

Dower block areas have specific issues that need to be addressed by the design and during installation. Further information regarding the issues identified bv the text are orovided below. This is not intended to be an all-inclusive list and the design and installation should carefully consider all potential impacts of a gaseous svstem on plant operation, and of olant ooeration on the effectiveness of a gaseous system. v

1) The gaseous sunnression system design a,n~ interface with the ventilation system needs to take into account prevention of over- oressurization during agent iniection, ventilation isolation• adequate sealing to prevent loss of aeent, and confinement of radioactive contaminants.

2) In any area reouired to be orotected bv both primary, and backuo gaseous fire suppression systems ("backup herein refers to a C O ; hose reel and not the primary and alternat, e banks of agent bottles that orovide a second system injection), a single active failure or a crack in any oit~e in the fire sunDression system should not impair both the Drimalw and backun fire suppression

3) Provisions for locally disarmii'~g automatic ga,se0us suppression systems needs to I?¢ sec,ared and under strict administrative control to prevent unauthorized disabling of the

4) Total flooding carbon dioxide systems should not be used in normally occuoied areas. (Insert existing text from A-3-10.6 here,)

5) Automatic total flooding carbon dioxide Systems should be eouiDned with an audible nre-discharge alarm and discharge delav su'fficient to permit egress of oersonnel as required by NFPA 12, The carbon dioxide system should be provided with all odorizer.

fi) Positive mechanical means should be provided to lock Ot~t total flooding carbon dioxide svstems during work in the nrotected

7) The possibility of Sccondaly thermal shock (cooling) damage should be considered during the design of any gaseous fire suppression system bt.lt particularly with carbon dioxide.

8) Particular attention should be given to corrosive characteristics of agent decomposition nroducts on safety systems.

A-3-11.3 remains vnehanged A-3-11.3.1 remains unchanged A-3-11.4 remains unchanged

SUBSTANTIATION: While the intent of the committee was to have Chapter 3 address the fundamental portions of the fire protection program at a nuclear power plant, it was done in a detailed, prescriptive fashion. The revised text in this comment addresses the fimdamental aspects of the fire protection program, but provides the opportunity for additional flexibility to address these flmdamental issues. COMMITTEE ACTION: Accept in Principle in Part.

Revise text to read as follows: Chapter 3 Fundamental Fire Protection Program and Design

Elements 3-1" General. This chapter contains the fundamental elements

of the fire protection program and specifies the minimum design requirements for fire protection systems and featvres. These fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard. Previously approved alternatives from the fimdamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein•

3-2 Fire Protection Plan. 3-2.1 Intent. A site-wide fire protection plan shall be established.

This plan shall document management policy and program

275


direction and shall def ine the responsibilit ies o f those individuals responsible for the plan 's implementa t ion . This section establishes the criteria for an integrated combinat ion of component.5, procedures , and personne l to i m p l e m e n t all fire protect ion program activities.

3-2.2* Managemen t Policy Direction and Responsibili ty. A policy d o c u m e n t shall be prepared that defines m a n a g e m e n t authori ty and responsibilities and establishes the general policy for the site fire protect ion program. The policy shall address the following:

a) Managemen t authori ty and responsibility b) Program adminis t ra t ion and control c) Interface with o the r organizat ions d) Applicable Authori ty Having Jurisdict ion (AHJ) for p rogram

e lemen t s 3-2.3* Procedures . Procedures shall be establ ished for

implementa t ion of the fire protect ion program. In addit ion to procedures that migh t be required by o ther sections of the s tandard, the p rocedures to accompl ish the following shall be established:

(a)* Inspection, testing, and ma i n t enance for fire protect ion systems and features credi ted by the fire protect ion program

(b)* Compensa tory actions i mp l emen t ed when fire protect ion systems and o ther systems credi ted by the fire protect ion program and this s tandard canno t per form their in tended funct ion and limits on impa i rmen t dura t ion

(c)* Reviews of fire protect ion p rog ram- re l a t ed per fo rmance and t rends

(d) Reviews of physical plant modif icat ions and procedure changes for impact on the fire protect ion program

(e) Long-term ma in t enance and conf igura t ion of the fire protect ion program

(f) Emergency response procedures for the plant fire brigade (ROP Log #19)

3-3 Prevention. A fire prevent ion program with the goal of prevent ing a fire from star t ing shall be established, documen ted , and imp lemen ted as part of the fire protect ion program. The two basic componen t s of the fire prevent ion program shall be 1) prevent ion of fires and fire spread by controls on operat ions activities, and 2) design controls that restrict the use o f combust ib le materials. T he design control r equ i rements listed in the r ema inde r o f this section shall be provided as described.

3-3.1 Fire Prevention for Operational Activities. The fire prevent ion program activities shall consist o f the necessary e lements to address the control o f ignit ion sources and the use of t ransient combust ib le materials du r i ng all aspects o f p lant operations. The fire prevent ion program shall focus on the h u m a n and programmat ic e lements necessary to prevent fires from start ing or, should a fire start, to keep the fire as small as possible.

3-3.1.1 General Fire Prevention Activities. The fire prevention activities shall include but not be l imited to the following program elements :

(a) Tra in ing on fire safety informat ion for all employees and contractors including, as a m i n i m u m , familiarization with plant fire prevent ion procedures , fire report ing, and plant emergency a la rms

(b)* D o c u m e n t e d plant inspect ions inc luding provisions for corrective actions for condi t ions where unanalyze:l fire hazards are identified

(c)* Administrative controls address ing the review of p lant modif icat ions and m a i n t enance to ensure that both fire hazards and the impact on plant fire protect ion systems and features is minimized

3-3.1.2" Control of Combustible Materials. Procedures for the control of genera l housekeep ing practices inc luding the control of t ransient combust ibles and combust ib le storage shall be developed and implemented . These p rocedures shall include bnt not be l imited to the following program elements :

(a)* Wood used within the power block shall be listed pressure- impregna ted or coated with a listed fire re ta rdant application.

Exception: Cribbing timbers 6 in.. by. 6 in. or larger shaU not be required to be .fire retardant treated.

(b) Plastic shee t ing materials used in the power block shall be fire-retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Fil~L large scale tests, or equivalent.

(c) Waste, debris, scrap, packing materials, o r o the r combust ibles shall be removed from an area immediate ly following the comple t ion o f work or at the end o f the shift, whichever comes first.

P'xceptio~t: Cribbing, scoffolding and sheeting in use for subsequent w~rrk shifts does not have to be 7emoved 'until completion of the work.

(d)* Combust ible storage or s taging areas shall be designated and limits shall be established on the types and quanti t ies of stored materials.

(e)* Controls on use and storage of f lammable and combust ible liquids shall be in accordance with NFPA 30, lqammable and Combu.~tible Liquids (;ode, or o the r applicable NFPA standards.

(f)* Controls on use and storage of f lammable gases shall be in accordance with applicable NFPA standards.

3-3.1,3 Control o f Ignition Sources. 3-3.1.3.1' A hot work safety procedure shall be developed,

implemented , and periodically upda ted as necessary in accordance with NFPA 51B, Standard for Fire l'~evention D'al~ng Welding, Cutting; and Other ltot Work, and NFPA 241, Standard fiJr Safeguarding ConsttTtction, Alteration, and Demolition Operations.

3-3.1.3.2 Smoking and o ther possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant. .

3-3.1.3.3 O p e n flames or combus t ion-genera ted smoke shall not be permi t ted for leak or air flow testing.

3-3.1.3.4" Plant administrative procedure shall control the use of portable electrical heaters in the plant. Portable fuel-fired heaters shall not be permit ted in p lant areas conta in ing e q u i p m e n t impor tan t to safety or where there is a potential for radiological releases result ing from a fire.

3-3.2 Structural. Walls, floors, and c o m p o n e n t s required to mainta in structural integrity shall be of noncombus t ib le construct ion, as def ined in NFPA 220, Standard on Types of Building Constraction.

3-3.3 Inter ior Finishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101", Life Safety Code*, requ i rements for Class A materials. Interior floor finishes in corridors and exits shall be in accordance with NFPA 101 requ i rements for Class I or Class II interior floor finishes. Interior floor finishes for the control room shall be in accordance with NFPA 101 requ i rements for Class I interior floor finishes.

Exception: The re are no limits on interior floor finish in areas protected by automat ic sprinklers.

3-3.4 Insulat ion Materials. The rma l insulat ion materials, radiat ion shie ld ing materials, ventilation duc t materials, and soundp roo f ing materials shall be noncombus t ib le or limited combus t ib le .

3-3.5 Electrical. 3-3.5.1 Wir ing above suspended ceiling shall be kept to a

m i n i m u m . Where installed, electrical wiring shall be listed for p l enum use, routed in a rmor metal jacketed, routed in metallic condui t , or routed in cable trays with solid metal top and bot tom c o v e r s .

3-3.5.2 Only metal tray and metal condui ts shall be used for electrical raceways.

3-3..5.3* Electric cable const ruct ion shall comply with a flame propagat ion test as accepted to the AHJ.

Exception: Existing cable installed in areas important to nuclear safet~ that has not passed a flame propagation test acceptable to the authority having jurisdiction shaU be/rrovided with a means to inhibit flame propagation unless a performance-based evaluation demonstrates acceptable risk.

3-3.6 Roofs. Metal roof deck construct ion shall be des igned and installed so the roofing system will no t sustain a self-propagating fire on the unders ide o f the deck when the deck is heated by a fire inside the building. Roof coverings shall be Class A as de t e rmined by tests described in NFPA 256, Standard Method of Fire Tests of Roof Coverings.

3-3.7 Bulk Flammable Gas Storage. Bulk compressed or cryogenic f lammable gas storage shall no t be permit ted inside s t ructures hous ing systems, equ ipmen t , or c o m p o n e n t s impor tan t to nuclear safety.

3-3.7.1 Storage o f f lammable gas shall be located outdoors, or in separate de tached buildings, so that a fire or explosion will not adversely impact systems, equ ipmen t , or c o m p o n e n t s impor tan t to nuclear safety. NFPA 50A, Standard for Gaseon.~ Hydrogen. Systems at Consumer Sites, shall be followed for hydrogen storage.

3-3.7.2 Ou tdoo r high-pressure f lammable gas storage conta iners shall be located so that the long axis is no t pointed at buildings.

3-3.7.3 Flammable gas storage cylinders not required for normal opera t ion shall be isolated from the system.

3-3,8 Bulk Storage o f Flammable and Combustible Liquids. Bulk storage o f f lammable and combust ib le liquids shall not be permi t ted inside s t ructures con ta in ing systems, equ ipmen t , or c o m p o n e n t s impor tan t to nuc lear safety. As a m i n i m u m , storage and use shall comply with NFPA 30, Flammable and CombuMible Liquids Code.

3-3.9* Transformers . Where provided, t ransformer oil collection basins and drain paths shall be periodically inspected to ensure

276


that they are free of debris and capable of performing their design function.

3-3.10" Hot Pipes and Surfaces. Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact with hot pipes and surfaces, incluthng insulated pipes and surfaces. Administrative controls shall require the p rompt cleanup of oil on insulation.

3-3.11 Electrical Equipment. Adequate clearance, free of transient combustible material, shall be maintained around energized electrical equipment .

3-3.12" Reactor Coolant Pumps. For facilities with nou-inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal condit ions such as accident condit ions or earthquakes. All o f the following shall apply:

(a) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil system.

(b) Leakage shall be collected and drained to a vented closed container that can hold the inventory of the reactor coolant pump lubricating oil sys tem.

(c) A flame arrestor is required in the vent if the flash point characteristics o f the oil presents the hazard of a fire flashback.

(d) Leakage points on a reactor coolant pump motor to be protected shall include, but not be limited to, the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connect ions on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps.

(e) The collection basin drain line to the collection tank shall be large enough to accommodate the largest potential oil leak such that oil leakage does not overflow the basin.

3-4 Industrial Fire Brigade. 3-4.1" On-Site Fire-Fighting Capability. All of the following

requirements shall apply: (a) A fully staffed, trained, and equipped fire-fighting force shall

be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable:

(1) NFPA 600, Standard on lndo.~trial Fire Brigades ( interior structural fire fighting)

(2) NFPA 1500, Standard on Fire Department Occupational Safety and tlealth Programs

(3) NFPA 1582, Standard on Medical Requirements for Fire Fighters

(b)* Fire brigade members shall have no o ther assigned normal plant duties that would prevent immediate response to a fire or o ther emergency as required.

(c) During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects o f fire and fire suppressants on nuclear safety performance criteria.

Exception to (c): Sufficie, nt training and knowledge shall be permitted to be provided by an operations advisor dedicated to fire brigade support.

(d)* The fire brigade shall be notified immediately upon verification of a fire.

(e) Each fire brigade member shall pass an annual physical examination to determine that h e / s h e can perform the strenuous activity required during manual fire-fighting operations. The physical examination shall de termine the ability of each member to use respiratory protection equipment .

3-4.2* Pre-fire Plans. Current and detailed pre-fire plans shall be available to the fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1-5. (ROP Log #187)

3-4.2.1" The plans shall detail the fire area configuration and fire hazards to. be encountered in the fire area, along with any nuclear safety components and fire protection systems and features that are present.

3-4.2.2 Pre-fire plans shall be reviewed and updated as necessary. 3-4.2.3* Pre-fire plans shall be available in the control room and

made available to the plant fire brigade. 3-4.2.4* Pre-fire plans shall address coordinat ion with o ther

plant groups during fire emergencies. 3-4.3 Training and Drills. Fire brigade members and other plant

personnel who would respond to a fire in conjunct ion with the brigade shall be provided with training commensura te with their emergency responsibilities.

(a) Plant Fire Brigade Training. All of the following requirements shall apply:

(1) Plant fire brigade members shall receive training consistent with the requirements contained in NFPA 600, Standard oft Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate.

(2) Fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire.

(3) A written program shall detail the fire brigade training program.

(4) Written records that include, but are not limited to, initial fire brigade classroom and hands-on training, refresher training, special training schools a t tended, drill a t tendance records, and leadership training for fire brigades shall be maintained for each fire brigade member .

(b) Training for Non-Fire Brigade Pev:~onnel. Plant personnel that respond with the fire brigade shall be trained as to their responsibilities, potential hazards to be encountered , and interfacing with the fire brigade.

(c)* Drills. All of the following requirements shall apply: (1) Drills shall be conducted quarterly for each shift to test the

response capability of the fire brigade. (2) Fire brigade drills shall be developed to test and challenge

fire brigade response, including brigade performance as a team, proper use of equipment , effective use of pre-fire plans, and coordinat ion witu o ther groups. These drills shall evaluate the fire brigade's abilities to react, respond, and demonstra te p roper fire- fighting techniques to control and extinguish the fire and smoke condit ions being simulated by the drill scenario.

(3) Fire brigade dills shall be conducted in various plant areas, especially in those areas identified to be essential to plant operat ion and to contain significant fire hazards.

(4) Drill records shall be maintained detailing the drill scenario, fire brigade member response, and ability of the fire brigade to perform as a team.

(5) A critique shall be held and documented after each drill. 3-4.4 Fire-Fighting Equipment. Protective clothing, respira to~

protective eqmpment , radiation moni tor ing equipment , personal dosimeters, and fire suppression equ ipment such as hoses, nozzles, fire extinguishers, and o ther needed equipment shall he provided for the fire brigade. This equipment shall conform with the applicable NFPA standards.

3-4.5 Off-Site Fire Department Interface. 3-4.5.1 Mutual Aid Agreement. Off-site fire authorities shall be

offered a .plan for their interface during fires and related emergencies on site.

3-4.5.2* Site-Specific Training. Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited 1o participate in a drill at least annually.

3-4.5.3* Security and Radiation Protection. Plant security and radiation protection plans shall address off-site fire authority response.

3-4.6* Communications. An effective emergency communicat ions capability shall be provided for the fire brigade.

(ROP Log #127aa) 3-5 Water Supply. 3-5.1 Two separate and reliable fire protection water supplies of

adequate quantity, and duration shall he provided by one of the two following methods:

(a) Providing a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies.

(b) Calculate the fire flow rate for two hours. This fire flow rate shall be based on 500 ga l /min (1892.5 L/ra in) for manual hose streams plus the largest design demand of an}, sprinkler or fixed water spray system(s) in the power block as de te rmined in accordance with NFPA 13, Standard for the hlstallation ~/'Spri~lk~ Systems, or NFPA 15, Standard .for Water Spray Fixed Systems ./or Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service. (ROP Log #70)

3-5.2* The tanks shall be interconnected such that fire pninps can take suction from either or both. A faihu-e in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard [m- Water Tanks for Private Fire Protectzon.

Exception No. 1: Water storage tanks shall not be mqutred when .fire pomps are able to take suction.firm a large body ~f water (.~uch a.~ a lake), provided each fire. pump has its own suction and both sm'tion.~ and pumps are adequately separated.

Fxeeption No. 2: Cooling tower ba.~ins are an acceptable wa/el source firr fire pumps whey, the volunw i.~ s,~fficient for both pn,poses and water quality is consistent with the demands of the .fire .w~vice.

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3-5.3* Fire pumps , des igned and installed in accordance with NFPA 20, Standard for the Installati¢m of Stationmy Pumps for lb~re Protection, shall be provided to ensure that 100 percent o f the required flow rate and pressure are available a s suming failure of the largest p u m p or p u m p power source.

3-5.4 Each p u m p and its driver and controls shall be separated from the remain ing fire p u m p s and from the rest of the plant as required to ensure that 100 percent o f the required flow rate and pressure are available.

3-5.5 Fire p u m p s shall be provided with automat ic start and manua l stop only.

3-5.6 Individual fire p u m p connec t ions to the yard fire main loop shall be provided and separated with sectionalizing valves between connec t ions .

3-5.7 An u n d e r g r o u n d yard fire main loop, des igned and installed in accordance with NFPA 24, Standard for the ln,~tallation of PTivate Fire Service Mains and their Appu, rtenances, shall be installed to furnish ant icipated water requi rements .

3-5.8 Means shall be provided to isolate port ions of the yard fire main loop for ma in t enance or repair wi thout s imul taneously shut t ing off the supply to both fixed fire suppress ion systems and fire hose stations provided for manua l backup. Sprinkler systems and manua l hose station s tandpipes shall be connec t ed to the plant fire protect ion water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppress ion systems. (ROP Log #276)

3-5.9 Threads compat ible with those used by local fire depa r tmen t s shall be provided on all hydrants , hose couplings, and s tandpipe risers.

Exception: Fire departments shall be permitted to be provided with adapters that allow interconnection between plant equipment and the fire department equipmeztt i f adequate training and procedures are provided.

3-5.10 Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equ ipped with hose and combina t ion nozzle and o the r auxiliary e q u i p m e n t specified in NFPA 24, Standard.for the Installation of Private Fire Selvice Mains and Their Appurtenances, shall be provided at intervals o f not more than 1000 ft (305 m) a long the yard main system.

Exception: Mobile. means of providing hose and associated equipment, such as hose carts or trucks, shall be permitted in lieu of hose house, s. Where provided, such mobile equipment shall be equivalertt to the equipment supplied by three haw hmtses.

3-5.11" The fire protect ion water supply system shall be dedicated for fire protect ion use only.

Excq~tion No. 1: Fire protection water supply systems shall be permitted to be 'used to p~vvide backup to nttclear safety systems provided the fire protection water supply systems are designed and maintained to deliver the combined fi~e and nuclear safe(~ flow demands for the duration spec¢ied by the applicable analysis.

Exception No. 2: Fire protection water storage can, be provided by Dlant system~ sert,ing other functions provided the storage has a dedicatect capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.

3-6 Standpipe and Hose Stations. 3-6.1" Power block buildings shall be provided with s tandpipes

and hose systems installed in accordance with NFPA 14, Standard for the lnstaUatio,n of Standpipe and Hose ,Systems.

3-6.2 A capability shall be provided to assure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision o f hose station pressure reducers where necessary for the safety of plant fire brigade m e m b e r s and off-site fire depa r tmen t personnel .

3-6.3 Where the seismic required hose stations are cross- connec ted to essential seismic non-fire protect ion water supply systems, the fire flow shall no t degrade the essential water system requi rement .

3-6.4 Provisions shall be made to supply water at least to s tandpipes and hose stations for manua l fire suppress ion in all areas conta in ing systems and c o m p o n e n t s needed to per form the nuclear safety funct ions in the event o f a safe shutdown ear thquake (SSE).

3-4 Fire Extinguishers . Where provided, fire ext inguishers of the appropr ia te number , size, and type shall be installed in accordance with NFPA 10, Slandard for Portable Fire Extin~uishers. Extinguishers shall be permi t ted to be posi t ioned outside of fire areas due to radiological condi t ions .

3-8 Fire Alarm and Detection Systems. 3-8.1 Fire Alarm. Alarm initiating devices shall be installed in

accordance with NFPA 72, National Fire Alarm Code ®. Alarm annunc ia t ion shall allow the proprietary alarm system to t ransmit fire-related alarms, supervisory signals, and t rouble signals to the

control room or o the r constant ly a t t ended location from which required notifications and response can be initiated. Personne l assigned to the proprietary alarm station shall be permit ted to have o ther duties. The following fire-related signals shall be t ransmit ted:

(1) Actuat ion o f any fire detect ion device (2) Actuat ion o f any fixed fire suppress ion system (3) Actuat ion of any manua l fire alarm station (4) Starting o f any fire p u m p (5) Actuat ion o f any fire protect ion supervisory device (6) Indicat ion of alarm system trouble condi t ion (ROP P319(a) Committee Proposal deleted 3-6.1.2) 3-8.1.1 Means shall be provided to allow a person observing a

fire at any location in the plant to quickly and reliably c o m m u n i c a t e to the control room or o the r suitable constantly a t t ended location.

3-8.1.2 Means shall be provided to promptly notify the following of any fire emergency in such as way as to allow them to de te rmine an appropr ia te course o f action:

(1) General site popula t ion in all occupied areas. (2) Members of the fire brigade and o the r groups suppor t ing

fire emergency response. (3) Off-site fire emergency response agencies. Two in d ep en d en t

means shall be available (e.g., t e lephone and radio) for notification o f off-site emergency services.

3-8.2 Detection. If automat ic fire detect ion is required to meet the pe r fo rmance or determinist ic r equ i rements o f Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes .

(ROP Log #296 and #74 deleted 3-6.2.2.1) (ROP Log #127ad deleted 3-6.2.2.2) 3-9 Automatic and Manual Water-Based Fire Suppression

Systems. 3-9.1" If an automat ic or manua l water-based fire suppress ion

system is required to meet the pe r fo rmance or determinist ic requ i rements of Chapte r 4, then the system shall be installed in accordance with the appropr ia te NFPA s tandards including:

(1) NFPA 13, Standard for the Installation qf Sprinkler Systems (2) NFPA 15, Standard,for Water Spray Fixed Systems for Fire

Protection (3) NFPA 750, Standard on Water Mist Fire Protection System~ (4) NFPA 16, Standard for the InstalLation of Foam-Water Sprinkler

and Foam-Water Spray ,Systems 3-9.2 Each system shall be equ ipped with a water flow alarm. 3-9.3 All a larms from fire suppress ion systems shall annunc ia te

in the control room or o the r suitable constantly a t t ended location. 3-9.4 Diesel~triven fire pumps shall be protected by automatic

sprinklers. 3-10 Gaseous Fire Suppression Systems. 3-10.1" If an automat ic total f looding and local application

gaseous fire suppress ion system is required to meet the pe r fo rmance or determinist ic requi rements o f Chapter 4, then the system shall be des igned and installed in accordance with the following applicable NFPA codes:

(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems (2) NFPA 12A, Standard on ftalon 1301 Fire Extinguishing System~ (3) NFPA 2001, Standard on Ck',an Agent Fire Extinguishing

Systems 3-10.2 Opera t ion of gaseous fire suppress ion systems shall

annunc ia t e and alarm in the control room or o the r constantly a t tended location identified.

3-10.3 Ventilat ion system design shall take into account prevent ion from over-pressurization dur ing agen t injection, adequate sealing to prevent loss o f agent, and con f in em en t of radioactive con taminants .

3-10.4" In any area required to be protected by both primary and backup gaseous fire suppress ion systems, a single active failure or a crack i u a n y pipe in the fire suppress ion system shall not impair both the primary and backup fire suppress ion capability.

3-10.5 Provisions for locally d isarming automat ic gaseous suppress ion systems shall be secured and u n d e r strict administrat ive control.

3-10.6" Total f looding carbon dioxide systems shall no t be used in normally occupied areas. (ROP Log#36)

3-I0.7 Automat ic total f looding carbon dioxide systems shall be equ ipped with an audible pre-discharge alarm and discharge delay sufficient to permi t egress of personnel . The carbon dioxide system shall be provided with an odorizer.

3-10.8 Positive mechanica l means shall be provided to lock out total f looding carbon dioxide systems du r ing work in the protected space.

3-11 Passive Fire Protection Features. This section shall be used to de te rmine the design and installation requ i rements for passive

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protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.

3-11.1 Building Separation. Each major building within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 hours or by open space of at least 50 ft or meets the requirements of NFPA 80A.

3-11.2 Fire Barriers. Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be designed and installed to meet the specific fire-resistance rating using assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of l"~re Endara~tce t?f BuiMing Construction a~d Materials, and ASTM E 119, Standard Tes't Methods .for Fire 7~sts of Building Con~traction and Materials.

3-11.3" Fire Barrier Penetrations. Penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire-resistance rating consistent with the designated fire-resistance rating of the barrier as determined by the performance requirements established in Chapter 4. (.See 3- I 1.3. I for penetration seaL~ for through p~netration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable:

(1) NFPA 80, Standard for Fire Doors and Windowt (2) NFPA 90A, Standard for the Installation of Air-Conditioning and

Ventilating Systems (3) NFPA 101, Life Safet~ Code Exception: Where Jire area bo.nndaries are not wall-to~waU, floor-to-

ceiling boundaries wtth all penetrations sealed to the fire rating required of the boundaries, a pMfirrmance-based analysis shall be required to assess the adeq'aacv of fire barrier forming the fire boundary, to determine if the barrier will withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to be protected by other n~ans a.s acceptable to the AttJ.

3-11.3.1" Through penetration fire stops for penetrations such as pipes, conduits, bus ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall be protected as follows:

(1) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.

(2) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and can be installed on either side of the barrier in a location that is as close to the barrier as possible.

Exception: Opemng~ inside cond'ait 4 in. or less in diameter shall be sealed at the fvre barrier with a fire-rated internal seal unless the conduit extend.~ great(* than 5 .feet on each sidle of the fire barrier. In this case the condnit @eninA~ shall be provided with noncombustible material to prevent the pas.sage ((smoke and hot gases. The fill depth of the material packed to a depth of 2 in. shall constitute an acceptable smoke and hot gas seal in this apphcation.

3-11.4" Electrical Raceway Fire Barrier Systems (ERFBS). ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area. ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10. Supplement 1, "Fire Endurance Test Acceptance Criteria For Fire Barrier Systems Used'to Separate Safe Shutdown Trains Within the Same Fire Area." The ERFBS needs to adequately address the design requirements and limitations of supports and intercening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, harrier design, raceway size and type, cable size, fill, and type shall be demonstrated.

Exception: When the temperatures inside the fire barrier .~),stem exceed the maximum temperature aUowed by the acceptance criteria of Generic Letter 86-10, Supplement 1, functionality of the cable at these elevated temperat'ure.~ shall be demon.strated. Qualification demonstration oJ these cables shall be performed in accordance with the elertrical te.~ting requirements of Gene.l~c l~tter 86-10, Supplement 1, Attachment I, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Bamer 3~,stems Daring and After Fire Endura'nce Test Expos ore. "

Chapter 3 Appendix

A-3-1 Fire protection systems that deviate from applicable NFPA design codes and standards should be supported by an engineering analysis acceptable to the authority having jurisdiction that demonstrates satisfacto~ 7 compliance with the performance objectives. (Orlando Pg. 168a)

A-3-2.2 The policy document that defines the management authority and responsibility should be consistent with other upper tier plant policy documents. The policy document should contain the following information:

a) Designate the Senior Management Position responsible for fire protection with the immediate authority and responsibility for the fire protection program. The senior plant management position responsible for fire protection should be the plant general manager or equivalent position. Fire protection needs the support of the highest level of management. This support is particularly important wbere various fire protection programmatic

' responsibilities go across organizational lines (i.e., operations, system engineering, design engineering, security, training).

b) Designate the position responsible for the daily administration and coordination of the fire protection program and its implementation The individual responsible for the day-today administration of the fire protection program on site should be experienced in nuclear fire protection. Preference should be given to an individual with qualifications consistent with member grade status in the Society of Fire Protection Engineers.

c) Define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In addition, the various plant positions having authority for implementing the various parts of the fire protection program should be identified. Fire protection impacts and is impacted by virtually all aspects of plant operations. These interfaces need to be considered on a plant-by-plant basis. Typically these interfaces include but are not limited to:

Plant operations Security Maintenance System engineering Design engineering Emergency planning Quality assurance Procurement Corporate fire protection (~nsurance) Chemistry Health physics Licensing d) A nuclear power plant is typically under the jurisdiction of

several authorities, which vary depending on the area of the fire protection program being considered. A federal authority will likely be the AHJ for portions of the program addressing nuclear safety, and a building code official at a local level might be responsible for life safety. The insurance coverage provider or the owner/operator can be the AHJ for plant damage/business interruption aspects of the program. To eliminate future confusion it is important to id@ntify the appropriate AHJ for the various aspects of the program in the program or policy document.

A-3-2.3 Most plants have procedure formats and hierarchies for controlling various operations and activities. Fire protection-related procedures should be consistent with other plant procedures to the extent possible.

A-3-2.3(a) Inspection, testing, and maintenance procedures should be developed and the required actions performed in accordance with the appropriate NFPA standards. Some AHJs such as insurers might have additional requirements that should be considered when developing these procedures. Performance- based deviations from established inspection, testing, and maintenance requirements can be granted by the AHJ. Where possible, the procedures for inspection, testing, and maintenance should be consistent with established maintenanc~ procedure format at the plant.

A-3-2.3(b) Compensatory actions might be necessary to mitigate the consequences of fire protection or equipment credited for safe shutdown that is not available to perform its function. Compensatory actions should be appropriate with the level of risk created by the unavailable equipment. The use of compensatory actions needs to be incorporated into a procedure to ensure consistent application. In addition, plant procedures should ensure that compensatory actions are not a substitute for prompt restoration of the impaired system.

A-3-2.3(c) In order to measure the effectiveness of the fire protection program, as well as to collect site-specific data that can be used to support performance and risk-informed considerations, a process to identify performance and trends is needed. Specific performance goals should be selected and performance measured.

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A procedure that establishes how to set goals and how to consistently measure the pe r fo rmance is a critical part o f this process.

A-3-3.1.1(b) Fire prevent ion inspect ions are an impor tan t part o f the overall fire protect ion program. Use o f fire protect ion personnel to perform these inspect ions should be only one part of the inspect ion program. Main tenance and opera t ions supervisors should be trained on fundamen ta l s of fire prevent ion so that they can incorporate fire prevent ion into their field walkdowns. In fact, t raining the general p lant popula t ion to recognize, repor t on and correct correct fire hazards is r e c o m m e n d e d . Not only does this increase the n u m b e r o f people looking for hazards, it also educates the employees to avoid creat ing the hazards in the first place. NFPA 601, Standard for Security Services in Fire Loss Prevention, provides a m e t h o d for developing and i mp l emen t ing a fire p reven0on surveillance plan.

A-3-3.1.1(c) In addit ion to reviews of ma in tenance activities, adequate controls need to be placed in the appropr ia te plant procedures to make sure that fire prevent ion considerat ions are

cluded in the modif icat ion and ma i n t enance process. These considerat ions should include not only informat ion on hot work and combust ib le materials controls, bu t also the impact o f modificat ion and ma i n t enance activities on fire protect ion systems, inc luding blocking sprinklers, detect ion devices, ext inguishers , hose stations, and emergency lights with scaffolding or staged equ ipment . The effect of hot work on detect ion in the area, (smoke or flame) as well as on suppress ion systems should also be considered, as well as the effect on fire barriers due to open doors or b reached barriers.

A-3-3.1.2 Combust ible materials in this section refer to t ransient type combustibles . In-situ combust ib les are addressed as part o f the specific equ ipmen t . Control o f t ransient combust ib le can be accompl ished in a variety of ways. Some plants have used a permi t system. Othe r plants have used procedural controls with oversight by supervision. Controls should not only cons ider quanti t ies o t combust ibles but also the actual location of the t ransient combustible. For example , 1000 lb o f t ransient Class A combust ible materials can be permi t ted and have only a small effect on the equivalent fire severity. However, if this 1000 lb is placed in the vicinity o f critical cables or equ ipmen t , t hen there is a significant impact on the level of risk.

A-3-3.1.2(a) Use o f f ire-retardant paint requires special care. Inconsis tent application and exposure to weather can reduce the effectiveness o f f ire-retardant coatings. Large t imbers are occasionally used to suppor t large pieces of e q u i p m e n t du r ing storage or main tenance . T he size o f these t imbers make them difficult to ignite, and they do not represent an immedia te fire threat.

A-3-3.1.2(d) The limits permi t ted in designated storage areas should be based on the type o f materials being stored, the type, if any, o f fire suppress ion in the area, and separat ion from e q u i p m e n t necessary to mee t the goals def ined in Chapter 1 o f this s tandard. Storage inside a power block building, such as the auxiliary building, turbine building, reactor or c o n t a i n m e n t building, control building, diesel gene ra to r building, or radioactive waste storage or process ing buildings, should be limited to that needed in a short period of time. Typically, one week's worth of supplies is appropriate .

A-3-3.1.2(e) For p lant areas con ta in ing e q u i p m e n t impor tan t to nuclear safety or where there is a potential for radiological release result ing from a fire, addit ional controls over f lammable and combust ib le liquids above those requi red by applicable NEPA s tandards should be considered. Power plants typically use a n u m b e r o f f lammable and combust ib le liquids and gases as part o f the operat ion of the plant. T he type of chemical and the quant i t ies used also change over time. T he administrative control p rocedures should be flexible e n o u g h to handle all types o f gases and liquids.

A-3-3.1.2(f) For plant areas conta in ing e q u i p m e n t impor tan t to nuc lear safety or where there is a potential for radiological release resul t ing from a fire, addit ional controls over f lammable gases above those required by applicable NFPA s tandards should be cons idered .

A-3-3.1.3.1 Hot work controls should include a permi t that is approved by the appropr ia te level o f m a n a g e m e n t prior to the start o f work. Permit dura t ion should be limited to one shift. Tra in ing on the hot work control p rocedure as well as the appropria te level o f hands-on fire ex t inguisher t raining should be provided to all who are assigned hot work responsibilities, including both the persons pe r fo rming the hot work as well as ,the person assigned hot work fire watch responsibilities. The administrative p rocedure should also include instruct ions for hand l ing use and storage o f oxygen and acetylene cylinders used for hot work.

A-3-3.1.3.4 The administrat ive p rocedures should include a m e t h o d to control the use of electric heaters so that only those that have been inspected and approved for use will be used. NFPA 241, Safeguarding Construction, Alteration and Demolition Operations, should be utilized for guidance when considering the use of temporary heating equipment.

A-3-3.5.3 Electric cable cons t ruc t ion should comply with a f lame propagat ion test as acceptable to the AHJ. Electric cable insulat ion should be of a type that has been tested using a s tandard recognized flame spread test. An example o f such a test is IEEE 1202, Standard for Flame Test ing o f Cables for Use in Cable Trays and Industrial and Commerc ia l Occupancies . In addit ion, IEEE 817, Standard Test Procedure for Flame-Retardant Coatings Applied to Insulated Cables in Cable Trays, is an example of mee t ing the f lame propagat ion requi rements .

A-3-3.9 Overflowing oil collection basins have spread fires in some incidents. In addit ion, upon overflow, the oil can go directly to a water source, such as a bay, lake, and so forth, which involves env i ronmenta l concerns . Periodic inspect ions by appropria te personne l is necessary. Also, d ra in ing the oil collection basins following heavy rains should be incorporated into plant p rocedures .

A-3-3.10 There have been a n u m b e r of fires within the industry that have occurred when h igh- tempera tu re lube oil has contacted hot pipes. Ignition has occurred, even t hough there has been no pilot fire source and the auto-ignit ion t empera tu re of the lube oil has been above that o f the pipe. This ignit ion is believed to be caused in part by the distillation o f the oil at the pipe surface after wicking th rough the insulation. T h e l ighter ends that are driven off by the distillation process then ignite since they have a lower auto- ignit ion tempera ture . Immedia te clean-up o f the oil is impor tan t to avoid s u c h fires.

A-3-3.12 Potential pressurized and unpressur ized leakages should be considered in des igning a lube oil collection system. Leakage points that should be evaluated to de te rmine if protect ion is warranted include the lift p u m p and piping, overflow lines, lube oil coolant, oil fill and dra in lines, plugs, f langed connect ions , and lube oil reservoirs where such features exist on the reactor coolant pumps . Lack of protect ion for any potential leakage point should be just if ied by analysis and should be d o c u m e n t e d for review by the AHJ.

A-3-4.1(b) Immedia te response as listed in these sections is considered to be achieved if nomina l act ions are taken to put associated e q u i p m e n t in a safe condi t ion.

A-3-4.1 (d) Verification o f a fire should result in a p ro m p t notification o f the fire brigade. Immedia te d ispatching of the fire brigade should occur upon verbal notification of a fire, two or more fire detectors being activated in a zone, or receipt of a fire suppress ion system flow alarm.

A-3-4.2. As a m i n i m u m , the pre-fire plans should include a descr ipt ion of the following:

(1) Available fire protect ion systems (2) Fire barriers (3) Fire doors (4) Locked doors (5) Inaccessible or l imited access areas (6) Safe shutdown e q u i p m e n t (7) Fire ex t inguisher locations (8) Ventilat ion capabilities (9) C o m m u n i c a t i o n e q u i p m e n t (10) Radiological hazards Special hazards(12) Areas subject to f looding

A-3-4.2.1 Pre-fire plans should detail radiologically hazardous areas and radiat ion protect ion barriers. Methods o f smoke and hea t removal should be identified for all fire areas in the pre-fire plans. These can include the use of dedicated smoke and heat removal systems or use o f the s t ructure ' s heating, ventilating, and air-condit ioning (HVAC) system if it can operate in the 100 percen t exhaus t mode .

Water drainage me thods should be reviewed and included in the pre-fire plan for each area.

Pre-fire plans should also conta in at least minimal informat ion on any hazardous materials located in the fire area, (i.e., acids, caustics, chemicals ) .

A-3-4.2.3 Considerat ion should be given to providing the pre-fire plans to public fire depa r tmen t s that migh t respond to the site so that they can use them in the deve lopment o f their own pre- plans. However, if pre-plans are provided to off-site fire depar tments , be aware that ensur ing that these copies remain cu r r en t can be difficult.

A-3-4.2.4 The pre-plans should cons ider coordinat ion of fire f ight ing and suppor t activities with o the r plant groups. These groups include but are not l imited to radmtion protection,

280


security, and operations. Coordinat ion issues include the following:

(1) Access into normally locked or limited access areas (due to radiological or security concerns

(2) Dosimetry (including dosimetry for the off-site fire departments

(3) Local and remote moni tor ing for radiological concerns (dose, contaminated smoke, contaminated fire-fighting water Funoff)

(4) Scene control by security (5) Escort o f off-site fire depar tment personnel and equipment

to the scene (6) Equipment shutdown by operat ions (electrical components ,

ventilation) A-3-4.3(c) Acceptable fire brigade drills should be provided

using realistic plant conditions to maintain fire brigade proficiency. Fire brigade drills should include the following:

(a) Fire brigade drills are to be a simulated emergency exercise ' involving a credible emergency requiring the fire brigade to perform planned emergency operations. The purpose of these drills is to evaluate the effectiveness of the training and education program and the competence of fire brigade members in performing required duties and functions. Fire brigade drills can be ei ther announced or unannounced to the fire brigade. However, the senior shift representative should be informed of all drills prior to their commencement .

(1) Announced. A fire brigade drill, including the scenario of the drill, that is announced in advance to the fire brigade and other personnel who can be alerted.

(2) Unannounced. A fire brigade drill that is not announced in advance to the fire brigade and other personnel who can be alerted.

(b) Generally, fire tnJgade drills are not considered training evolutions. ttowever, announced drills can incorporate a degree of training while performing an evaluation of the fire brigade. Announced fire brigade drills can vary in type~ of response, speed of response, and use of equipment. Unannounced fire brigade drills are to be used specifically to evaluate the fire fighting readiness of the fire brigade, fire brigade leader, and fire protection systems and equipment.

(c) At least annually, each shift fire brigade should participate in an unannounced fire brigade drill. Unannounced fire brigade drills should be per formed in a realistic manner , using real-time evolutions, full personal protective equ ipment (PPE) including self-contained breathing apparatus (SCBA), and, where appropriate, charged hose lines. Assessment of the following items should be performed:

(1) Fire alarm effectiveness (2) Timeliness of notification of the fire brigade (3) Timeliness of assembly of the fire brigade (4) Selection, placement, and use of equipment , personnel , and

fire-fighting strategies (5) The brigade members ' knowledge of their role in the fire-

fighting strategy (6) The brigade members, knowledge and ability to properly

deploy fire-fighting equipment and proper use of PPE, SCBA, arid communlca tmns equ ipmen t

(7) The brigade members ' conformance with established plant fire-fighting procedures

(8) A critique of the drill per formed by all of the participants, including brigade members, drill planners, and observers.

A-$-4.5.2 Training of the plant fire brigade should be coordinated with the local fire depar tment so that responsibilities and duties are del ineated in advance. This coordinat ion should be part of the training course and should be included in the training of the local fire depar tment staff. Local fire depar tments should be provided training in operational precautions when fighting fires on nuclear power plant sites and should be made aware of the need for radiological protection of personnel and the special hazards associated with a nuclear power plant site.

A-3-4.5.3 Items to be addressed should include overseeing the issuance of security badges, film badges, and dosimetry to the responding public fire-fighting forces and ensuring that the responding off-site fire depar tment(s) are escorted to the designated point of entry to the plant.

A-3-4.4 The fire brigade communica t ion system should not interfere with o ther plant groups such as security and operations. Multichannel portable radios are used for communicat ions at nuclear power plants. This section does not prohibi t sharing of radio channels by various station groups. The use and assigmnent of channels should ensure that the fire brigade, operations, and security all can use the radios to carry out their fimctions during a fire emergency.

The potential impact o f fire on the plant 's communicat ion system should be considered. For example, separation of repeaters from o ther forms of communicat ions to ensure that communicat ion capability will remain following a fire is one such consideration.

In unique or unusual circumstances where equipment cannot be designed to prevent radio frequency interference, the authority having jurisdiction can permit the area around the sensitive equ ipment where portable radios cannot be used to be identified and marked so that fire fighters can readily recognize the condit ion. Training in this recognition also should be provided.

Fire brigade personnel need to be aware of the use of portable radios by the off-site fire depar tments responding within these areas. Off-site fire depar tment radios are typically of a higher wattage output than plant fire brigade radios and can affect plant equ ipment in areas where plant radios would not.

A-3-5.2 Due to the 100 percent redundancy feature of two tanks, refill times in excess of 8-hour are acceptable.

A-3-5.3 For maximum reliability, three fire pumps should be provided so that two pumps meet the maximum demand including hose streams. Two fire pumps can be an acceptable alternative, provided ei ther o f the fire pumps can supply the maximum demand including hose streams within 120 percent of its rated capacity.

A-3-6.1 The standpipe and hose systems needs to address the flow rate and nozzle pressure at each standpipe to provide adequate pressure and flow for an effective hose stream given the hose length and nozzle type. Nuclear power plants often have high capacity fire pumps and below grade deviations, so hose station pressure reducers may be necessary for the safety of plant fire brigade members and off-site fire depar tment personnel .

A-3-9.1 An adequate capability should be provided to drain water from fire suppression systems away from sensitive equipment .

A-3-10.1 The possibility of secondary thmlnal shock (cooling) damage should be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide. Particular at tention should be given to corrosive characteristics of agent decomposi t ion products on safety systems.

A-3-11.3 Openings in fire barriers can be protected by methods such as a combination of water and draft curtains. Such alternate protect ion can be used if justified by the FHA [What does FHA stand for?] and approved by the AHJ.

A-3-10.4 The back-up gaseous suppression system referred to in this section would be a CO 2 hose reel. This back-up system does not refer to the primary and alternate bottle banks on a Halon or CO 2 system.

A-3-10.6 If total flooding carbon dioxide systems are used in rooms that require access by personnel engaged in actions to achieve and maintain safe and stable conditions, provisions within the applicable procedures should ensure that ei ther the room is ventilated prior to entry or the response personnel are provided with self-contained breathing apparatus.

A-3-11.3.1 Various fire test protocols are available to assess the performance of a through penetrat ion fire stop's ability to prevent the propagation of fire to the unexposed side of the assembly. These protocols include ASTM E 814, Standard 7~.st Method for Fire Tests of Through Penetration Fire Stops, IEEE 634, Standard Cable Penetration Fire Stop Qualification Test, and UL-1479, File Tests of Through Penetration Fire Stbps. A-3-11.4 Additional fire test protocols are available to assess the

capability of a barrier system when used to separate redundant safety systems from the effects of fire exposure. Use of these test methods should be addressed with the AHJ. These test methods include ASTM E-1725, Standard Test Methods fo~ Fire Tests of Fire Resistive Barrier Sy,*tems for Electrical Components, and UL-1724, Fire Test for Electrical Circuit Protective .Systems.

The ERFBS should meet o ther design-basis requirements including, seismic position retent ion and ampacity derat ing of electrical cables. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION: COMMITTEE STATEMENT: The Committee incorporated many of the accepted comments into the rewrite of Chapter 3.


EXPLANATION OF NEGATIVE: ALPERT: Please change my vote on this comment to negative.

See comments by Sinopoli and Garrett. I am concerned that the Authority Having Jurisdiction may not agree with my interpretation of "previously approved alternative" as outl ined in Paragraph 3-1 [see my initial affirmative with comment vote on Comment 805-58 (Log #82)]. This would require existing plants to backfit many aspects of their fire protection systems to meet the fundamental

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requirements of Chapter 3 in order to adopt 805. My initial affirmative vote with comment read as follows:

Comment 805-58 (Log #82) Paragraph 3-6.4 states: "Provisions should be made to supply water at least to standpipes and hose stations for manual fire suppression in areas containing systems and componen t s needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE)."

Comment 805-58 (Log #82) Paragraph 3-1 states: "This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. These fire protection program elements and min imum design requirements shall not be subject to the performance-based methods permit ted elsewhere in this standard. Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein."

Branch Technical Position CMEB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," Rev. 2 Section C.6.c(4), and Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976," Section E.3.(d) (for plants whose application was docketed but the construction permit was not received as of 7-1-76) state in part:

"Provisions should be made to supply water at least to standpipes and hose connect ions for manual firefighting in areas containing equipment required for safe plant shutdown in the event o f a safe shutdown earthquake. The piping system serving such hose stations should be analyzed for SSE loading and should be provided with suppor ts to ensure system pressure integrity. The piping and valves for the portion of hose standpipe system affected by this functional requirement should, as a minimum, satisfy ANSI B31.1, Power Piping."

Note: Draft Regulatory Guide DG-1097, "Fire Protection for Operat ing Nuclear Power Plants," Section 3.4.1, does not contain similar seismic requirements for s tandpipe systems.

For plants under construction and for operat ing plants there is no equivalent requirement for seismic protect ion of the s tandpipe systems. In addition, 10CFR50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operat ing Prior to January 1, 1979," has no requirement for the seismic design for the standpipe and hose systems required in Section III.D.

Plants licensed to e i ther Section C.6.c(4) of the BTP CMEB or Section E.3.(d) o f Appendix A to BTP APCSB (no construct ion permit) will have an approved standpipe system that meets the requirements of Comment 805-58 (Log #82), Paragraph 3-6.4 due to the similar requirements in the BTPs and Paragraph 3-6.4 of NFPA 805, or by the provision for previously approved configurations " allowed in Paragraph 3-1 due to an approved exempt ion or deviation from the BTP.

For a plant licensed to the requirements of 10CFR50, Appendix R Section III.D or Section E.3.(d) o f Appendix A to BTP APCSB there was no requi rement for seismic design of the s tandpipe system, however, as a member of the Technical Commit tee it is my unders tanding that the standpipe system would be considered a "previously approved alternative" as stated in Paragraph 3-1 and the facility would meet the fundamental requirements as required in Paragraph 3-1 and no backfit is required. This would also apply to o ther aspects o f the fire protect ion systems and features that did not have a specific requi rement under the plant 's licensing basis, but are r q u i r e d under Chapter 3 of NFPA 805.

Since no backfit would be required for the s tandpipe systems of plants licensed to Appendix R Section III.D or Section E.3.(d) o f Appendix A to BTP APCSB or o ther accepted aspects o f the fire protection systems and features that did not have a specific requirement under the plant 's licensing basis, but are required under Chapter 3 of NFPA 805, I vote affirmative on this Log.

CARLISLE: I concur with Mr. Sinopoli 's commen t dated 8/9/2000.

CONNELL: The last sentence of Section 3-1 states: "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." This s ta tement could be interpreted as to imply that the prior approval by the NRC of a specific deviation, at a specific plant, in a specific location in that plant, is now approved for all locations, at all plants, without any support ing engineer ing analysis or prior review by the AHJ." This was not the intent of the technical commit tee and the sentence should be revised to indicate that previously approved alternatives are simply acceptable.

DAVIS: I concur with Mr. Sinopoli 's comment . The original submit ted draft of Chapter 3 by Mr. Sinopoli should be included.

GARRETT: While the intent of the committee has been to have Chapter 3 address the fundamental elements of a fire protection program, the chapter was written, and remains overly prescriptive. NFPA 805 specifically states in the Scope, that the standard was written to address existing operating plants. No existing plant is known to comply with Chapter 3 as currently written. The fundamental requirements of Chapter 3 preclude the ability to use performance based or risk informed approaches to address those issues that are truly fundamental programmatic issues. The changes (plant modifications) required to meet the chapter as currently written would create costs beyond the benef i t sga ined from implement ing the standard. The requirements in Chapter 3, as currently written, will likely prevent many plants from adopting this standard.

The approach that was taken in the original comment was to maintain fundamental elements of the fire protection program, but relocate the detailed compliance methods to the appendix as an example of a method of compliance.

The commit tee acted on this commen t on a line by line basis. Certain portions were accepted and many others were rejected. Therefore, the preferred me thod of addressing this comment would be to accept the original comment in total.

KASSAWARA: The present version of Chapter 3 is too prescriptive. It would impose condit ions that would require backfits by many plants with no concomitant safety benefits. The essence of the original comment (by C. Sinopoli) was to provide a rewrite of Chapter 3 and the sections of Appendix A related to Chapter 3 in order to reduce the prescriptive requirements that it contained. The resulting committee modification did not accomplish this. The intent of this negative vote is to request a return to the original rewrite (submitted by C. Sinopoli) in its entirety.

NOTLEY: Over the course of its first several meetings on the development of 805, the committee discussed at length both the purpose and the character of Chapter 3. Consensus was finally reached on both of these issues. With respect to purpose, the commit tee agreed that Chapter 3 would have two principal consti tuents with separate but overlapping requirements - - the nuclear power industry and the NRC staff. The industry would require detailed guidance regarding the several elements that contr ibute to an acceptable fire protection program and the specific design requirements for acceptable fire protection systems and features. The staff would require [essentially these same] details in order to have a basis for review and acceptance of licensee fire protection submittals and a yardstick against which to measure their cont inued compliance with commitments and cont inued acceptability of a plant 's fire protection program and its implementat ion.

With respect to the character o f Chapter 3, the committee agreed that the elements of the fire protection program and the design requirements for the fire protection systems and features would have to be stated in considerable detail in order to satisfy the needs of both of the above constituents.

Therefore, the first sentence in 3-1 properly states the rationale for Chapter 3.

This chapter contains the fundamental eleme~lts of the fire protection program and specifies the minimum design requirements for fire protection systems and features [italics added for emphasis] .

The committee deleted several paragraphs which describe various fundamental e lements and minimum design requirements. These deletions, which are essential to fully satisfy the s tatement in 3-1, form the basis for my negative .vote and are listed below.

3-2.2.1 The policy document shall designate the senior management position with immediate authority and responsibility for the fire protection program.

3-2.2.2 The policy document shall designate a position responsible for the daily administration and coordinat ion of the fire protection program and its implementat ion.

3-2.2.3 The policy document shall define the fire protection interfaces with o ther organizations and assign responsibilities for the coordinat ion of activities. In addition, this policy document shall identify the various plant positions having the authority for implement ing the various areas of the fire protect ion program.

3-2.2.4 The policy document shall identify the appropriate AHJ for the various areas of the fire protection program.

3-3.5.2 [The first sentence, "Only metal tray m~d metal conduits shall be used .for electrical raceways" is already included. Only the last two sentences are missing and should be reinstated.] Thin wall metallic tubing shall not be used for power, instrumentat ion

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or control cables. Flexible metallic condui t s shall only be used in shor t lengths to connec t componen t s .

3-5.13 Headers fed from each end shall be permi t ted inside bui ldings to snpplv both sprinkler and s tandpipe systems, provided steel piping and fittings mee t ing the requ i rements of ANSI B31.1, Code for l'ower P~ping, are used for the headers (up to and inc luding the first valve) supplying the sprinkler systems where such headers are part of the seismically analyzed hose s tandpipe system. Where provided, sucb headers are considered an extension of the yard main system. Each sprinkler and s tandpipe system shall be equ ipped with an outside screw and yoke (OS&Y) gate valve or o ther approved shutoff valve.

3-5.14 MI fire protect ion water supply and fire suppress ion system control valves shall be u n d e r periodic inspect ion program and shall be s u p e : d s e d by one of the following methods .

(a) Electrical supervision with audible and visual signals in the main control room or o ther suitable constant ly a t t ended location.

(b) Locking valves m their normal position. Keys shall be made available only to author ized personnel .

(c) Sealing valves in their normal positions. This opt ion shall be ntilized only where valves are located within fenced areas or unde r the direct control o f the owner /opera to r .

3-6.3 The proper type of hose nozzle to be supplied to each power block area shah be based on the area fire hazards. The usual combina t ion spray/s t ra ight s t ream nozzle shall no t be used in areas where the straight stream can cause unacceptable damage or present an electrical hazard to fire f ight ing personnel . Listed electrically safe fixed fog nozzles shall be provided at locations where high voltage shock hazards exist. All hose nozzles shall have shutof f capability and be able to control water flow fi-om full open to full closed.

3-10.9 The possibility o f secondary thermal shock (cooling) damage shall be considered dur ing the design of any gaseous fire suppress ion system, but particularly with carbon dioxide.

3-10.10 Particular at tent ion shall be given to corrosive characteristics of agent decompos i t ion products on safety systems.

In addit ion to lhe above specific technical concerns , there are three editorial or typographical concerns that should be corrected. RATCHFORD: While the in tent o f the commi t tee has been to have Chapter 3 address the fundamen ta l e lements of a fire protect ion program, the chapter was written, and remains overly prescriptive. NFPA 805 specificall,y states in the Scope, that the s tandard was written to address existing opera t ing plants. No exist ing plant is known to comply witJ~ Chapte r 3 as current ly written. The f lmdamenta l requ i rements of Chapter 3 preclude the ability to use pe r fo rmance based or risk in formed approaches to address those issues that are truly fundamen ta l p rogrammat ic issues. The changes (plant modifications) required to mee t the chap te r as currently written would create costs beyond the benefits ga ined fi-om imp lemen t ing the s tandard. "The requ i rements in Chapter 3, as currently written, will likely prevent many plants f rom adopt ing this s tandard.

The approach that was taken in the original c o m m e n t was to mainta in f imdamenta l e lements o f the fire protect ion program, but relocate the detailed compl iance me t hods to the append ix as an example o f a me thod of compliance.

The commit tee acted on this c o m m e n t on a line by line basis. Certain port ions were accepted and many others were rejected. Therefore , the preferred me t hod of address ing this c o m m e n t would be to accept the original c o m m e n t in total.

SINOPOLI: While the intent of the commi t t ee has been to have Chapte r 3 address the fundamen ta l e lements of a fire protect ion program, the chap te r was written, and remains overly prescrpt ive . NFPA 805 specifically stales in the Scope, that the s tandard was written to address existing opera t ing plants. No existing plant is likely" to m e e t the r e q . i r e m e n t s o f Chapter 3 as current ly written. The f imdamenta l require inents of Chapter 3 preclude the ahility to use pe r fo rmance based or risk in formed approaches to address those issues that are truly f imdamenta l p rogrammat ic issues. The changes (plant modifications) required to meet the chap te r as current ly written would create costs beyond the benefits gained from imp lemen t ing the s tandard. The requ i rements in Chapte r 3, as current ly written, will likely prevent many plants from adopt ing this s tandard. The approach that was taken in the original c o m m e n t was to mainta in fundamen ta l e lements o f the fire protect ion program in the body of the s tandard, but relocate the detailed compl iance m e t h o d s to the append ix as an example of a m e t h o d of compliance. This approach would permi t the user to credit exis t ing o r new me thods of achieving the f imdamenta l p rogram e lement . The commit tee acted on this c o m m e n t on a line by line basis. Certain port ions were accepted and many others were rejected. Therefore , the preferred me thod of address ing this c o m m e n t would be to accept tire original c o m m e n t in total.

NOTE: Since the ballot on this Comment did not confirm the Committee Action, the comment is being rejected.

(Log #24) 805- 59 - (3-3.5.3 Except ion) : SUBMITTER: Edward A. Connel l , U.S. Nuclear Regula tmy C o m m i s s i o n C O M M E N T O N PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Revise text to read as follows:

Existing cable installed in areas impor tan t to safety thai has not passed a f lame propagat ion test acceptable to the authori ty having jur isdic t ion shall be provided with a coat ing acceptable to the authori ty having jur isdic t ion to inhibit f lame propagat ion. S U B S T A N T I A T I O N : None. COMMITTEE ACTION: Accept in Principle.

Revise text to read as follows: "Existing cable installed in areas impor tan t to nuclear safety that

has not passed a f lame propagat ion test acceptable to the authori ty having jur isdic t ion shall be provided with a means to inhibit f lame propagat ion unless a pe~Tormace-based evaluation demons t ra te s all acceptable risk." COMMITTEE STATEMENT: The r equ i r emen t for" inhibi t ing f lame propagat ion could be accompl ished th rough a per formance- based analysis. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 15 NEGATIVE: 9 NO T RETURNED: 2 Biggins, Vieira

EXPLANATION OF NEGATIVE: ALPERT: See my Explanat ion of Negative Vote on C o m m e n t 805-

58 (Log #82). CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated

8 /9 /2000 . CONNELL: The except ion provided for Section 3-3.5.3 allows a

per formance-based evaluation in place of a flame propagat ion test. As there is current ly no per formance-based evaluation, acceptable to the NRC, which assesses the risk impact o f us ing unqual i f ied cable, the except ion is presumptive. This except ion also excludes the considerat ion of alternative means of protection, such as in tray sprinkler protect ion or coat ing of cables with a material to inhibi t f lame propagat ion, which would provide adequate protect ion.

DAVIS: This r equ i r emen t goes above and beyond basic fire protect ion for cable installations. If there is a specific test which can be referenced, fine. However, as written, this leaves a wide open door to interpretat ion.

GARRETT: The c o m m e n t , as incorporated, will impose requ i rements that are not considered appropr ia te as a " fundamenta l e lement" of the fire protect ion program. The r equ i r emen t would require existing facilities that are in compl iance with existing regulat ions to backfit addit ional fire protection features or under take an extensive eng ineer ing evaluation for a large area o f the facility. In addit ion the term " impor tan t to safety" is no t def ined and likely will be imposed inappropriately and inconsis tent with ach ievement of the nuclear safety pe r fo rmance criteria.

HENNEKE: The change provides a performance-based approach in Chapte r 3, base requi rements . Performance-based requ i rements should be moved to Chapter 4, where the risk can be accounted for and determinat ion of the need to provide propagat ion can also be used.

KSOBIECH: This section should address the addit ion of new cables into the plant. Any new cables installed shonld be tested and mee t flame propagat ion s tandards acceptable to the AHJ.

Older plants installed cable with jacket ing that may not pass today's fire tests. The cables and potential fire hazards they present to nuclear safety have ah-eady been evaluated as part o f the fire hazards analysis, regulatoLy and l icensing process. This r equ i r emen t (in the prescriptive part of Chapter 3) would require plants to again reevaluate the cable jacket ing fire exposure . Existing cable jacket will be considered in a pe r fo rmance evaluation of a fire area u n d e r the per formance based port ion o f the code when necessary. Only new cable should be prescriptively required to mee t ~he cur ren t flame propagat ion standards.

RATCHFORD: The c o m m e n t , as incorporated, will impose requ i rements that are not considered appropriate as a " fundamenta l e lement" o f the fire protect ion program. The r equ i r emen t would re.quire existing facilities that are in compl iance with existing regulat ions to backfit addit ional fire protect ion features or under take an extensive eng ineer ing evaluation for a large area of the facility. In addit ion the term " impor tant to safety" is no t def ined and likely will be imposed

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inappropriately and inconsis tent with ach ievement o f the nuclear safety pe r fo rmance criteria.

SINOPOLI: As stated in the Scope, this s tandard applies to existing light water reactors. Older plants were cons t ruc ted prior to the formal adopt ion o f a s tandard test for rating the f lame propagat ion characteristics o f cables. T he r equ i r emen t to address the cables beyond what has already been found acceptable to the AHJ would const i tute a significant economic hardship . The text, as revised by commit tee action on this commen t , uses the phrase " important to safety" which is not def ined by the s tandard.

NOTE: Since the ballot on Comment 805-58(Log #82) did not confirm the Committee Action, the comment is being rejected.

805- 60 - (3-3.11 ): (Log #68) SUBMITTER: Stanford E. Davis, PA Power & Light Co. COMMENT O N PROPOSAL NO: 805-244 RECOMMENDATION: C hange text to read as follows:

"Electrical equ ipmen t , adequate clearance, free of t ransient combust ib le material , shall be..." S U B S T A N T I A T I O N : Add the word t ransient clearly differentiates that this r equ i r emen t is no t in tended to apply to in-situ combus t ib l e s COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commit tee Action on C o m m e n t 805-58 ( Log #82 ). NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:


805- 61 - (3-4.1): Reject (Log #76) SUBMITTER: T.A. O ' C o n n o r , A m e r G e n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Curren t wording requires "all o f the following" be compl ied with for on site fire f ighting capability (NFPA 600, 1500, and 1582). Revise to require compl iance with NFPA 600. S U B S T A N T I A T I O N : O u r plant has a license r equ i r emen t to meet NFPA 600. Reviewing NFPA 600, note that only NFPA 1500 is a reference in the body of NFPA 600. The NFPA 1582 d o c u m e n t is only referenced in the append ix o f NFPA 600. Having wording in NFPA 805 as presently stated in Section 3-4.1 is contradictory to NFPA 600 and presents new excessive requ i rements beyond the requ i rements of the existing s tandard and the requ i rements imposed by the AHJ. Adopt ion of this s tandard with the cu r ren t wording could represent a significant bu rden on implementa t ion with no improvemen t in nuc lear safety - these types o f burdens are barriers to adopt ion. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The s t a n d a r d says as applicable. It is no t the in tent to follow all of the standards. Also the submi t te r did no t provide specific wording for change. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:


805- 62 - (3-4.4): Reject (Log #53) SUBMITTER: Franklin D. Garret t , APS/NEI COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Delete the last sentence in Section 3-4.4.

Thia e q u i p m e n t ~ha:' co::form with the applicaS!e ~.~,FPA

standards:-. S U B S T A N T I A T I O N : T he last sen tence o f this section implies that whenever this s tandard is utilized the most cu r ren t version o f the NFPA s tandards mus t be used for fire f ight ing equ ipmen t . COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: Section 1-8 establishs the code o f record. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N C O M M I T T E E ACTION:



(Log #25) 805- 63 - (3-5.1(b)): SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATIO N: Add text to read as follows:

Provide at least two separate water supplies based on the following: S U B S T A N T I A T I O N : The cur ren t wording allows a single water supply. COMMITTEE ACTION: Accept in Principle.

See C o m m e n t 85-58 (Log #82). COMMITTEE STATEMENT: The in ten t was incorpora ted into C o m m e n t 805-58 (Log #82). N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:



805- 64 - (3-5.4, 3-5.5): (Log #52) SUBMITTER: Franklin D. Garrett , APS/NEI CO MMENT O N PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Revise text to read as follows:

Section 3-5.4 "At least one diesel engine-driven fire p u m p and one electric driven fire p u m p or two more ~e:~mic Category 1 C!a~ tE- electric motor-driven fire p u m p s canable o f be ing connec ted to r edundan t a Class IE emergency power buses capable o f providing 100 perc6nt o f the required flow rate and pressure shall be provided."

Section 3-5.5 "Each p u m p and its driver and controls shall be separated f . . . . . . . . . . . . . . . . . . . . . ~ ,:,~ rump~ and from the rest of the plant by rated fire barriers. In addit ion, each p u m p and its driver and controls shall be separated from each o ther by rated barriers if con ta ined inside a c o m m o n s t ructure or by a physical distance greater than 30 ft if in an ou tdoor a r rangement ." S U B S T A N T I A T I O N : This section is no t clearly related to the requ i rements for addit ional p u m p s and the r equ i r emen t for safety related fire water supplies is too restrictive. Section 3-5.3 implies that more than one fire p u m p is required to deliver water flow while Section 3-5.4 identifies that only one diesel p u m p or two electric p u m p s are an acceptable configurat ion. The use o f the word "more" in Section 3-5.4 implies that more than two pumps will be required. The r equ i r emen t for a Category 1 Class 1E p u m p is too restrictive. The p u m p s should be capable o f connec t ing to a 1E source however, the p u m p should be purchased in accordance with NFPA p u m p code requ i rements u n d e r a quality assurance program. Section 3-5.5 implies that more than one p u m p is required and requires rated barriers and does no t allow credit for adequate physical separat ion between the pumps . COMMITTEE ACTION: Accept in Part.

Do no t accept Part 1 as 3-5.4 was deleted. See Commi t t ee Action on C o m m e n t 805-58 (Log #82) for Part 2.

COMMITTEE STATEMENT: See C o m m e n t 805-58 (Log #82). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #54) 805- 6 5 - (3-5.16 Except ion No. 1 ): Reject SUBMITTER: Franklin D. Garrett , APS/NEI COMMENT O N PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Revise text to read as follows:

Except ion No. 1: Fire protect ion water supply systems shall be permit ted to be used to provide backup to nuclear safety systems t..9_o mitigate severe accident events, prcM~ed the fire protectie.n ;','attar -

~pec:fie~ b;.' tkc aFp!:ea~le analyg,s. S U B S T A N T I A T I O N : The except ion end the explanatory text (shown in under l ine) seem to contradict one another .

Exception No. 1: Fire protect ion water supply systems shall be permi t ted to be used to provide backup to nuclear safety systems

284


provided the fire Protection water supply systems are des igned and main ta ined to deliver the combined fire and nuclear safety flow d e m a n d s fqr the dura t ion sDecified bv the applicable analysis. A-3- 5.16 Mitigating severe accident events that can result in a fnel-clad damage is a top priocity. Since fires and o the r severe plant accidents are not a s sumed to occur s imultaneously, fire nrotect ion systems do not need to be des igned to hand le both d e m a n d s s imultaneously. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT" The applicable analysis would de te rmine if the fire event would be required to assume s imul taneously with an accident . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


NOTE: Since the ballot on C o m m e n t 805-58(Log #82) did not conf irm the Committee Action, the c o m m e n t is being rejected.

(Log #69) 805- 66 - (3-6.5): SUBMITTER: Stanford E. Davis, PA Power & Light Co. COMMENT O N PROPOSAL NO: 805-276 RECOMMENDATION: Delete the following text:

"...cause unacceptable damage or pre~ent an elect~ca! hazard to f i . . . . ~ . . . . . . 6 V . . . . . . . . . . . . .

SUBSTANTIATION: This d o c u m e n t is to be used at an electric genera t ion facility. The re are electrical hazards t h roughou t the facility. As written, a fire brigade could not use a combina t ion nozzle anywhere on site. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t t ee Action On C o m m e n t 805-58 (Log #82). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


NOTE: Since the ballot on this C o m m e n t did not conf irm the Committee Action, the c o m m e n t is being rejected.

(Log #26) 805- 67 - (3-6.4 Exception): SUBMrrTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Delete except ion. SUBSTANTIATION: Except ion is vague and unnecessary. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t t ee Action on C o m m e n t 805-58 (Log #82). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: ALPERT: See my Explanat ion of Negative Vote on C o m m e n t 805-58

(Log #82). CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated 8 /9 /2000 . DAVIS: The except ion, which was removed by this commen t ,

provided an acceptable and reasonable alternative for suppor t ing manua l f irefighting post SSE. T he except ion should remain.

GARRETT: The commen t , as incorpora ted , will require existing plants that are in compl iance with current regula t ions to perform extensive modificat ions to the s tandpipe and hose stations. The except ion, which was removed by this commen t , provided an acceptable alternative for suppor t ing manua l f irefighting post SSE. In addit ion, the r equ i r emen t is in conflict with Paragraph 2-3.1 assumpt ions for the deterministic approach .

HENNEKE: The except ion removed allowed plants that mee t the cu r ren t regulat ions to con t inue to operate u n d e r this s tandard. The re are several planlLs that will no t mee t the s tandard without the except ion, and will not be able to use the s tandard without significant cost. The except ion should remain.

KASSAWARA: Exper ience obta ined th rough investigation of over 25 major ear thquakes worldwide over the last 30 years has shown ( R e f - EPRI Report NP-6989) that ea r thquakes do not cause fires in convent ional commercia l power plants that are properly built, ma in ta ined and operated. These ear thquakes are many times the size of nuclear p lant Safe Shutdown Ear thquakes (SSE) in terms of

m a x i m u m ground acceleration and spectral con ten t in the damagin! f requency range. Further , nuclear plants are built and operated to an even h igher s tandard than these facilities. Rare events should not impose requ i rements on plant 's design basis unless demons t r a t ed tc cause severe consequences . Nonetheless , delet ion of this r equ i r emen t impacts only one of three e lements o f defense-in-depth Seismic requ i rements apply to the o ther two, i.e., prevent ion and saf shutdown e lements o f the defense-in-depth. The except ion provide~, flexibility for us ing alternative means o f supplying firewater u n d e r an SSE (which will be reviewed by the AHJ).

KSOBIECH: Leave except ion in the code. Older plants may not have been designed to mee t today's SSE

requirements for fire protection piping and equ ipment . They should not have to prescriptively meet these requ i rements if the fi 'eqnency and severity of such an event does not warrant it. Plants should have the option of us ing a pe r fo rmance based approach to define the risk and alternatives to address the risk using a pe r fo rmance based approach.

RATCHFORD: The c o m m e n t , as incorpora ted , will require existing plants that are in compl iance with cur ren t regulat ions to perform extensive modificat ions to the s tandpipe and hose stations. The except ion, which was removed by this co m m en t , provided an acceptable alternative for suppor t ing man u a l f irefighting post SSE. In addition, the r equ i r emen t is in conflict with Paragraph 2-3.1 assumpt ions for the determinis t ic approach.

SINOPOLI: The Scope of NFPA 805 states that this s tandard applies to existing light water reactors. Many cur ren t plants were not specifically des igned with s tandpipes and fire protect ion water supplies systems that are des igned for a Safe Shutdown Ear thquake (SSE). The impact on existing plants that have s tandpipe systems that have already been found to be acceptable to the AHJ would be significant. In addit ion, the r equ i r emen t to use the SSE as the design basis vs. the lesser Operat ional Basis Ear thquake (OBE) will also have a significant impact on the need to modify existing s tandpipe systems. The exception, which was removed by this commen t , provided an acceptable alternative for suppor t ing manua l fire f ighting following an ear thquake.

(Log #56) 805- 68- (3-9.4): Reject SUBMITTER: Franklin D. Garrett , APS/NEI COMMENT ON PROPOSAL NO: 805~t04 RECOMMENDATION: Delete the following text:

Sectiou 3-9.4 "Dic~el driven fire izu,=~p~ ~hal! he pro:erred by

SUBSTANTIATION: Section 3-9 o f this s tandard is describing the general system per formance requ i rements and not the locations which may require suppress ion systems. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The commit tee feels that a diesel- driven fire p m n p mus t be protected b)' a spr inkler system. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: GARRETT: Fire p u m p installations are typically located remotely

from unclear safety systems and equ ipmen t . The r e q u i r e m e n t for protect ion of diesel fire p u m p s is m u c h more associated with prevent ing economic loss than nuclear safety and should not be a fundamen ta l e l emen t of the fire protection program. This type of r equ i r emen t should only. be required when necessary to comply with the pe r fo rmance criteria.

HENNEKE: Sprinklers on Diesel Fire Pumps should be per fo rmance based. If the pumps are not near o the r vital equ ipmen t , then fire suppress ion is no t necessary.

KSOBIECH: Accept original commen t . Sprinklers should not be prescriptively required for the diesel fire

p u m p area. A per fo rmance based approach should be allowed to de te rmine the risk of such an event on the plant ' s pe r fo rmance objectives and allow for alternatives" to address the most appropria te protect ion to reduce the risk.

RATCHFORD: Fire p u m p installations are typically located remotely from nuclear safety systems and equ ipmen t . The r equ i r emen t for protect ion of diesel fire p u m p s is m u c h more associated with prevent ing economic loss than nuclear safety and should not be a fundamen ta l e l ement of the fire protect ion program. This type o f r equ i r emen t should only be required when necessary, to comply with the per fo rmance criteria.

285



(Log #27) 805- 69 - (3-11.1 Exception): SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatol 7 Commiss ion COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Delete except ion. S U B S T A N T I A T I O N : There are no per fo rmance criteria for the exception. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t tee Action on C o m m e n t 805-58 (Log #82). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:



' (Log #28) 805- 70- (3-11.4 Exception No. 2): SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Delete Exception No, 2. SUBSTANTIATION: This except ion could be in te rpre ted as allowing thermo-lag barriers to remain as is. COMMITTEE ACTION: Accept in Priuciple. COMMITTEE STATEMENT: See Commi t t ee Action on C o m m e n t 805-58 (Log #82). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 17 NEGATIVE: 6 ABSTENTION: 1 NOT RETURNED: 2 Biggins, Vieira

EJf..PLANATION OF NEGATIVE: ALPERT: See nay Explanat ion of Negative Vote on C o m m e n t 805-

58 (Log #82). CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated

8 /9 /2000 . GARRETT: NRC Generic Letter 86-10 "Fire Endurance Test

Acceptance Criteria for Fire Barriers Used to Separate R e d u n d a n t Safety Shutdown Trains Within the Same Fire Area" states that the guidance conta ined therein will be used by the NRC to review and evaluate the adequacy of fire e n d u r a n c e tests and fire barrier systems proposed by licensees or applicants in the future to satisfy existing NRC fire protect ion rules and regulat ions and as, such is not applicable to those ERFBS employed prior to its issuance. The exception deleted by this c o m m e n t was in tended to reflect this regulato D' condit ion.

HENNEKE: The exception that was deleted by this c o m m e n t was needed to meet cu r ren t design and current regulations.

RATCHFORD: NRC Generic Let ter 86-10 "Fire Endurance Test Acceptance Criteria for Fire Barriers Used to Separate Redundan t Safety Shutdown Trains Within the Same Fire Area" states that the guidance conta ined therein will be used by the NRC to review and evaluate the adequacy of fire endu rance tests and fire barrier systems proposed by licensees or applicants in the flmlre to satisfy existing NRC fire protect ion rules and regulat ions and as such is not applicable to those ERFBS employed prior to its issuance. The except ion deleted by this c o m m e n t was in tended to reflect this regulatoiy condit ion.

SINOPOLI: NRC Generic Letter 86-10, Supp lemen t 1 "Fire Endurance Test Acceptance Criteria for Fire Barriers Used to Separate Redundan t Safe Shutdown Trains Within the Same Fire Area" states that the guidance conta ined therein will be nsed by the NRC to review and evaluate the adequacy of fire endu rance tests and fire barrier systems proposed by licensees or applicants in the flmlre to satisfy existing NRC fire protect ion rules and regulat ions and as such is not applicable to those ERFBS employed prior to its issuance. The except ion deleted by this c o m m e n t was in tended to reflect this regulatory condit ion. EXPLANATION OF ABSTENTION:

DAVIS: While I agree with the phi losophy of the exception, 1 believe there is a middle g round that would address both sides o f this issue.

(Log #7) 805- 71 - (4-1.1): Accept SUBMITTER: James E. Lechner , Nebraska Public Power District COMMENT O N PROPOSAL NO: 805404

I RECOMMENDATION: Revise text to read as follows: "One success path ,of-systems necessary' to achieve..."

S U B S T A N T I A T I O N : Could imply the need for r edundan t syslems within a success path. A success pa th is defined later as cables and e q u i p m e n t in 4-1.3.1 and 4-1.3.2. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #29) 805- 72- (4-1.1): Accept SUBMITTER: Edward A. Conne l l , U.S. Nuclear Regulatm3J C o m m i s s i o n C O M M E N T ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete except ion. S U B S T A N T I A T I O N : Should be covered unde r Section 4-1.4 as a recovery action. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #111) 8 0 5 - 7 3 - (4-l.1 Exception): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services CO MMENT ON PROPOSAL NO: 805404

[ RECOMMENDATION: Delete the following text: Exception: U~e of repairs ~hall ~e !imite,-! ta long te rm

I , ~ , , e ~ a ~ , c c o ( c . - ~ ' . . ' . ~ ' . . . . . . . " . ' . f ~ [ pe r fo rmance cr;teria aa ~t appl:e~ tc the ;':ta[ au:;:ha:-iea :unct ion I ur lca : p:'i,or app:'c, va! ° f t,U..e repair; are o~tained from tl~,'~ AHJ.

S U B S T A N T I A T I O N : Recovery actions are pe r fo rmance based and therefore mus t be proven to be feasible. Additional l imitations do not have to be established. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log # l 12) 805- 74- (4-1.2): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Remove the r emainder of the sentence after "area" so that the sen tence reads as follows:

"...The per formance-based approach shall be permit ted to utilize determinis t ic me thods for simplifying assumpt ions within the

[ analys s area . . . . v . . . . . . . . . . . . . . . . . . u . . . . . . . . . . . : . . . . . . . . . I demons t ra te that t};e require~ a pc:Tar 'nance criteria i,3 met."

S U B S T A N T I A T I O N : Unnecessary detail in s tandard. COMMITTEE ACTION: Accept, NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins,,Vieira

(Log #84) 805- 75 - (Figure 4-1.2): Accept SUBMITTER: Chr i s topher P r agman , PECO Energy CO MMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Revise Figure 4-1.2 as shown on the next page:

286

Nuclear Safety


Deterministic approach

Performance-based approach

I ~ same fire

Document I achievement of I performance

criteria

No

Provide one of the following protection schemes: • 3-hour encapsulation of one

success path • 1-hour encapsulation of one

success path with suppression and detection

• 20 ft of separation without intervening combustibles and suppression and detection throughout the area

Document achievement of performance

criteria

Using information from Chapter 2, identify physical location of targets (equipment requiring protection)

Establish damage thresholds in ,¢

Provide one of the following protection schemes: • Radiant energy shield • 20 ft of separation without

intervening combustibles • Suppression and detection

throughout the area

accordance with Chapter 2 methodology

Determine limiting condition(s)

Document I achievement of ~ Establish fire

pe rfc ortmeda: ce I- I scenario(s)

~ No

Specify additional fire protection systems

and features

Figure 4-1.2 Nuclear safety capability assessment flowchart.

SUBSTANTIATION: The number of success paths for a given plant is typically not two. The true number of success paths available for consideration is de te rmined by the plant design, physical layout, and how much effort the licensee is willing to expend in creation and maintenance of analysis models. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #113) 805- 76 - (4-1.3.1): Accept SUBMITTFAR: Elizabeth A. Kleinsorg, Duke Engineering Services COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

"One success path of required cables and equipment to achieve and maintain the nuclear safety performance criteria without the use of ma-n,aal ~ actions shall be protected by the requirements specified in ei ther 4-1.3.2, 4-1.3.3, or 4-1.3.4, as applicable. Use o f manuat r_~.C.o_.~ actions to demonstra te availability of a success path for the nuclear safety performance criteria automatically implies use of the performance-based approach as outl ined in 4-1.4." SUBSTANTIATION: Reflects the change made in terminology. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #4) 805- 77- (4-1.3.3B): Reject S U B M r r r E R : Ronald Rispoli, Entergy Coq3. COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise last sentence to state:

"fire suppression system shall be installed in the analysis area such that the required cables and or componen t s are adequately protected." SUBSTANTIATION: Revised wording removes unnecessary requirement to provide protection capability throughout a large analysis area for which such protection may have minimal if any mitigating effect. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The proposed requirement is deterministic in a performance section. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


COMMENT ON AFFIRMATIVE: HOLMES: The Commit tee Statement, "The proposed

requi rement is deterministic in a performance section," is incorrect.

The title of Section 4-1.3 is "Deterministic Approach." Paragraph 4-1.3.3(b) describes deterministic requirements . The proposed revision in Comment 805-77 (Log #4) would introduce a performance-based approach in a section on deterministic requirements .

287


The Commi t tee Action should be corrected to read: "The proposed r equ i r emen t would i rmoduce a per formance-

based requ i rement in the determinist ic Section ,t-1.3.3."

(Log #30) 805- 78 - (4-1.3.3(a) Exception ): Accept SUBMrFrER: Edward A. Connel l , U.S. Nuclear Regulato D" Commis s ion COMMENT O N PROPOSAL NO: 805404

] RECOMMENDATION: Delete except ion. SUBSTANTIATION: No technical basis for" except ion. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: ALPERT: See c o m m e n t s by Sinopoli and Garrett. CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated

8 /9 /2000 . DAVIS: Sufficient informat ion to suppor t these except ions exists

in the SFPE Handbook of Fire Protection Engineer ing, NFPA Fire Protect ion Handbook , and ~,arious research project documen ta t i on following ASTM El 19 guidelines. The except ions should remain.

GARRETT: The m a x i m u m tempera tu re on the unexposed side of concrete slabs has been de t e rmined by full scale ASTM-E119 testing. The results o f this testing, as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapte r 7, indicate that a thickness of greater than 6 irrches meet the end point t empera ture criteria of the ASTM protocol for a 3 hour fire resistive rating. Additional substant iat ion is provided ira the Journa l o f Fire Protection Engineer ing 7 ('t) 1995 PP 125-132 "Modeling the The rma l Behavior of Corrcrete Slabs Subjected to the ASTM E l l 9 Standard Fire Condit ion." This research involved mathemat ica l s imulat ions of the heat and mass transfer th rough concrete when exposed to an E l l 9 fire. T he resuhs of this work indicate tha t the fire restiveness of a 6 inch thick concrete slab is capable of limiting the end point t empera ture helow 250°F for a period o f 196 minutes .

HENNEKE: Basis for the except ion has been found, arrd except ion should remain.

RATCHFORD: The m a x i m u m tempera tu re On the unexposed side of concrete slabs has been de t e rmined hy full scale ASTM- El 19 testing. The resuhs o f this testing, as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapter 7. indicate that a thickness of greater than 6 inches meet the end point t empera ture c r i t e r i a of the ASTM protocol for a 3 hou r fire resistive rating. Additional substant iat ion is provided in the Journal of Fire Protection Engineer ing 7 (4) 1995 PP 125-132 "Modeling the Thermal Behavior of Corwrete Slahs Subjected to the ASTM El 19 Standard Fire Condit ion." This research involved mathemat ica l s inmlat ions o f tire heat and mass t ransfer th rough concrete when exposed to an E119 fire. Tire results of this work indicate that the fire restiveness o f a 6 inch thick c o r r c i e t e slab is capable of l imiting the end point t empera tu re below 250°F for" a period of 196 minutes .

SINOPOLI: The m a x i m u m tempera tu re on tire unexposed side of concrete slabs has been de t e rmined by fifll scale ASTM-E119 testing. The results of this testing, as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapter 7, indicate that a thickness of greater than 6 inches meet tire end point t empera tu re criteria of the AS3"M protocol for a 3 hour fire resistive rating. Additional substant iat ion is provided in the Journa l of Fire Protection Engineer ing 7 (4) 1995 PP 125-132 "Modeling tire Thermal Behavior of Concrete Slabs Subjected to tire ASTM E l l 9 Standard Fire Condit ion." This research involved mathemat ica l s imulat ions o f the heat and mass transfer th rough concrete when exposed to an E l I9 fire. Tire resnhs of this work indicate that the fire restiveness of a 6 inch thick concrete slab is capable o f l imiting the end point t empera tu re below 250°F for a period of 196 minutes . (Technical inpu t provided by Ron Rispoli .)

(Log #114) 805- 79 - (4-1.3.3(a) and (c) Exceptions): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Sepdces COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Delete the following exceptions:

Exception: Conduit~ e m b e d d e d in concrete at a dept.h o f greater than 5 in. is co::sidercd to ha;'c an e~::ivalent le;'cl c f "-rotection as required above. "1 v

Excei~tio~}: C~mcluhs ~,-,,'nScdded i;" co.ncrcte at a depth of greater { . t ~ A * ~ . . ~ of protection as

SUBSTANTIATION: No technical basis. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:



8 /9 /2000 . DAVIS: See nay Explanat ion of Negative Vote on C o m m e n t 805-

78 (Log #30). GARRETT: See my Explanat ion o f Negative Vote on C o m m e n t

805-78 (Log #30). HENNEKE: See my Explanat ion of Negative Vote on C o m m e n t

805-78 (Log #30). RATCHFORD: See my Explanat ion of Negative Vote on

C o m m e n t 805-78 (Log #30). SINOPOLI: See my Explanat ion o f Negative Vote on C o m m e n t

805-78 (Log #30).

( Log #31 ) 805- 80 - (4~1.3.3(C) Except ion) : Accept SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatoty C o m m i s s i o n COMMENT ON PROPOSAL NO: 805-404

[ RECOMMENDATION: Delete except ion. SUBSTANTIATION: No technical basis for exception. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:



8 /9 /2000 , DAVIS: See my Explanat ion of Negative Vote on C o m m e n t 805-

78 (Log #30). GARRETT: The m a x i m u m tempera ture on the unexposed side

of concrete slabs has been de te rmined by full scale ASTM-E119 testing. The resuhs of this testing as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapter 7 indicate that a thickness o f greater than 4 inches mee t the end point t empera ture criteria of the ASTM protocol for a 1 h o u r fire resistive rating. Additional substant iat ion is provided in the Journa l of Fire Protection Engineer ing 7 (4) 1995 PP 125-132 "Modeling the The rma l Behavior of Concrete Slabs Subjected to

• the ASrlM E l I 9 Standard Fire Condit ion." This research involved mathemat ica l s imulat ions of the heat and mass transfer th rough concrete when exposed to an El 19 fire. Tire results o f this work indicate that the fire restiveness of a 4 inch thick concrete slab is capable of l imiting the end point t empera ture below 250°F for a period of 87 minutes .

HENNEKE: See my Explanat ion o f Negative Vote on C o m m e n t 805-78 (Log #30).

RATCHFORD: The m a x i m u m tempera ture on the unexposed side o f concrete slabs has been de t e rmined by full scale ASTM- E 119 testing, The results of this testing as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapter 7 indicate that a thickness o f greater than 4 inches meet the end point t empera tu re criteria o f the ASTM protocol for a 1 h o u r fire resistive rating. Addit ional substant iat ion is provided in the

Journa l of Fire Protection Engineer ing 7 (4) 1995 PP 125-132 "Modeling the The rma l Behavior of Concrete Slabs Subjected to the ASTM El 19 Standard Fire Condition." This research involved mathemat ica l s imulat ions of the heat and mass transfer th rough concrete when exposed to an El 19 fire. The results of this work

2 8 8

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Z . . . . . .


indicate that the fire restiveness o f a 4 inch thick concrete slab is capable of l imiting the end poin t t empera tu re below 250°F for a period o f 87 minutes .

SINOPOLI: The m a x i m u m tempera tu re on the unexposed side of concrete slabs has been de t e rmi ned by full scale ASTM-E119 testing. The results o f this testing as described in the SFPE Handbook of Fire Protection Engineer ing, Section 3, Chapte r 7 indicate that a thickness of greater than 4 inches mee t the end point t empera ture criteria of the ASTM protocol for a 1 hour fire resistive rating. Additional substant ia t ion is provided in the Journa l of Fire Protection Engineer ing 7 (4) 1995 PP 125-132 "Modeling the The rma l Behavior o f Concrete Slabs Subjected to the ASTM E l l 9 Standard Fire Condit ion." This research involved mathemat ica l s imulat ions o f the hea t and mass t ransfer th rough concrete when exposed to an E l l 9 fire. T he results of this work indicate that the fire restiveness of a 4 inch thick concrete slab is capable o f l imiting the end poin t t empera tu re below 250°F for a period of 87 minutes . (Technical inpu t provided by Ron Rispoli.)

(Log #115) 805- 81 - (4-1.3.4(b)): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Seiwices COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

(b) Separation o f required cables and e q u i p m e n t of r e d u n d a n t success paths by a non-combust ib le radiant energy shield. These assemblies shall be capable of wi ths tanding a m i n i m u m 1 /2 -hou r fire exposure when tested in accordance with NFPA 251, Standard Methods o f Tests o f Fire Endurance o f Building Const ruct ion and Materials. SUBSTANTIATION : Typo between versions. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VO TE O N COMMITTEE ACTION:


(Log #116) 805- 82 - (4-1.4): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services CO MMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

"This subsect ion shall provide for a per formance-based alternative to the determinist ic approach provided in 4-1.3. When the use o f man-trot r_~..Qy..¢.~ actions has resulted in the use o f this approach, the addit ional risk p resen ted by their use shall be evaluated. W h e n fire mode l ing for o the r eng inee r ing analysis, inc luding the use of-man-trot recovery_ actions for nuclear safety analysis, are used, the approach descr ibed in 4-1.4.1 shall be used." SUBSTANTIATION: Accurately reflects new terminology. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log # 117) 805- 83 - (4-1.4.1): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

Use o f fire mode l ing a~,8 nuclear zafety for the per formance- based approach. The following approach shall be used. S U B S T A N T I A T I O N : Editorial. COMMITTEE ACTION: Accept in Principle.

Revise 4-1.4.1 as follows: "Use of Fire Modeling." Revise 4-1.4 as follows: 4-1.4" Performance-Based Approach . This subsect ion shall

provide for a per formance-baseda l te rna t ive to the determinis t ic approach provided in 4-1.3. W hen the use o f manua l act ions has resulted in the use of this approach, the addit ional risk presen ted by their use shall be evaluated. W h e n the fire mode l ing o r o ther eng inee r ing analysis, is used the approach descr ibed in 4-1.4.1 shall be used. When fire risk evaluation is used, the approach described in 4-1.4.2 shall be used.

COMMITTEE STATEMENT: Editorial. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #118) 805- 84 - (4-1.4.1.2): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services CO MMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

"Establish Damage Thresholds . Within the analysis area u n d e r considerat ion, the damage thresholds shall be establ ished in accordance with Section 2-4 for the e q u i p m e n t and cables n eed ed to achieve th__.£ nuclear safety pe r fo rmance criteria." S U B S T A N T I A T I O N : Editorial. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #119) 805- 85 - (4-1.4.1.5): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text as follows:

"Protection o f Required Nuclear Safety Success Path(s) . Th e effectiveness of fire protection systems and features shall demons t r a t e that the circuits and c o m p o n e n t s requi red to achieve the nuclear safety per formance criteria are main ta ined free of fire damage and there is sufficient margin between the m a x i m u m expected fire scenarios and the limiting fire scenarios. Fire suppress ion activities in the fire area of concern shall no t impae~ RE_e.y._e_m the ability to achieve the nuclear safety pe r fo rmance criteria." S U B S T A N T I A T I O N : Editorial. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #120) 805- 86- (4-1.4.1.6): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Sexwices C O M M E N T O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

"Opera t ions Guidance. Guidance shall be provided to plant personne l that details the credi ted success path(s) for each analysis area, inc luding the per fo rmance of ma-mmt- r g . c . 9 , ~ act ions and repairs. ~ Re__._eCg.y._e~ actions az=d repair~ credi ted to achieve the nuclear safety pe r fo rmance criteria shall be feasible. (See Appendix B for acceptable mct.Sc, d~ fc, r cons idera t ions when assessing the feasibility o f ~ recovery actions and reFair~.)" S U B S T A N T I A T I O N : Editorial - - makes consis tent with new terminology. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 V O T E ON COMMITTEE ACTION:

AFFIRMATWE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #121) 805- 87 - (4-1.4.2): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services C O M M E N T O N PROPOSAL NO: 80.%404 R E C O M M E N D A T I O N : Bold first sentence. S U B S T A N T I A T I O N : Editorial. COMMITTEE ACTION: Accept in Principle.

Change the title to read: "Use of Fire Risk Evaluation".

COMMITTEE STATEMENT: Editorial. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26

289


VOTE O N COMMITTEE ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

VOTE ON COMMITTEE ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieil-a

(Log #32) 805- 88 - (4-2): Accept SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory Commis s ion COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Delete last sen tence . SUBSTANTIATION: Sta tement does not make sense. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #122) 805- 89 - (4-2): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Ser¢ices COMMENT O N PROPOSAL NO: 805-404

I RECOMMENDATION: Delete the following: "Both a determinis t ic and per formance-based approach shall be

met if radiation release f rom all sources due to fire or fire suppress ion activities is within the limits specified in 1-5.2." SUBSTANTIATION: Editorial. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #33) 805- 90 - (4-2.1): Accept in Principle SUBMrrTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Add text to read as follows:

Determinist ic Approach. The protect ion specified in Section 4- 1.3.4 provides an acceptable determinist ic m e t h o d for radiat ion release. SUBSTANTIATION: Section 4-2 does not specify a determinis t ic approach. COMMITTEE ACTION: Accept in Principle.

Add text to read as follows: "Determinist ic Approach. T he protect ion specified in Section 4-

1.3.4 shall provide an acceptable determinist ic m e t h o d for radiation release." COMMITTEE STATEMENT: The text was revised to make it mandatory. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #123) 805- 92 - (4-3): Accept SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineer ing Services COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Delete text as follows: "Life Safety. Life safety shall be provided for both nonessent ial

and essential facility occupants per the life safety per formance criteria o f 1-5.3. Both a determi:~i~tic and peffe, rmancc ba~ed

SUBSTANTIATION: Editorial. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #42) 805- 93 - (5-3): Accept SUBM1TrER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: After "plant decommiss ion ing" insert a c o m m a and "commensura t e with the changes in fire hazards and the potential release of hazardous and radiological materials to the env i ronment . " SUBSTANTIATION: To be consistent with the same wording used in Section 5-2 for the Fire Protection Plan. Fire protection p rogram e lements du r ing decommiss ion ing should be c o m m e n s u r a t e with the hazard. COMMITTEE ACTION: Accept. COMMITTEE STATEMENT: Editorial. In 5-2 change "environmental" to "environment" . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #43) 805- 94 - (5-3.1): Accept SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT O N PROPOSAL NO: 805-404

] RECOMMENDATION: Delete "as specified by Chapter 3". SUBSTANTIATION: See c o m m e n t on Section 5-3. If fire protect ion capability is ma in ta ined c o m m e n s u r a t e with the hazard, the reference to Chapter 3 is no t necessary. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #34) 805- 91 - (4-2.2): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory Commis s ion COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add text to read as follows:

Performance-Based Approach . The per formance-based approach specified in Section 4-1.4 provides an acceptable per formance-based approach for radiation release. SUBSTANTIATION: Section 4-2 does not specify a pe r fo rmance based approach . COMMITTEE ACTION: Accept in Principle.

I dd text to read as follows: "Performance-Based Approach . The per formance-based

approach specified in Section 4-1.4 shall provide an acceptable per formance-based approach for radiat ion release." COMMITTEE STATEMENT: Tile text was revised to make it mandatory. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

(Log #35) 805- 95 - (5-3.5.1): Accept in Principle SUBMITTER: Edward A. Connel l , U.S. Nuclear Regulatory C o m m i s s i o n COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Change m i n i m u m fire brigade staffing from 4 member s to 5 members . SUBSTANTIATION: The commi t tee has not provided a technical basis for the reduct ion in staff. COMMITTEE ACTION: Accept in Principle.

Delete the second sentence. COMMITTEE STATEMENT: R e d u n d a n t with Chapter 3. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


290


(Log #44) 805- 96 - (5-3.5.3): Reject SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Change the existing sentence to read as follows:

"On-site fire brigade equipment requirements shall be commensurate with thehaza rd levels." SUBSTANTIATION: To be consistent with 5-3.5.1 and 5-3.5.2,

• equipmen(requirements should be commensurate with remaining hazards, rather than as specified i~ Chapter 3. COMMITTEE ACTION: Rejec t . / COMMITTEE STATEMENT: The fire brigade requirements specified in 5-3.4 are necessary regardless of comissioning or aecomissioning. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(LOg #45) 805- 97 - (5-3.5.4): Accept SUBMITTF.,R: Frederick A. Emerson, Nuclear Energy Insti tute COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Replace the existing text to read as follows:

"Fire brigade drills and training sl-/ould be performed commensurate with the hazard. The off-site department interface requirements shall also he commensurate with the hazard." SUBSTANTIATION: To be consistent with 5-3.5.1 and 5-3.5.2, drills, training, and interface requirements should be commensurate with remaining hazards rather than as specified in Chapter 3. COMMITTEE ACTION: Accept. NUMBER OF COMMI'lq'EE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMrFrEE ACTION:


(Log #46) 805- 98 - (6-1.1): Accept SUBMITrER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Numerous NFPA standards are referenced in NFPA 805 but not listed here ( includingl0, 24, 101, 251,600, 1500, and 1582). All referenced standards shot~ld be added to this section, if they are listed in NFPA 805 as mandatory requirements. SUBSTANTIATION: In line with Section 6-1, the NFPA standards which are listed here in NFPA 805 as mandatory requirements should be included in Section 6.1.1. C O M M ~ ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMrrTEE ACTION:


(Lbg #67) 805- 99 - (A-l-6): Accept SUBMITTER: Stanford E. Davis, PA Power & Light Co. COMMENT ON PROPOSAL NO: 805-118

I RECOMMENDATION: Add app, endix material to definition for Through Penetration Fire Stop as follows:

A-1-6 Through Penetration Fire Stops should be installed in a tested configuration. These installations should be tested in accordance with ASTM E814 or equivalent test. SUBSTANTIATION: Through Penetration Fire Stops need to be installed in a tested configuration. Adding the reference to the ASTM test provides guidance to the designer/installer. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITrF~ ACTION:


(Log # 7 0 ) 805- 100 - (A-2o2(i)): Accept in Principle SUBMITrEg: Stanford E. Davis, PA Power & Light Co. COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Appendix item A-2-2(i) should be changed to A-2-2(h). SUBSTANTIATION: Appendix item needs to tie to correct section. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Committee Action on Comment 805-103 (Log #47). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITI'EE ACTION:

AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira ,

(Log #73) 805- 101 - (A-2-2.5.2): Accept in Principle SUBMITTER: Ira Heatherly, Tennessee Valley Authority COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise text to read as follows:

Once NFPA 805 is adopted as the Code of Record for a facility.. future equivalency evaluations (previously known as Generic Letter 86-10 evaluations) are to be conducted using a performance-based approach. • SUBSTANTIATION: Clarification of the applicability of NFPA 805 in this area will ensure that existing facilities will continue to be evaluated against their current requirements. COMMITTEE ACTION: Accept in Principle.

Revise text of first sentence of the second paragraph I follows: to read as I "Once NFPA 805 is adopted for a facility, future equivalency

evaluations. (previously known as Generic Letter 86-10 evaluations) I are to be conducted using a performance-based approach."

COMMITTEE STATEMENT: Clarifies the intent. Code of record is covered in 1-8. Note: The second sentence of the second paragraph will remain as shown in the ROP. NUMBER OF COMMrI"TEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #77) 805- 102 - (A-2-2.5.2): Accept in Principle SUBMrIq'ER: T.A. O'Connor, AmerGen COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Changc: "once 805 is adopted..." to ~once 805 is adopted by the AHJ and applied by an end user or licensed nuclear power generating plant operator...-" SUBSTANTIATION: The phrase ~once adopted" is not defined and in the case on NFPA 805 this adoption requires Nuclear Regulatory Commission approval and probably acceptance by individual licensees of nuclear power generating facilities. I f Nuclear Regulatory Commission-adoption is done by other means to be applicable to all facillities/licensees, separate reviews and approvals will be required along with consideration of regulations that address application of possible backfi t requirements. NFPA needs to accurately address this or avoid becoming involved in

lication under other regulations. MITTEE ACTION: Accept in Principle.

COMMITTEE STATEMENT: See Committee Action on Comment 805-101 (Log #73). ___ NUMBER OF COMMrITEE MEMBI~S ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #47) 805- 103 - (A-2-3.4.2): Accept in Principle SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete the paragraph beginning "The defense in-depth philosophy..." SUBSTANTIATION: This paragraph repeats the information in A-2-2(i). COMMITTEE ACTION: Accept in Principle.

Rewrite A-2-2(i) and A-2-3.4.2 as follows: A-2.2-(i) should be A-2-2(h) and no appendix material.

291


A-2-3.4.2 The intent of this requirement is not to prevent changes in the way defense-in<lepth is achieved. The intent is to ensure defense-in-depth is maintained.

Defense-in-depth is defined as the principle Rimed at providing a high degree of fire protection and nuclear safety. It is recognized that, independently, no one means is complete. Strengthening arty means of protection can compensate for weaknesses; known or unknown, in the o ther items.

For fire protection, defense-in-depth is accomplished by achieving a balance o f the following:

(1) Preventing fires from starting (2) Detecting fires quickly and suppressing those fires that

occur, thereby limiting damage (3) Designing the plant to limit the consequences of fire relative

to life, praperty, environment , continuity of plant operation, and nuclear safety capability

For nuclear safety, defense-in-depth is accomplished by achieving a balance of the following:

(1) Preventing core damage (2) Preventing conta inment failure (3) Mitigating consequence

Where a comprehensive fire risk analysis can be done, it can be used to help determine the appropriate extent of defense-in-depth (e.g., the balance among core damage prevention, conta inment failure, and consequence mitigation as well as the balance among fire prevention, fire detection and suppression, and fire conf inement) . With the current fire risk analysis state of the art, traditional defense-in-depth considerations should be emphasized. For example, one means of ensur ing a defense-in-depth philosophy would be providing adequate protect ion from the effects o4" fire and fire suppression activities for one train of nuclear safety equ ipment (for the nuclear safety e lement) and ensuring basic program elements are present for fire prevention, fire detect ion and suppression, and fire conf inement (for the fire protect ion e lement) .

Consistency with the defense-in-depth philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:

(1) A reasonable balance among prevention of fires, early detection and suppression of fires, and fire conf inement is preserved.

(2) Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

(3) Nuclear,safety system redundancy, independence , and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers).

(4) Independence of defense-in-depth e lements is not degraded. (5) Defenses against human errors are preserved.

An example of when a risk acceptance criteria could be met but when the defense-in-depth philosophy is not, is when is it assumed that one e lement of defense-in-depth is so reliable that another is not needed. For example, a plant change would not be justified solely on the basis o f a low fire initiation frequency or a very reliable suppression capability. COMMITTEE STATEMENT: The commit tee deleted redundancy and combined discussion on Defense-in-depth. NUMBER OF C O M M r r r E E MEMBERS ELIGIBLE TO VOTE: 26 VOTE' O N COMMITTEE ACTION:


(Log #75) 805- 104- (A-3-2.2.2): Reject SUBMITTER: T.A. O 'Connor , AmerGen COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Current wording states that preference be given to an individual with qualifications consistent with SFPE member grade status as the person assigned responsibility for day to day fire program responsibilities. Revise this sentence to put more emphasis on plant operations, maintenance and nuclear safety knowledge than on SFPE qualifications. SUBSTANTIATION: Nuclear safety remains the primary goal and focus. Fire protect ion programs and changes must haye nuclear safety as a paramount consideration since these changes may create risk. Having a standard that requires giving preferential t rea tment (perhaps it could be worded as a consideration) without addressing the nuclear safety issue is inappropriate. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: This material is in the appendix and not a requirement . NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

VOTE O N C O M M I T r E E ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #38) 805- 105 - (A-3-3.2(a)): Accept SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute C O M M E N T O N PROPOSAL NO: 805-404 RECOMMENDATION: Change "currenttransformer" to "current transformer". SUBSTANTIATION: Typographical error. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #37) 805- 106- (A-3-11.3): Accept SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute CO MMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete the words "[what does FHA stand for?] ". SUBSTANTIATION: This phrase is not needed since FHA is defined in Section 1-6. COMMITTEE ACTION: Accept. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 V O T E ON COMMITTEE ACTION:


(Log #124) 805- 107 - (Appendix B): Accept in Principle SUBMITTER: Technical Committee on Fire Protection for Nuclear Facilities CO MMENT ON PROPOSAL NO: 805-404 R E C O M M E N D A T I O N : Rewrite Appendix B as follows:

B-1 Nuclear Safety Capability Assessment. The primary purpose of the nuclear safety capability assessment is to demonstrate that

~ iven cable and equipment damage due to a fire postulated in any re area plant le,~atic, n, sufficient systems and equipment remain

available to achieve the following nuclear safety performance criteria (see Section 1-5):

([) Reactivity control (2) Inventory and pressure control (3) Decay heat removal (4) Vital auxiliaries (5) Process moni tor ing The purpose of this appendix is to identify attributes that should

be considered when ou:!ine accepta~Ic mcthad~ of demonstrat ing this capability. Other risk informed - nerformance based methods acceptable to the AHJ are permitted.

B-2 Nuclear Safety Capability Systems and Equipment. A list of systems and equipment that ensure the nuclear safety performance criteria can be achieved during and after a plant fire, regardless of fire location, should be developed. Th isprocess can be iterative and can require revisions to incorporate fire risk significant systems and equipment , if fur ther analysis in the circuit analysis or an@Ms f i ~ area assessment de termine additional systems or equ ipment to be fire risk significant. The process that follows describes the initial a t tempt to de termine which systems and equ ipment require evaluation. Other risk informed - oerformance based methods acceptable to the AHJ can be used to refine the list o f nuclear safety systems and equipment .

The set of systems and equipment to be evat',tated cons idered for nuclear safety 4s-an all inc!uaive list should addressing, as a minimum, the following:

(1) Systems and equipment required to place the plant in a safe and stable condit ion following a fire originating at power or while maintaining hot standby or hot shutdown. This fire also could result in a loss of off-site power, which would require achieving safe and stable conditions using power from on-site ac sources (i.e., emergency diesel generators) . This is typically a traditional Appendix R to 10 CFR 50 post-fire safe shutdown analysis.

292


(2) Systems and equipment required to maintain shutdown cooling capability following a fire originating while the plant is in the shutdown cooling mode.

(3) Systems and equipment required to maintain spent fuel pool cooling.

B-2.1 Assumptions (Plant Cond/tions at Time of Postulated Fire). In addition to the assumt)tions in Chapter 2. the following assumptions apply to this appendix.

Note: The follc;;=.ng aaanmption~ are provided to perform a

criteria arc met. Pcffo..~nance baaed information (i.e., e e o i p m e ~

" ~ ° : ~ basis c;'cnta) arc integral parts ~c ~ DCA ~.~ oz.~..,.4 z.~ cona:,dercd when pc,'-fo,,'-m, ancc r.~.~.4 -z't" . . . . . . . . . . . . . . . . . . . . .

( I ) The p lant is in a s tandard l i neup governed by ope ra t i ng procedures, ope ra t i ng modes, o r admin is t ra t i ve con t ro ls at the onset o f the f i re.

consequence of fire damage) of s;,'atcm~, eqnipment, . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . , or power o~t. t, . . . . . . . . . . . . . . . . . . . . . . . , . . . . , ' ~ g , O . . . . . O W , , , 8 . . . . . I r e . . . . . . . . . . . , . . . . . . . . 1 . . . . . . . . . . . . . . . .

(3) No abnormal a;.'atem transient;, ~.~.~.,:~. . . . . . ~,, .~ :+ . . . . . . h~o;o . . . . . . . . . v . . . . . . . . . . . . . . . . . . n o r do any o f tEese . . . . . . . wh ich are na t a d i rec t conacquc: 'cc o f .qre damage, nccur d n r i n g

(~) LOODS or bypasses within a system where mur ious oneration would not result in a loss of flow or inadequate flow to nuclear safety success paths shoutd- need not be-considered i ne !nded in t,~ae Nggh~

(3) For tanks, all outlet lines should be considered ~ f o r their functional requirements. For lines not required to be functional, a means of isolation should be included when necessary, to prevent unnecessary drawdown of the tank. Tank fill lin¢~ should also be considered e-'aluatcd as nccc~sa:%

(4) Properly oriented check valves function to prevent reverse flow in process systems.

(5) Normally closed manual valves (hand-operated only) will remain undamaged by a fire and can be relied upon for system boundary' isolation.

(6) Instruments exposed to a fire (e.g., RTDs, thermocouples, pressure transmitters, flow transmitters, and mechanically linked remote/ local indications) are assumed to suffer damaee that results in failure of the instruments. The instrument fluid boundary associated with these devices, with the exception of soldered fittings, is assumed to remain intact.

(7) Pioinm check valves, strainers, tanks, manual valves, heat exchangers, safety relief valves, and pressure vessels are assumed to remain functional durin~ and after a fire. The integrity of instrument tubing, with the exception of soldered fittings, is also expected to be maintained, though the accuracy of the instrument readin~ can be affected due to heating of the process fluid.

B-2.2 . . . . . . . . . . . ~ : Considerations for the Selection of Nuclear Safety Systems and Equipment.

Step 1: System Identification. Based upon documentat ion o__fon- plant design, risk insights and operation, plant systems required to achieve each of the nuclear safety ~ objeet~es should be identified. :r4~.is identification aho--!d 5c accomp!iahed by . . . . ~ ..... 8 . . . . s " , ~ v . . . . . . . 8, " " ~ hcena:ng . . . . . .

(1) Piping and instru:n, cnmtion diagram= ~.(DO'rn~'~.~o, {-2-) Single line .4: . . . . . .

t ~ DZ~.* . . . . . . . ca!culation.~ a n ~ -(4) Operating instructions

(7) FSAR accident ana!)'aea procc-uread (8) Emergency a n ~ a n o r m a ! cpcra=ing (9) .~c~ . . . . . .... ~.~,'*:~ including .u . . . . . . . . p . . . . . . xmr, P . . . . . . :~ ~ . . . . . .

88 29, Supplement t~m m . . . . . . : . . . . . . a :..a . . . . . . . . . . . . histories- Step 2: System Inter-Relationships. The selection of systems and

the documentation of how these systems fulfill the nuclear safety performance criteria should ~ be depicted in system-level logic diagrams, fault trees, or some other method that shows equipment dependencies. The documentat ion should co.......nsider4netude not only the required process systems but also the essential mechanical /environmental suppor t and essential electrical systems required to support the nuclear safety performance criteria. TaMe-

B 2.2 i : a . . . . . . . : j . . . . . j . . . . . . . : . . . . . . . . . . . . . . . . : . . . . . . . . . . . . . .

step 3: Equipment ld~ti~a~on. (a) P & l D s / f l o w d iagrams shou ld be used to iden t i f y the

equipment in the flowpath and the boundmy equipment within the systems that are required to achieve the nuclear safety objectives. This procedure hounds the act.m! components needed to mit.gate

(b) Equipment that is not directly in a required system flowpath, but whose spurious operation (undesired operation) could prevent achieving adversely affect aek.omp!iahing the nuclear safety objectives should be identified (e.g., boundary valve componen t whose spurious opening could divert flow away from critical equipment) . The potential for spurious operations of equipment should must- be considered when determiuing boundary valves and equipment selection,

For example, if two normally closed valves in series must spuriously open to result in an unrecoverable condition, then both valves should be identified on the nuclear safety equipment list (NSEL). If positive means is provided to preclnde spurious operation of one valve/component for non-high low pressure interface component [such as removing power to one of the two motor-operated valves (MOVs) during normal operation], then consideration additional analysis of the additional component (the other series valve) is not reqnired.

(c) Careful consideration must-should be given to equipment that could result in a fire-induced plant transient. The following is guidance on cons idera t ions that should be given in the identification of equipment that could result in a fire induced plant transient:

(1) l')'m-indured plant tn~tmtinlZ et,e'nt.~ Itransient~ and los.* of roolant m.cidents (LOCA.~)/. Transients are defined as anticipated operational occurrences (e.g., inadvertent safety injection actuation, loss of off-site power, over cooling, over filling of steam generators, spurious closure of containment isolation valves, significant loss of safety systems, station blackout, rapid cooldown, etc.) that initiate as a resuh of fire-induced circuit failures.

( g o ) Loss of prima U ~y.~tem inventory, The potential for fire initiated spurious actuation at reactor coolant pressure boundaries that could cause an uncontrolled loss of reactor coolant inventory [e.g., spurious actuation of primary coolant interfaces such as at the reactor head vents, normal and excess letdown at a pressurized water reactor (PWR), main steam isolation valves (Mg, t~BWRs)] should be considered. It should be noted that spuriona opening of reactor coolant hounds:) ' valves that could v-~.l,U...~:~" .... . . . . . . . . . . . v~ . .~ ~ m piping on an in t e r fac ing ayatcm or t h e l o s s 4 inventory conld not he mitigated i:" a':mcient time ',o achieve the • , .~..-. ' . . . . . . o-'--~er ..... performa::ce criteria are c!assi.qcd as high .~,.' .... p - m s s u ~ ~ ~ m oval ~r,x,m ~ . . . . . ~, preaanrizer

co@ling ~.],ratcin at boiling water reactor:; x~' ~xAa, . . . . z,~ 3 .,,..~".,. . . . . . . . . . . . . . ~':a~.~a m-ttm-m~e+estr4cN'eew-fi+eda-for circuit analysis and on,sequent ~.a.~:,: . . . . ~ c. . . . . . . . . . . . . :~.. r . . . . . . . . docun'ented in Chapter o ~. . . . . . . . , , . , ,,1~ t . , ~ . . . . . . . . . .~. .~o, aa Q ~r this standard.

(gb__) Rapid gook/own, PWR Transients that could result in an uncontrolled plant cooldown due to spurious operat ion of boundary valves should be considered, Interaction of plant systems such as ~ steam generator (PWR) power-operated relief valves, feedwater, reactor trip, turbine trip, and main steam isolation should be considered as well mu-" ~-- nnde:'stcod a.'-~l- revic;ved in order to make t!'is assesan'ent,

(4c_) Uncmtlndled przma O, injection. Transients tbat could potentially result in an undesired or uncontrolled injection into the reactor coolant system should be assessed. This can include spurious actuation of high-pressure injection sources (i.e., HPCS, RCIC, HPCI, feedwater for BWRs, high-head ECCS pumps for PWRs).

(,Sd) Electric power transients. "[ransients that could result in a loss of an_y_att-ac power supplies should be considered assessed. This loss can include spurious breaker actuations, onsite generating capability emeD,,ency diesel generator spurious starts or failures, or inadvertent paralleling of ac sonrces due to fire- induced circuit failures.

(d) Equipment that requires support such as cooling water, instrument air, HVAC, motive, and control power should be considered ..~a . . . . . . .~ _~., :Ao...~.~., in order to understand componen t and system inter-relationships and seqnential equipment loss impact.

(a) Loops or ~)'passea ;;.thin a ayate..m ;v.~.ere apma, ou~ operation would not rea'-!t in a Ions of rio;;' or inadequate fie;.; to ::'-clear safety~emeess-patl~Maould not be includcd in the NSEL.

293


c ..... :~..~i ...... : .......... For lines not required to ~'~ . ............

a m e a n s o f i so la t ion s h o u l d be. inch:deal w h c n ncceaaa: 7 to p rcvem- . . . . . . . . . . . . . . a . . . . . . * 4 . . . . . . . C *1 .~ * ~ * . 1 . T~ . . I , ¢ '~d l I ; . . . . . R ~ . , I A also b e e ........ as ne,cessa: 7.

(Ne_) Off-site power can be used as a sou rce of power tb r n u c l e a r safety e q u i p m e n t i f an analysis can prove it ia u n a f f e c t e d . All e q u i p m e n t r e q u i r e d to s u p p o r t off site, power c r the po r t ion of off-site power r e l i ed u p o n to ach ieve the n n c l e a r sat~'ty p e r f o r m a n c e ~ g - e a l s s h o u l d a lso be iden t i f i ed . :g4m

Off-site p o w e r s h o u l d conserva t ive ly be c o n s i d e r e d ava i lab le for those cases w h e r e avai labi l i ty of off-site power c o u l d adverse ly i m p a c t n u c l e a r safety (i.e., r e l i ance c a n n o t be p l aced on fire c a u s i n g a loss of off-site power i f the c o n s e q u e n c e s of off-site power avai labi l i ty a re m o l e severe than its p r e s u m e d loss). No c r ed i t s h o u l d be t aken for a f ire c a u s i n g a loss of off-site p o w e r p r e v e n t s p u r i o u s o p e r a t i o n s .

( h 0 I n s t r u m e n t s e n s i n g l ines s h o u l d be c o n s i d e r e d for ~ v i e w e d to case,as the p o t e n t i a l fo~ i n a c c u r a t e i n s t r u m e n t i n d i c a t i o n s a n d / o r s p u r i o u s e q u i p m e n t a c t u a t i o n s tha t c o u l d o c c u r as a resu l t o f an i n s t r u m e n t s ens ing l ine b e i n g e x p o s e d to a f ire a n d i nc r ea sed t e m p e r a t u r e s . Any i n s t r u m e n t s e n s i n g l ines tha t c o u l d p r e v e n t the fu l f i l lmen t of adverse ly affe,ct the n u c l e a r safety p e r f o r m a n c e c r i t e r i a s h o u l d be iden t i f i ed , a ssoc ia ted with the e q u i p m e n t tha t it cou ld impac t , a n d i n c l u d e d in the n u c l e a r safety a s se s smen t for review on a f ire a rea basis.

( i S ) I n s t r u m e n t a i r p i p i n g a n d c o m p o n e n t s (e,g., a c c u m u l a t o r s ) s h o u l d be c o n s i d e r e d -assessed- for ~fiability d u r i n g and a f te r the f ire in p r o v i d i n g the mot ive force for c r e d i t e d c o m p o n e n t s .

( j h ) Power supp l i e s , i n c l u d i n g a l t e r n a t e p o w e r suppl ies , for n u c l e a r safety e q u i p m e n t s h o u l d be ideu t i f i ed . I n t e r - r e l a t i o n s h i p s be tween p o w e r s u p p l i e s ( such as bus-t ie capab i l i ty a n d a h e r n a t e power supp l i e s ) s h o u l d also be iden t i f i ed . Th i s i n f o r m a t i o u is essent ia l iu d e t e r m i n i n g n u c l e a r safely e q n i p m e n t losses d u e to loss of a p o w e r supply. Ex,':-eme c a r e - s h o u l d bc t aken to e n s u r c . . . . . . . . . . . v . . . . . . . . . . . . . . . . . . . . . f unc t ion is rc!ie,d u p o n to ac . a . . . . _ ~ - e - r i t e r i a have powe,r . . . . . . i : - o ** ................... ~ ] r ....... gent ' , po;ve,r sources, If ;lot I . . . . . . . . . . . . . . . . . . . . . . . 8 . . . . 1 . . . . . . . ~ . . . . , e m e r g e s ' o r - . . . . .

g e n e r a t o r s ) , t h c : ~ b i t i t y - ~ f the n o n e m e r g e n c y - s o u r c e . . . . o,~. t" ..... ~" ......... °~"""" I ...... u,otr, ........ t ...•.o. ~.

e e e ~ p o n e n t ' s ac f i v e 4 m e ~ m . S t e p 4 : E q u i p m e n t I n t e r r e l a t i o n s h i p s . T h e necessa ry r e l a t i o n s h i p s

be tween i nd iv idua l n n c l e a r safety e q u i p m e n t a n d sys tems s h o u l d be u n d e r s t o o d a n d d o c u m e n t e d {-i.e., with a I c ~ ~ ~ of fires on . ~ l u i l m m m = a + ~ a y s ~ p:-o~e;-!y u n d e r s t o o d . Th i s is n e c e s s a : 7 to e n s u r e t l ' a t s u p p o r t

• , 4 . . . . . . . 4 . . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 : ~ , , , a n d e q n : p : ~ e n t . . ~ . . . . . . . . . . . . . . . .

p m p e v t y - m ~ d e f ~ . , a m p l e ,~s,~'~-:~ d g a g ~ m 4 s - ~ m ; ; ' n in Figu~-e B 4 ~ .

Step 5: Documentation. ~+..,:,,. . . . . . . ~ .~ . .w~ ~.~ d o c u m e n t e d ; . . . . . 1,... t o

d o c u m e n t e d i ~ a f e t y cqu ipme ,n t list (NSE4=%: . . . . ;12. ...... ; ........ + ..... I~.. (1) ""l"'~'" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m T~.... p l an t "t . . . . . . . . " P " t . . . . . . . . . . . . . . . . . . .

s h o u l d hc i den t i f i ed . {2) D c . ; c ; ~ p ; i ~ ; ; ~ d e : ; c r i p t i o ~ f :he e q n i p m ~

( ~ ' 1 t ' . , . . t . . . . . T I . . . . . . . *~ - - . A~o : . . . . * ~ C~ . - +I . . . . . . . ~ . . ~ * . . . . L . ; ~ I . t he

~ E - . . . . . . 8 ° ' %'1 . . . . . . ) ' . . . . . . . . . . . . . " . . . . .

I A ~, L ' . . , , g ; . . . . . . . . ~ f ; . . . . . . . . . . ' Y ' L . ~ C - . . . . . . . I ~ * ; . . . . . C * I . - - nuc~e,ar

. . . . . . . . . . . . . . . . . H " ' v . . . . . . . . . . . . . . . . . . . . . . . . . c i r c u

w i t 4 H 4 m m ~ y c o m p o n c ; ~ t - ( i n c l u d i n g c ~ n t r o l power , if "I]V II~"I'I~ / 0 II~'II~ ..... ~ ............... " ....... 8 a n n

v,~..,~, .... . o , o,,,~5.~ ...,~o, ..... / v, '~I" u'""'"8°" x ~ , , ~ .... lq~..~o

a~ - ' - no t r e q u 4 f e d - ~ r a c o m p o n e n ~ ' s safety f u n c t i o n (i.e., mot ive

power for a m o t o r o p e ; a t e d valve v.'hoac on!) . . . . . . ....~.~..I . . . . . . o-*~-lr . . . . . . . . . . . . . . : . . . . . . : . . . . . . . . . . . :^-~ Supp!cme,nta! inform..ation such as

oleo+vie s ing le l ine d r a w i n g s h o u l d be used to a id in the ide,nti ,qcation of tl~c .. . . . . . . . . . . . i .p,y . . . a ~..-~., . . . . . . . . . t.~.,

1 ¢ . . . . . . . . . . . . * . . | . ~ .

i. . . . . . . . . . . . . * l . .............. I . . ^ ~ ; * : ~ - SUCh . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v . . . . . . . . . as a p u m p tha t cou ld k . . . . . . . - ~ + : . . . . . A . . . . . . . . : ~ A A . , . - ; . . . . . . . . . . I . . . . . . . . : ~ _ k ~ . K

. . . . . ; , ; . . . . . . . . l . . . . . I . | I . . . . . * . . . . A ~ ¢ ~ ; I n - - . ; . ; . . . . . A . I ~ . ; . ~ A

pe~ ,+ i~n / s ta te s h o u l d be i den t i f i ed for each compone,n~ based on . . . . , . . . . . I . . . . . :_.: . . . . . . . . . . ,: . . . . . . . . . . . .~ . . . . . D0.Tno a n d . . . . . i - . .

. . g - , , , c . . . . . , ; ^ . . i ~.a0-..:. T I . . g..:l ..--.:,:--n iqfol-na, a t ion ,.-g ..... , - . . . . . . . . /~-.~ loss of a i r for the poa idon of a e o : n p a n e , n t on loss of v . . . . . , . . . .

a i r o p e r a t e d valves), : v x o . , r . . . . . . . . . . . . . D ~ ¢ . . . . . . . . . . . . t . . . . . :..: . . . . . . "~ ina t run 'en ta t ic , n

a; . . . . . . . . . : . ~ l - I:..~ .1: . . . . . . . . a I^~:~ d i a g r a : n a / f a u l t t ree A; . . . . . . . . I . ~ . . , I . 1 k ~ I ; ~ * ~ A ; . . o l . . A l : ~ . ~ i S i O n n n n ' . h e y , ............

. . , ; . I . ,, . . . . . . . . ; . . . . . . . . . . .~ _,,;A . . . . . . ¢ . . . . . . : . . . . . : - - !)'sea in

as h igh ' . . . . . . . . . . . . . . . : . . . . . q c . . . . ~ . ^ . . l a be c lear ly : a . . . , ; c - a : - o rde~ t O "lq"/ ............ 8~,,u .............. ~ ........ 7 . . . . . . .

(b_a) The ratiamfie, and methodMo~, bases for selection and

e x c l u s i o n o f n u c l e a r safety sys tems a n d e q u i p m e n t s h o u l d be d o c u m e n t e d a n d m a i n t a i n e d , Ca l cu l a t i ons aud analyses tha t have b e e n prev ious ly p e r f o r m e d in s u p p o r t o f o t h e r n u c l e a r safety ob jec t ives (i.e., s t a t ion b l ackou t , se i smic qua l i f i ca t i on ) can be u t i l i zed p r o v i d e d t he resul ts of these analyses have p roper ly c o n s i d e r e d the app l i cab i l i t y to post-f ire n u c l e a r safety.

( e b ) To d e v e l o p a n u c l e a r safety e q u i p m e n t list (NSEL) in a c o n s i s t e n t a n d r e p r o d u c i b l e m a n n e r , the fo l lowing s h o u l d be c o n s i d e r e d m e t h o d o ! o g / s h o u l d c o n s i d e r the fo l lowing:

( I ) V a l v e s / d a m p e r s c o n s t i t u t i n g system b o u n d a r i e s s h o u l d be i n c l u d e d in the NSEL. N o r m a l l y c losed m a n u a l valves a n d p r o p e r l y o r i e n t e d c h e c k valves c r e d i t e d as system b o u n d a r i e s a re n o t r e q u i r e d to be l is ted in the NSEL.

(9) M a n u a l d ra in , veut , and i n s t r u m e n t roo t valves s h o u l d uo t be i n c l u d e d in the NSEL.

(3) V a l v e s / d a m p e r s in the f lowpa th whose spu r ious o p e r a t i o n c o u l d adverse ly affect system o p e r a t i o n s h o u l d be i n c l u d e d in the NSEL. M a n u a l v a l v e s / d a m p e r s r e q u i r i n g r e p o s i t i o n i n g d u r i n g the post-f i re s h u t d o w n s h o u l d also be i u c l u d e d . M a n u a l w d v e s / d a m p e r s / c h e c k valves tha t do no t r equ i r e m a n u a l ac t ions d u r i n g the post-f ire s h u t d o w n s h o u l d n o t be i n c l u d e d in the NSEL.

(4) S a f e t y / r e l i e f vah, es p r m ~ d e d for e q u i p m e n t and p i p i n g p r o t e c t i o n s h o u l d n o t be i n c l u d e d . However , s a f e t y / r e l i e f valves p r o v i d i n g an act ive n u c l e a r safety func t ion , such as the s team g e n e r a t o r re l i e f valves a re excep t ions .

(5) Pi lo t s o l e n o i d valves s h o u l d be l is ted as s epa ra t e c o m p o n e n t s in the N S E L T h e c a b l i n g assoc ia ted with the s o l e n o i d valves s h o u l d be l is ted e i t h e r u n d e r the s o l e n o i d valve o r the a s soc ia t ed p roces s vah'e (e.g., the a i r o p e r a t e d valve) as d i c t a t e d bv the p r o j e c t i m p l e m e n t a t i o n p r o c e d u r e .

(6) Pumps , fans, tu rb ines , tanks, h e a t e x c h a n g e r s , and o t h e r e q u i p m e n t s h o u l d be i n c l u d e d on the NSEL.

I n s t r m n e n t a t i o n r e q u i r e d for p rocess m o n i t o r i n g o f the n u c l e a r , . " • T ~ I . I ~ 111 O O I R \ . . . ~ ;A~ . safety systems s h o u l d be ] d e n n f i e d . . . . . . . . . . ~ , v . ~ ; ' . ~ o

. . . . . . i . . . . . . . . . . . . . . . . . :,~.-:..~ : . . . . . . . . . . . . . . . . : . . . . . . : ~ . . , r e l i ed u p o n for h u e ! e a r safcty Moni ta r4n~ .

For c c n a i s t e , n c y in :'~ .... :r.,;..~ . . . . . . . . . I " "8 po;', 'cr s a p p l ) . . . . . . ; . . . . . . . . :" :°

s~dl~hgear, load ccntc~-ontro~ canto,re, distribution

...... i~ r. ...... :~,. ~ .......... t ........... a : ......... i.~ identified as

i m e : - r u p t i n g dev ices -such ms b reakers , switche,s, a n d f'uaca. Circuit- . . . . . I , , o 1 ~ r . . . - * l . ~ " . . . . ; . . . . . . . * o l . . . . . IA ; . . ~ 1 , . ~ 1 ~ y i . . . . . . y . . . . . . . . . . pr :ma@' ~ ' w , ~ . . . . . . . . . . . . . . . . . . . . . . . . . . an n t e r a c f i a n s , . , : . 1 . i . . . . . 1 . . . . . . . . . . :.~1. . . . . .a r . . . . . . . . "~ e n s u r e tha t al l p o t e n t i a l o l ~ d 4 a i t m : e m m d e s are addressed-..

294


" l P ~ k l ^ ' 0 0 0 / ~ T . _ : ~ , o,__. e . . . . . . . ,,r . . . . . . . . to A-~.: . . . . ~.r..~,~-- Safc.~" o ^ . c . . . . . . ~-: '^- '~ (Reactor m~- , ch...a . . . . Only) "IF .......... 1 ............... g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Decay heat rcmo; 'a !

Vital a-axi!iarica

Prc,ccaa mc,nitofing

Cc,ntrc,! ~^a. (ahc,rt * ~ ' ~ "~'~--;"-

. . . . . . . . : ~ - (no,final if v . . . . . . . . . . spray spray l O P D . . . . . ; l * k l . . . . . . : 1 ; ~ . ~ . I C ....................... ! 3 p ; - a ) ' ,,

• ,~.,-~'~, preac.urizer PORVs or

L-TOP

safety, . . . . . va . . . . and atmn:phcric ~1 . . . . . . . I . . . . D U D 1 . . ~R..*A . . . . . .

( ' ~ ' . . . . . . . . . . . . . . . . . . . i . e . , C H i C r g e n C y

: ~ o , . . . . . . . . . • . . . . . . . . . . . . A ~ ; , -

Cc,ntrc,! rods

'~;- ' - inj ) , ,~D • , . . . O , . . p r e 3 3 u , r c e c t i o n , . . . . . . ,

va!:'ea/autc,matic deprcasurizatic,n S"S*e~ ' l . . , y . D T J D . ~ * . . . . I { = f . . - - I . . . .

c ~ l ' ~ . . . . . . I I ~ C . , ~ l . . . . D L I D : n

° ~ l q . . . . . . . . . V . . . . . . . . . . 8 D I ~ D : ~ . I ~ , , . A . . . . . . . . I ; . . . . " . . . . . . . . . . . . . . ~ " ' . . . . . . . . . 5 ~ " ~ I * . . . . , ~ o K . , . A . . . . . . . . I : . . . . A ~

L" . . . . . . . . . . . . . . . . . / ' . . . . . . 8 . . . . 1 e . . . . . ~:.e., - ~ ' - -~m. . senc / dieael gencratc,ra, aaac,ciated pc,wet diatributian), reactor plant

se:-dce watcr, ~¢*AT~,,, HVAC,

pe~e~

P!*L~ I:;.=tr:=.v~C:;tatia~';. . . . . . . . .-:_~_ ~ . . . . i (A2ac, . . . . . . I 1 . . . . I a . . = ~ refueling?) [D. " : - a q::ca~oa far Rcactc,r coolant :;,'atom pret~ure and v . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . e, . . . .

• ~" . . . . . . . . L~ "1": . . . . . . . ,~^- r^- ,~-^ co,mac'-tree to reae!vc?] Rcactc,r cc,o!ant hc,t Ic~ temperature or exit core thermocc,aplca

~ , . . . . . . . . . * . . . . . . . . . . . . . A I . . . . I

. . . . . . . . . . . . . . . . . . . . . . . . . . . ( g F . . . . . . . . . . . ~ , ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . e . . , t ank level ind ica t ions , tam c ra tu re . . . . . . . . . . .

l.~..i ........... i.,12 ..... ~.l..,d.,d f~. DDI/D..

Supprcaaion pcc'c,! Ic;'cl and temperature Emergency c,r iaolation condcnac= le;'~-4 ~: . . . . . . :~ ; . . . . . . . . . . . . :^~ (c@, tank level indications, :cmperature, tic,;;" ratc~)

B-3 Nuclear Safety Capability Circuit Analysis. Circuit analysis for required nuclear safety equipment should be performed and documented. Circuit selection criteria include the following:

(1) Identification of all circuits that can adversely affect nuclear safety, including circuits with auxiliary contacts in the control circuits of nuclear safety equipment

(2) Exclusion of circuits that cannot prevent nuclear safety equipment from performing its nuclear safety function

Other risk informed - performance based methods acceptable to the AHJ can be used to refine the circuit analysis or its assumptions on circuit failure modes.

The equipment listed in the nuclear safety equipment list should be reviewed to determine all circuits (cables) that could prevent the equipment from performing their nuclear safety performance criteria. This analysis is focused on cables, typically either directly associated with componen t operat ion or associated with the componen t via auxiliary relays and contacts.

Aaac,ciated Other required circuits are defined as cables whose failure due to fire damage can affect nuclear safety capability. The following are three types of other required ~ac,ciatcd circuits:

(a) Spurious Actuations. The concern is fire damage to a cable whose failure could result in the spurious operat ion or real- operation of equipment that could affect nuclear safety. Identification of cables that could cause spurious components is an integral part of the circuit analysis methodology described in Section B-3.1

(b) Common Power Supply. The concern is fire damage to a cable that could result in a loss of power to a power source required to support nuclear safety equipment. This damage could occur if the upstream breaker/fuse for a nuclear safety power supply is not properly coordinated with a downstream breaker / fuse for the cable that is damaged. This is discussed in Section B-3.2.

(c) Common Enclosure. The concern is fire damage to a. cable whose failure could propagate to other nuclear safety cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) or the fire could somehow propagate along the cahle and impact nuclear safety equipment. This situation is discussed in Section B-3.3.

B-3.1 Definitions. The following circuit failure modes should be considered in the evaluation.

(a) Open Circuit. A condition that is experienced when an individual conductor within a cable loses electrical continuity.

(b) Shorts-try-ground. A condition that is experienced when an individual conductor comes in electrical contact with a grounded conducting device, such as a cable tray, conduit, or metal housing.

(c) llot Short. A condition that is experienced when individual conductors of the same or different cables come in contact with each other and may result in an impressed voltage or current on the circuit being analyzed.

(d) Sh:::'t Ci'r;';;it. A condkian that ia experienced ;vben indi;~dual cond'-ctc,r; (energized ar uncnergized) c,f a cable co,me i n con tac t w i th each o ther .

B-3.2 McthodcAc.gy Considerations for Performing Nuclear Safety Circuit Analysis. The following describes a process and the guidelines for performing the nuclear safety circuit analysis.

Step 1: Equipment Parameter Input. All equipment requiring circuit analysis and their functional

parameters (see Step 5, Section B-2) should be identified in the NSEL (see Sertion B-2). The functional parameters include normal and desired positions and failure posit ion/status upon loss of power and air (if applicable). These parameters establish the framework for circuit analysis.

Step 2: El~¢t6cal Drawing l~new. For each component to be analyzed, applicable elementary

diagrams, single-line drawings, connection diagrams, instrument loop drawings, and vendor drawings should be reviewed as necessary to identify all circuits connected to the component , including control cables associated with the componen t via auxiliary contacts in the component ' s control circuit.

Step 3: Consideratio'n.~ J b r ~ Cin'uit ldenti[ieation. Each circuit should be evaluated to determine which cables are

necessary-to support the nuclear safety function of the component . a fire-induced circuit failure of the cable can place the componen t in a posi t ion/condi t ion other than any desired posi t ion/condi t ion necessary hy that compouent in any mode of operation required to

295


achieve the n u c l e a r safety p e r f o r m a n c e criteria, the cable shonld be identified as a required cable. If a fire-induced circuit failure of the cable cannot spuriously reposition or prevent the desired operation of the eompoueut , it is not a required cable.

The following provides general evaluation guidelines for identification of required circuits:

(a) For spurious actuations and signals, all possible failure states, including all possible consequential damage, must be evaluated - - that is, the c o m p o n e n t could be energized or de-energized by one or more fire-induced circuit failure modes (e.g., hot shorts, short circuits, shorts-to-ground, and open circuits). Therefore, vah'es could fail open or closed; pumps could fail running or not running: and electrical d i s t r i b u t i o n breakers could fail open or closed• Multiple spurious actuations or signals originating fi-om fire-induced circuit failures could occur as result of a fire in a given tire area. Ho ...... , .... ~ ......... 1 ................................. ~ ......... he-

(b) A cable fault is a maintained condition that exists for the duration of the fire until an action has been taken to isolate the given circuit from the fire area or other actions have been taken to negate the effects of the spurious actuation. It cannot be assumed that the fault condition should terminate to a ground or open circuit condition that should stop tbe spurious actualion. C, enerally upou clearing of a fauh, the equipment would return to its desired state (if the de-energized state is the safe state of the safe shutdown component) . An example is dc solenoid vah'e or air-operated valve (AOV) with dc solenoid pilot that would fail to its safe position on a loss of control power to the solenoid. This would not he applicable to an MOV, which, if spuriously operated, would remain in that position following a loss o f control power.

(c) When evaluating a component failure (e.g., required nuclear safety component or a related component) , all circuits that are associated with the component ' s operation and all those circuits whose failure could affect the component ' s operation should he considered as required.

(d) A single hot short from an external source is postulated to affect any and all conductors, one at a time, within a cable that is required or is associated with a required nuclear safety component . In addition to evahmting the affects of a hot short circuit failure, these failures are postulated to occur, in conjunction witb other s imuhaneous individual or muhiple fire-induced circuit failures (i.e., shorts-to-ground, open circuits, internal conductor to conductor shorts) on other conductors associated with the circuit or cable

of concern. For h igh/ low pressure interfaces, properly aligned three phase and proper polarity dc hot shorts must be postulated to occur and these circuits must be included in the analysis.

(e) For muhi-conductor cables, energized conductor(s) could short circuit individual or multiple conductors within the cable that is ~equired or is associated with a required nuclear safety component .

(f) Plant-specific design features can preclude certain circuit failures f r o m o c c u r r i n g . For example, the use of grounded, metallic, armored cable or dedicated conduit would preclude external hot shorts from fl~rther consideration.

S l e p 4: l )orume*dal iom The circuit analysis for each component analyzed should clearly

docuinenl the desiguated nuclear safety cables and, more importantly, document and justify excluded cables and circuits. The rationale for selection or exclusion of a cable should also be documented. The circuit analysis should list required power supplies and identify those cab'los associated that must remain flmctional followiug a control transfer (i.e., vah'e control cables that remain functional following control transfer from a remote shutdown panel). Any use of risk-informed methods to reduce the scope of circuit analysis should be adequately documented. ~eump!cs of ca! ' !c disposition codes and consequences that could

." " " menting ch-c~iit analysis are pro;'idcd in Tab!c B

Specific Circuit Failure Scenarios The following are examples of typical circuit failure modes

(simplified circuits fignre have been used for illustrative purposes only).

(a) Lm~,-Voltage dc ( I V d c - 4 8 V do). These circuits are typically instrument signal cables for monitoring, protection systems, or control valves circuits (sensor to I /P converter). Shielded, grounded signal are cables typically used. Note that a conductor- to-conductor short on instrument cable or an intermediate resistance ground can result in a false instrument signal (same failure consequences as cable hot short, a fire-induced high signal). See Figure B-3.2(a) for an example of this failure mode.

Gode

Tabtc B ~.2 Example Diape, aitloa of & , a ~ C : d r e u i t s Nuclear Safer?' Cah!e Fa::tt Designation C o d e s For Required Cables tre . . . . . . ~_\

. . . . . . . . . . e . . . . . . . . . . . . . . . . . . . v . . . . . . . . m o t : r e p o w e r t o : h e c o m p o n e n t . A c a b l e f a i l u r e

e~ .... V ................. l .............. t ............ ~ ................ t ............

e o m p o n e n t 4 m m pc:forming its nuc!eawsMety4unet4on-by di~ab!ing control capability (loss of n * ~ l . . . . . . . . . . . I . . . . . C C. . . . . . :~ . . - -W; . . . \ T U n C~;~ . . . . . . . . . ,a . . ~ , ~ . . , ; ~ , . . c a u s e t h e c o m p o n e n t t o f a i l

~ , ~ r - s h - u 4 p opea or d m e - ( e i + e m ~

_ '" "- _rument signal cable. A cablc failure could p r e v e n t the c o m p o n e n t from

Y4;c cable is a power feeder c a b ~ ~ ~ - t o u.~'t" . . . . . . . . . . . . c,. .-~,,,~ ,,,, ,m,v~;~,,t . . . . . . . . . . , ,,~,~'l- failure so;:'ca :o fail the component to the desired nuclear safety posi t ion/condit ion "or othe:;vise . . . . . . . . ! . . . . . . . . . . . . . . . . . v . . . . . . . . . . . . . jx':-fcrm:n g :~ ..... ' . . . . . . c . . . . c. . . . . :~..

oemponen t to the desired : . . . . " " /condi t ion a n d / o r the los; of control does not prevent the co:npo::ent from pc:Worming it~ nuclear safe:}' functi'.m. T h e cable :s an instrumen: sigaa~ ~..k,... h . . . . . . . . . . . . J-,- g..:, . . . . . . . . . . b:..g in a ' . . . . r : . . . . . . . . . . . . o. s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 . . . . . . . . . not prevent the component from pe:Torming its e,-uclear o~r . . . . functiom.

..~, . . . . . r . . . . . . ~-'- by The cable is spared out and ia not e l e { : ~ ~ e ~ e 4 4 n - t , h e circuit; not a n . . . . . . . . . . . 7 . . . . .

w ~ . - ~ . ~ : . . . . . . . . q , . : . - d c ~ .~ . . . . . . p . . . . . . . . . p e r f o r m i :~ n u c l e a r s a f e t y c . . . . . . : . . . . . A : . . . . . . . . . a

. . . . . . . . . . . . . . . . . I . ~ , a b l e - . . . . . . e . . . . . v e n e : , ~ t o q ~ f i ~ r - m - I t s , , : . . . . . r s . . . . . y . u n c . m n .

The cable is not r e q t d v e d ~ m + o m p e , nent to peHbrm iu nuclear safety function and ia isolated by- ~1 :.l-S4

remain operable for the compone:-t to "~'*̂ "~v-"~'"' i~. ::::cleat- safer?' function.

296


( ~ Indicator

Protection-,~-~Power ' t24V A s ign. .~ ~ Short system/ supply I Jfdc Curren~~

interlock ~ (,~ ]

Transmitter Figure B-3.2(a) Instrumentation circuits.

Note: An internal shor t at po in t A could result in an e r roneous i n s t r u m e n t signal, which could cause erratic indicat ion or related e q u i p m e n t mal funct ion such as pro tec t ion system actuat ion or actuat ion of system interlocks.

Many plant protect ion circuits have logic circuitry that require mult iple inpu t signals in o rde r to actuate (i.e., one ou t of two taken twice, two out o f three, etc.). In these instances mult!]~le failures on i n s t r u m e n t inpu t signal cables (or in some specmc designs, mult iple conduc tors within the same cable) could cause a f i re- induced protec t ion system actuation.

(b) Medium Voltage Single Phase ac ( I I S V ac-220V ac). These circuits are typically used for secondary side control t ransformers o f Motor Control Cen te r (MCC) loads (e.g., MOV control, fan and small p u m p controls, ac solenoid opera ted valves and dampe r s ) .

For g r o u n d e d ac circuits, a hot shor t "source" can be from any energized circuit internal or external . For u n g r o u n d e d ac, external and internal ho t shor t "source" mus t be f rom same the

s a m e power supply. See Figures B-3.2(b) t h rough B-3.2(e) for examples o f open

circuits, shorts- to-ground, and hot shorts on typical g r o u n d e d 120V ac control circuit.

Valve open/close

,,: N A N . . _ _ ~ - - " - - ' - I ~ : I I

Notes:

/ Solenoid valve (energize to open)

_t_

Figure B-3.2(b) Ouen circuit.

1. An open circuit at points A and B will no t affect the valve posit ion if the control switch is open. However, it will no t permi t the operat ion o f the valve.

2. An open circuit at points A or B will de-energize the valve and change its posi t ion if the control switch is closed. In addit ion, the open circuit will no t permi t the opera t ion o f the valve.

Valve open/close

I1"

Solenoid valve (energize to open)

B

Figure B-3.2(c) Short-to-ground.

Notes:

A

m

1. Short- to-ground at po in t A will not affect the valve position if the control switch is open .

2. Short- to-ground at point A will affect the valve position if the control switch is closed by de-energiz ing the solenoid.

3. Short- to-ground at point B will no t affect the valve position irrespective o f the control switch position.

= I Open/ Open/ ' [ No ~ close/ No close/ No.-"

~ auto auto A C

,~ Solenoid valve y (energize to open)

3_ m

Figure B-3.2(d) Hot short example (external cable).

Notes: 1. A hot shor t at point A (from internal or external source) and

an internal short inside the cable connec t ing to PS-1 will energize the valve and change its position.

2. A hot shor t at point B will energize the valve and change its posi t ion.

3. A hot shor t at C and D (from an external source) need not be postula ted as f ire-induced circuit failure for this valve, except for high-low pressure interface valves.

Valve open/close PS- 1

- - - ~ : : : ~ c * ~ l " ~ . II . '~ Closes when t / pressure is /

high

Solenoid valve (energize to open) >

3 m

Figure B-3.2(e) Hot short example (conductor-to-conductor within same cable).

297


Note: Internal shor t inside the cable connec t ing to PS-1 and a shor t across the valve's control switch at po in t A will energize the valve and change its position.

(c) Medium Voltage dc (125V dc to 250V dc). These circuits are typically used for breaker controls for large pumps , switchgear, diesel generators , dc distr ibution, controls for solenoid-actuated air- opera ted valves, solenoid valves, dc MOVs, and so forth.

For an external ho t short , the hot shor t "source" (o ther than a 2 conduc to r p rope r polarity hot short) mus t be f rom the same dc source (battery) as the dc circuit u n d e r evaluation in o rder to provide a comple te circuit path.

For high-low pressure interface valves, removing power at dc panel du r ing normal opera t ion does hOe el iminate two conduc to r p roper polarity dc hot shorts from caus ing spur ious actuation. See Figure B-3.2(f).

<+)---L---qI:21

A i l Valve open/close

125V dc Solenoid (ungrounded) (energized to open)

B Valve open/close

T Figure B-3.2(f) Two-conductor proper polarity dc hot short example (solenoid valve or solenoid pilot valve for AOV).

Multiple shor ts- to-ground on an u n g r o u n d e d dc system on circuits f rom the same battery (or on u n g r o u n d e d ac circuits f rom the same t ransformer) could spuriously energize a normally de-energized circuit. See Figures B-3.2(g) and B-3.2(h). In addit ion, a ho t shor t on an u n g r o u n d e d dc circuit could also result in a protective device actuat ion (blowing a fuse or t r ipping a breaker) if a positive conduc to r shorts to a negative conduc to r f rom the same dc source or vice versa,

Ctq. B, CKT.&

" ( - ' ~ ' [ ~ ] Hot short Valve example

open/close-- H ~ ;ho'~ / sot rce

125V dc A ~ (san te dc (ungrounded) t sotJ rcel~

• Solenoid valve~'--,,~ (energize to open)

(-/ ~ - n

CKT. B, CKT. A.

/

! _=r--J B Valve _L . . . . . . . . . . . . . . . . T I

°pen/cl°se T } -2.

t2sv do A I- -T } ~ame-ac- ( (ungrounded) ' L L,_~°u rce~c~

< Solenoid valve- - Vl~T - _ . (energize to open) bno~s-t% y(e~ ,pen) .J~n~ rrtgu, ,u

~ ~ iL ~) ~-example

2. A short to g round at point A and a shor t to g ro u n d at point B will energize the valve and change its posit ion (bot tom figure).

(+)

_T_ C

CKr. B < CKT. A.

(-)

Figure B-3.2(h)

Notes: 1. A shor t - to-ground at po in t A and a shor t - to-ground at point B

will no t affect the valve position. 2. A shor t - to-ground at point A and a g r o u n d at po in t B and a

s imul taneous hot shor t at point C need not be postulated as a fire- induced failure for this valve except for high-low pressure interface valves.

Figures B-3.2(b) t h rough B-3.2(e) show simplified examples of open circuits, shorts- to-ground, and hot shorts on typical 120 vac control circuit. For illustration purposes , these figures I~ould also represen t de control circuits. "

(d) High Voltage ac Power (Three-Phrase Power 230V to 4KV ac). Circuits are typically used for MOVs, p u m p and fan motors, and ac power distr ibution. See Figure B-3.2(i) for an example of a three-phase ac hot short .

O1

0 2

Breaker MCC (open or closed)

MCC

430 V a c

IT MCC

I control power !

- L q- . . . . . . . . . J

: 42

Valve motor

Figure B-3.2(i) Three-ohase ac hot short examole (MOV),

Note: Hot shorts at po in t A, B, and C (from an external source, such as ad jacent power cables in the cable tray o r raceway) will energize the valve motor and change its position. This type of fire- induced circuit failure needs only be postula ted for a high-low pressure interface valve.

It should be noted that for high-low pressure interface vah'es, removing power at motor control cen te r (MCC) du r ing normal at power opera t ions does not e l iminate three-phase p roper phase ac ho t shorts at the power cable from MCC to valve from

Figure B-3.2(g) Ungrounded system - - multiple shorts-to-ground,

Notes: 1. A shor t at point A (from an internal or external source) will

energize the vah'e and change its position (top figure).

298

N F P A 805 m N o v e m b e r 2000 R O C

B-3.3 Other Required Circuits (Formerly Known as Associated Circuits). This section conta ins a descr ipt ion of addit ional circuits that should be cons ide red ~ for their impact on the ability to achieve the nuclear safety criteria. These circuits are divided into the following two categories: 1) c o m m o n power supply and 2) c o m m o n enclosure . These categories are descr ibed in the following paragraphs . Note that circuits requi red due to spur ious opera t ion ' cons ide ra t ions are addressed i n S e c t i o n B-2.2.

C o m m o n Power Supply Issues. The concern is fire damage to a cable that cou ld result in a loss o f power to a power source required to suppor t nuc lear safety equ ipmen t . This si tuation could occur if the ups t ream b reake r / fu se for a nuc lear safety power supply is no t properly coordina ted with a downst ream breaker / fuse for the cable that is damaged . Circuits associated by c o m m o n power supply are circuits that are powered from nuclear safety power supplies whose overcur ren t protect ion device is no t properly coordina ted with the overcur rent protect ion o f the power supply feeder [see Figure B-3.3(a)].

U A BL ;B

/ ) / ) t/

shutdown ~ N /k~) ) * ) A - ~ shutdown ° o = , , Ii ,urn,=

I P ~ xFire l " a 7

,.area,\ / , , , . .re., , ~ - - Redundant pumps ~

Common Enclosure Issues. "Circuits, which share enclosures with nuclear safety e q u i p m e n t circuits, mus t be analyzed to de te rmine their effect on r e d u n d a n t nuclear safety equ ipment . Th e concern is that the effects o f the fire can ex tend outside o f the immedia te area by m e a n s of f i re- induced faults on inadequately protected cables and could cause an overcur rent condi t ion resul t ing in secondary ignition. This si taution could result in remote damage that could disable the r e d u n d a n t train o f equ ipmen t .

A type o f c o m m o n enclosure failure mode is that o f secondary ignition. As shown in Figure B-3.3(b), the fire at i n s t rumen t B induces a fault on a cable that is no t adequately electrically .protected. The fault then results in the ignition o f a secondary fire in the vicinity o f the in s t rumen t A cables due to an overcur rent condi t ion resul t ing in cable jacket ignition. Again, the result could be the loss of the two r e d u n d a n t trains. A special type of c o m m o n enclosure issue involves cu r ren t t ransformers . Cur ren t t ransformers are used t h r o u g h o u t the electrical dis tr ibut ion system to mon i to r bus cu r r en t and provide differential protection. The cur ren t t ransformers are physically located on the electrical conduc to r (i.e., bus bar) and provide a signal th rough their secondary winding that is proport ional to the cu r ren t flowing th rough the conductor . Cur ren t t ransformers are des igned to t ransform high primary cur ren t into low secondary current . Theoretically, the secondary voltage is as high as required to main ta in a cons tan t primary-to-secondary cu r r en t ratio. An open ing in the secondary circuit causes excessively high voltages in the cur ren t t ransformer secondary circuit in an a t t empt to main ta in this ratio, which can result in an ignition of the t ransformer materials .

O n e type of cu r r en t t ransformer circuit s cheme is provided in Figure B-3.3(c). A fire in the control room could cause a fire- induced open ing in the secondary circuit of the cu r ren t t ransformers , caus ing an electrical fault at the cu r r en t t ransformers in both nuclear safety buses A and B. This si tuation could result in a fire within each bus, thereby disabling both buses and the r e d u n d a n t nuc lear safety p u m p s powered from them.

Figure B-$.3(a) Common power supply issues.

In Figure B-3.3(a), the postula ted fire could render nuc lear safety p u m p B inoperable . Fire-induced electrical faults affecting p u m p X or its power cable could cause breaker 1 to tri.'p, thereby render ing p u m p A inoperable if the two respective overcurrent isolation devices (breaker 2 and breaker 1) are not properly coordinated. ~fhis si tuation results in the in ter rupt ion of power to r e d u n d a n t nuc lear safety pumps .

Fire barrier

A / ~ Fire area I ~ Fire area II Potential t/~"JV ~ # I

Ignition 6 ~ ~ ~ Area of common enclosure

eway circuits " ~

EL ~ ~ ~ - R e d u n d a n t ~ instruments B-3.3(b) Concern common enclosure i s sues - - secondary ignition.

299


Switchgear room A Switchgear room B

Safe shutdown Safe shutdown Bus A Bus B

:)

A [3

Fire location / Control room

- - Safe shutdown components - -

Figure B-3.3(c) Common enclosure issues - - current transformer secondaries.

B-3.3.1 Mc*&.c.dolc.gy Considerations When foe Evaluating Common Power Supply Issues. The nuc lea r safety e q u i p m e n t list will be used to identify the electrical d is t r ibut ion buses, switchgear, load centers , and panels requ i red to provide electrical power to nuc lear safety equ ipmen t . The foUowing steps describe t he process r e c o m m e n d e d to evaluate those ci rcui ts (nuc lea r and non- nuc lear safety) that share a c o m m o n powersupp ly with nuc lea r safety equipm ent.

Step 1: Identification of Required Power Supplies. A listing of all the electrical distr ibution system buses, switchgear,

load centers, panels, and so forth, should be prepared utilizing the nuclear safety e q u i p m e n t list (Section B-2.2).

Step 2: Evaluation of Electrical Protective Device and Coordination. Each load on the nuclear safety power supplies for circuit

coordinat ion should be reviewed per accepted design practices. A list of the circuits that are not properly coordinated should be prepared .

Step 3: Disposition and Docun~ntation. The results o f the c o m m o n power supply issues should be

documen ted . For cases in which coordina t ion between series protective devices c anno t be demons t ra t ed , a c o m m o n power supply associated circuit should be as sumed to exist. These circuits should be disposi t ioned by one of the following means:

(1) Demons t ra te coordina t ion by ref in ing the available short circuit cu r ren t a n d / o r trip device characteristics.

(2) Identify protective device setpoint changes ( inc luding changes in fuse size a n d / o r clearing characteristics) that should establish coordina t ion .

(3) Incorporate the cables of concern into the nuclear safety analysis as required circuits for the affected power supply. This p rocedure should ensure that, on a fire area basis, the impact o f c o m m o n power supply circuits is evaluated.

B-3.3.2 Mct.~'~a-.c!c~'f~r- Considerafons When Evaluating Common Enclosure Issues .

Step 1: Common Enclosure - - Secondary Ignition. As descr ibed previously, s econdary ignition can be a real has the

potent ia l to be a conce rn d u e to f i re- induced e l e c m c a l faults o n inadequately protec ted cable resul t ing in secondary i gnilion. Th i s c o n c e r n can be add re s sed by evaluat ing the p lan t e lecuica l des ign to ensure that adequa te ove rcu r ren t pro tec t ion for all c o n d u c t o r s exists t h roughou t the plant.

Step 2: Current Traus former Seconda(y lgtHtion.

Tbc moat feaaib!e way to rcaolvc ~:e p rob!cm o f o p e n cu r r en t ~ a m f o m . c r acccndary; ; ' indinga ia to ena '-re "~at "&e cu:Tent

follov.a.ng ateps-desefibe "&c proceaa tb.at wa! b ~ a e ~ e d ~ , a t ~

u-ansfommrs that are cons t ruc ted such that an o p e n secondary circui t could cause igni fion o f the t ransformer .~9_uld vd-lt-be cons idered associated circuits of the power supply i n which they are located. Cur ren t t ransform ers t h a t a r e suscept ib le to igni fion _due to opeu s e c o n d m y windings and have secondarycircuits extending outside the fire a rea tbat are not isolated by u-ansducers _will have their circuits inc luded in the nuc lear safety assessment .

(I) Identify the ~witc..~ear and ofl~er . . . . . . . . . . . e - " ~" o~vt-'l:~°'~o a...~. . . . . . . . ~ . . . ~ m CU;Tent unn sfo..'-m.e=-a.

(2) Screen o u t ~he cum'ent tmna fonne ra -,~q~m_-.c aecondary

a. Do not exit tbc 5:-e a rea in which fl:e a;;a.tchgear ia located. u . . l l v v . . t u . ~ . , | ' ; . ~ 1 ~ + ~ . 1 k . , * . . . . . I1 . . . . .

{4)--Dctcn'ndne dm-eombus~ib!c ~ t c n t i a l for cacl" t y ~ o f

forth. The rcault~ aho'-!d he dc-c:'mcnted~ Ua~-gur ren t tmnsfo:.-n.era t h a t a r e proven no t to he amcep t ibk ' - to

; ~ . . ; d . . . . . . i i i k . . . . 1 . . A ~ A ~ . - ~ g ' . . . - * l . . . . . . a . . . .

Step 3: Disposition and Documentation. The results o f the c o m m o n enclosure issues should be

d o c u m e n t e d and mainta ined. Calculations and analyses that have been pe r fo rmed in suppor t o f o the r issues can be utilized provided the results o f these analyses have properly considered the

. . . . . . • ~ . . o C . . . . . . + 1 . ~ + anr)licabilitv to nest-fire nuclear safetv, C',~ . . . . . t . . . . . . . . . . . . . . . . . . . a r c 3 u a c c p f i b l c t o : ~ : * ; ~ - - A . . ~ , . . . . . . . . . . . . i . . . . . . . ; . . A ; . . . . .~

isotafed by waned " ' ir circuit~ inc luded in ",he

B4 Nuclear Safety Equipment and Cable Location. The purpose o f this step is to identify the location of the nuclear safety e q u i p m e n t and cables identified in Sections B-2 and B-3. Equ ipmen t and cables should be located utilizing available informat ion from existing databases, plant layout drawings, and, if necessary, p lant walkdowns. E q u i p m e n t and cables should be located by the smallest des ignator (room, fire zone, fire area, or anatysisfir_~_area) for ease o f analysis. WhRe-physie~ location o~ ~ t y e q u i p m e n t ~; . . . . ~ . . . . " ~ ' ~ , . . , . , . . . . . . , pumpa, matrumcnt~ . . . . .

c4foee~. The cable and raceway database (or the hard copy drawings)

establish the relat ionship between cables and their routing. Based on this informat ion, s u p p l e m e n t e d as necessary by plant walkdowns, the location of each required circuit can be identified. The process usually entails the following process:

(1) Cable-to-raceway rela t ionship (what condui t , cable trays, j unc t i on boxes, pul lboxes that a cable is rou ted th rough)

(2) Raceway-to-fire z o n e / r o o m relat ionship (what fire zones, rooms, or analysis fire areas that a raceway is routed th rough)

(3) Cable endpo in t location (fire zones, rooms, or sub-fire areas, where a cable terminates; this can be necessa W since cable endpo in t s can exist in p lant areas with no raceways conta in ing the cable - - i.e., a cable te rmina t ing in the control room, coming up th rough the floor o f a cable spread ing room)

(4) Fire z o n e / r o o m to ana tys i s -a re~f i re area relat ionship (what sub-fire areas comprise a fire area or ana!yai~ area)

B-5 ¢matysls- Fire Area Assessment. In Sections B-2, B-3, and B- 4, the in te rdependenc ies of nuclear safety systems, equ ipment , and cables have been established. An ana])'si~ area assessment for each fire area should be pe r fo rmed in order to demons t ra te ach ievement of the nuclear safety pe r fo rmance criteria. The nuclear safety e q u i p m e n t and their inter-relat ionships should be analyzed for each area to ensure that success pa th(s ) are available based upon the postulated e q u i p m e n t a n d / o r cable losses in that fire area. O the r risk in formed - per formance based me thods acceptable to the AHJ ~ can be used to refine the analysis" _fiaL¢_ area assessments and the assumpt ions made regarding the impact on a_ hhe success path. A success path ~ must- be identified that ensures each o f the nuclear safety pe r fo rmance criteria is met.

B-5.1 Methodology. Step I: Ident!fication ~?[ a[fected Systems aod Co.mpone,ts.

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NFPA 805 - - November 2000 ROC

All urlnrotected cables and e n u i o m e n t within a fire area mav be affected bv the fire• This assumof ion does no t imnlv that the fire ins tantaneously soreads throu~h 'out the fire area. bu t ra ther is in tended as a conservative a s sumnt ion to address the fact that• for this anaivsis, ne i ther the fire size n o r the fire intensity is rigorously de te rmined . As a m i n i m u m , this shou ld inc lude the following:

(a) Nuclear safety e q u i p m e n t affected by e i ther the equ ipmen t , its associated cables, or associated i n s t r u m e n t tub ing located in the area.

(b) Equip.ment affected by loss of a suppor t system (cooling water, ventilation) or loss o f power supply

Using the e q u i p m e n t compor .en t in ter - re la t ionship tool (logic diagrams, fault trees, etc.), identify the e q u i p m e n t affected by a fire in the area and the reason for the e q u i p m e n t loss (i.e., e q u i p m e n t located in the area, cable located in the area, power supply lost, etc.)

Step 2: Success Path Determination:. T h e following guidl ines shou ld be cons idered in the

de te rmina t ion o f ".:~c~ to a-.ctcrm.!ne :~e ~?'~tcm~ a~-:crzc!y ^..ffccte~ a success pa th ( s ) ~ctz..":=:'r=z*=on:

(a) Suppor t systems should be reviewed first in o rder to assess the impact on the systems be ing suppor ted . For example , it is r ecommef ided that the impact on the electric power distr ibution system be reviewed first. This review is pe r fo rmed so that an unnecessary a m o u n t o f t ime is no t spen t analyzing process systems (i.e., charging, auxiliary feedwater, residual hea t removal) that migh t no t have necessary power to suppor t the specific train o f c o m p o n e n t s to suppor t safe shu tdown.

(b) Similarly, cool ing systems such as service water and c o m p o n e n t cool ing water should be reviewed for the systems that they suppor t .

(c) r - _ a a : . : ^ _ :, : . . . . . . . . . "~'~ *~'"* T he impact o f a part icular fire on process mon i t o r i ng i n s t r u m e n t a u o n should be reviewed prior to assessing o t he r 1~-oc-ess system availability, to ensu re that process mon i t o r i ng ins t rumenta t ion is available foc the e o u i o m e n t credi ted c.::. the P.CS Inc. F ( : team gcacra tor ) :~a: ":

g w - ~ . (d~ t t . ; - - . t . ~ ~ . c r . . . . , 4 . . . . : . . . . . . . .~ . . . . . . . . . : _ , . .

A . . - - : . . . . . 4 . C * - - ~ ~ . . . . . 4 + k . . . . . . I ~ . . . . . . . , ,4 . . . .

~v . . . . . . . . . . . . . . . . . . . . . . y . . . . . . . . . . . . . . . . . . ~ . Note that some ecluipment can per form m o r e than one funct ion, such as RHR isolation valves that are normally closed du r ing plant opera t ion at power and du r ing the initial phases o f p lant cooldown bu t are requi red to be open to suppor t RHR opera t ion and the safety func t ion of long-term decay hea t removal.

. . . . . 1 + - : - - C - : I . . . . . C . 1 ~ - : - - . . . . . . . + ^ " I ' 1 , . . - : - - . + . . . . . + ~ . . : A

~'7~ ~ "&:2 :" ::":.~%.';: . _ _ _ ? . L - ~ : L ~ 2 2 : : ' ~ - - , : : - L : . . . . . . . , . . . . . . . .

. . . . I ~ . . . . . . . ~ ' - . . . . . I : ^ F . . . l . . . . . . ...1 . . . . . . . . . . . . . . . 1 . . . . . . . . . . . A * ~

; - - . + . . . . . + h . l . ~ ; . . . . . ; + ~ # k . . . . . . . # : - - - - ~ . ~ I A . - - - - A ~ . ~ + : - - - - - : - . I . ~

r c a ~ n g car, 5e affcctc~ ~uc to Ecat':r,g of Lkc process fiu'~. ( g e) Exclusionary analysis.

(1) An acceptable alternative to the me t hods described above is the pe r fo rmance o f an exclusionary analysis. This analysis would assure that cables and c o m p o n e n t s for at least one success pa th are no t located iu the area.

(2) In taking the pe r fo rmance based approach, factors discussed in (d) above should again be cons idered in the event e q u i p m e n t (or its cables) requi red for a success pa th to achieve and mainta in the nuclear safety pe r fo rmance criteria are in the analysis ~ area o f concern .

(3) .Although an exclusionary analysis uses a different approach than the step 2 success pa th de te rmina t ion process described in i tems (a) t h rough (f), all o f the fundamenta l s , such as c o m p o n e n t inter-relat ionships, ~ ~ action feasibility, and sb forth, should mus t - s r tg be cons idered . Potential losses o f e q u i p m e n t due to c o m m o n power supply, c o m m o n enc losure concerns , and spur ious ac tuat ions and signals shou ld also be addressed. Success paths relied upon for achieving the nuc lear safety pe r fo rmance criteria us ing the exclusionary analysis rqethod should must- be clearly d o c u m e n t e d for each a n ~ y s i s fire area in o rder to ensure

future changes do no t ~ a~vcr~c! 7 affect the results o f the analysis•

{D T h e basis for successful demons t r a t ion of the nuc lear safety ne r fo rmance criteria for a fire in each ~ d - y ~ f i r e area should be ~locumented for ea~h p 'ccc of e q u i p m e n t potentially affected by a fire in the area. This documen ta t i on is narticularlv imnor t an t when r e d u n d a n t nuc lear safety e q ¢ i p m e n t is located i n the area and reliance is nlaced on ohvsical senarafion, fire nro tec t ion svstems and features, o r recovery actions, in o rder to achieve an d main ta in the nuclear safety ne r fo rmance criteria. B-5.2 MethodologY. ~ . S u e c e m Path Resolution Consi d e r a l l o n & ~ .

(a) The magni tude , dura t ion , o r complexi ty o f a fire c an n o t be foreseen to the ex ten t o f predict ing the t iming and quant i ty o f f ~ induced ~rc-~i-~ failures. Nuclear safety circuit analysis is no t in t ended to be pe r fo /med at the level o f a failure modes and effects analysis since it is no t conceivable to address every. combina t ion o f ~ failures. Rather, for all potential spur ious operat ions in any anatysis fire area, focus shou ld m a s t - b e on

~-a'~ . . . . . . . . . . . :~_a each potential spur ious opera t ion and mit igat ing the effects o f each individually. Mult inle sour ious actuat ions or signals or ig inat ing f rom f ire- induced circuit failures

v . . ~ -

¢901d occur i~ result o f a ~ v e n fire. The s imul taneous e q u i p m e n t or c o m o o n e n t mal-oDerations resul t ing f rom fire induced failures. Unlgss the circuit failur¢ affects m~lt ip/e c o m n o n e n t s , is no t expected to initially occur. However. as the lqre nronauates , any and all sour ious e q u i p m e n t or c o m o o n e n t actuat ions, if no t orotected o f oronerlv mititrated in a timely m a n n e r , could occur. u . . . . . . . . . . k - : , ; . . . . . • ~ " ~ . * - - A • . . . . . . . . : ~ . . t . . . . . . . . Z . . . . . I - - . . . L ^

...= . . . . . . . . +: ............ .4 K., +I~^ ~ - :--.4 .... .4 C~:l .... :~

orcu:~. :ha: car. affec t mult :p!z c cmpcnzn : z . Spur ious ac tuauons or signals that can prevent a requi red c o m p o n e n t f rom accompl i sh ing its nuc lear safety func t ion mus t be appropriately mit igated by fire protect ion features.

(b) Reliance on an assumpt ion o f only a single spur ious opera t ion without opera tor in tervent ion (i.e., h a v i n g t w o normally closed MOVs in series with spur ious cables rou ted th rough an area, and a s suming only one of the valves could spuriously open) should no t be relied' upon for ensuring a success path remains available. Therefore , in identifyiug the mit igat ing act ion for each potential spur ious .opera t ion in any given ~ y s i s f_tr_~ area, should not be relied upon to mitigate the effects of one spur ious opera t ion while ignor ing the effects o f a n o t h e r potential spur ious opera t ion.

Where a single fire can impact the cables for ~ High-low pressure interface valves in series, the potential for ~ valves to spuriously opera te s imul taneously should be cons idered . Removin~ power to two or m o r e normally closed bigh-iow pressure mterface valves in series du r ing no rma l opera t ion (which reduces credible spur ious opera t ions to mult iple three-phase ac ho t shorts o r mult iple p roper polarity dc ho t shorts on mult iple valves) is an acceptable m e t h o d o f ensu r ing RCS integrity without addit ional analysis or fire protec t ion features. This criteria applies to all ~ fire areas, inc luding the control room, and to a l l circuits regardless o f whe the r or no t they can be isolated f rom the control room by the actuat ion o f an isolation t ransfer switch.

(d) T~c ~-.~i~ for ~'accc=~f'a! ..acmc.n~:ra:ior. o f :ke n-ac!car ~^~=~'

f i . . . . . . . . ~ - . e . . . . . . . . . . ~ . , . ~ . . ~ . . . . . . . i ~ . . . . . ~ . . . . 7 . . . . ! " ~ . . . . .

( d e ) Succc~: ~_ , t . . . , . . . . . . a: . . . . . : . . . . . : . . . . . . "~ b?' a £ re :.n the ar.a!ys:,z fire area ~ c , uM be. d~cur,,,cntc~ u~:,r,g czr.s:,ztcr.t s tatcmcnt~. The per formance-based approach shou ld cons ider the fire protect ion systems and features o f the room and what effects the fire scenarios would have on the nuclear safety e q u i p m e n t within the area u n d e r considerat ion.

(2) P.cp!accra.cnt of c=E!:.ng

add:. : . . . . . . . u . . . . r . . . . . . . . . . . . . . . ";'"~ . . . . . . . . . . . . . . . . . . . . . . . . addrz~.~e~a:

to rcmos'c = capped fit*-ng are. no t cc.r.~M::rcd rcpa!r~. T~c "a3c of

301


...... ' ......... ':~- i~ not ccn~i~crcd a repair=

and ma:c~a!~ required to i mp l emen t the repa:r~ a,,c, no t need to, bc

Remc;'a! c f f'-~eo" for i~c.lation caa bc considered a man' -a! ac t ic~ ~o..u: . . . . . . ~.~ ,:m:,. ,:~..o u~,^...~ if found accep~b!e to the

( e g ) Mancmt Recover~ actions can be pe r fo rmed as part o f a per formance-based approach subject to the l imitations o f Chap te r 4 of the s tandard to mit igate a spur ious actuat ion or achieve and mainta in a nuclear safety pe r fo rmance criteria. For the e q u i p m e n t requi r ing manua l actions, informat ion regarding the location requi r ing the action (fire area in which a fire requires the action), the location of the action (location at which the action is to be performed) , and the time constraints to per form the actions s h o u l d b e . . . . . ~.,~.~ ~k..~.~ ~ , u ...... h . ^c .~. . . . . . . . . func t iona l analysis-)- to assess the feasibility of the proposed action.

(1) The proposed ~ m a m i a l act ions should be verified in the fo ld to ensure the action can be physically per formed u n d e r the condit ions expected du r ing and after the fire event.

(2) When ~ m a m ~ actions are necessary in the analysis fir._._c_e area u n d e r considerat ion, the analysis should demons t r a t e that the area is tenable for the actions to be pe r fo rmed and that fire or fire suppressan t damage will no t prevent the manua l action f rom bei~;~ per formed.

The l ighting r e q u i r c m e n ~ should be evaluated to ensure sufficient l ighting is available to per form the in tended action.

(4) Walk-through of opera t ions ~'uidance po; t fire . . . . . . "~ . . . . . (modified, as necessary, based on the analysis), should be conduc ted to de te rmine if adequate manpower is available to perform the vc~{~r-e4 potential recovery actions within the time constraints (before an unrecoverable condi t ion is reached) .

(5) The commun i ca t i ons system should be evaluated to de te rmine the availability o f communica t ion , where required for coordinat ion of ~ ~ actions.

(6) Evaluations for all actions, which require traversing th rough the fire area a n d / o r an action in the area of the fire, should be pe r fo rmed to de te rmine acceptability.

(7) $~fficient t ime to travel to each. action location and perform the action should exist. T he action mus t be capable o f beinu identified and oer formed in the t ime reouired to suonor t the associated shu tdown funct ion(s) such that an unrecoverable condi t ion does not occur. Previous action locations should be considered when seauent ial act ions are re(mired.

(8) The re should be a sufficient n u m b e r o f essential personne l to oerform all o f the reouired actions in the t imes required, based on the m i n i m u m shift staffinm The use of essential oersonne l to oer form actions should not interfere with any collateral fire brigade or control room duties they may need to oer form as a result; of ~he fire.

(9) Any tools, e u u i n m e n t or keys reuuired for the action shall be availabie and accessible. This includes considerat ion o f SCBA and per~gnnel protective e q u i p m e n t if reouired. SUBSTANTIATION: This is a result o f a detailed review by the nuclear safety task group. The in tent ion of the changes is to ensure the d o c u m e n t is consis tent with industry positions. COMMITTEE ACTION: Accept in Principle.

Editorial: 1) Change "associated circuit" with "other required circuit" everywhere in Append ix B.

Revise text to read as follows: Appendix B Nuclear Safety Analysis

B-I Nuclear Safety Assessment. The primary purpose o f the nuclear safety assessment is to demons t ra te that given cable and e q u i p m e n t damage due to a fire postulated in any fire area, sufficient e q u i p m e n t remains available to achieve the following nuclear safety pe r fo rmance criteria (see Section 1-5):

(1) Reactivity control (2) Inventory and pressure control (3) Decay heat removal (4) Vital auxiliaries (5) Process mon i t o r i ng The purpose o f this append ix is to identify attr ibutes that should

be considered when demons t r a t i ng this capability. O t he r risk

in fo rmed - pe r fo rmance based me thods acceptable to the AHJ are permit ted.

B-2 Nuclear Safety Systems and Equipment. A list of systems and e q u i p m e n t that ensure the nuclear safety per formance criteria can be achieved du r ing and after a plant fire, regardless of fire location, should be developed. This process can be iterative and can require revisions to incorporate fire risk s ign i fcan t systems and equ ipment , if fur ther analysis in the circuit analysis or fire area assessment de te rmine addit ional systems or equ ipm en t to be fire risk significant. The process that follows describes the initial a t t empt to de te rmine which systems and e q u i p m e n t require evaluation. O the r risk in formed - pe r fo rmance based me thods acceptable to the AHJ can be used to refine the list of nuclear safety systems and equ ipmen t .

The, set o f systems and e q u i p m e n t to be considered for nuclear safety should address, as a m i n i m u m , the following:

(1) Systems and e q u i p m e n t required to place the plant in a safe and stable condi t ion following a fire occurr ing while the plant is at power, or while main ta in ing hot s tandby or hot shutdown. This fire also could result in a loss of off-site power, which would require achieving safe and stable condi t ions us ing power from onsite ac sources (i.e., emergency diesel generators) . This is typically a traditional Appendix R to 10 CFR 50 post-fire safe shutdown analysis.

(2) Systems and e q u i p m e n t required to mainta in shutdown cool ing capability following a fire or iginat ing while the plant is in the shutdown cooling m o d e .

B-2.1 Assumptions (Plant Conditions at Time of Postulated Fire). In addit ion to the assumpt ions in Chapter 2, the following assumpt ions apply to this appendix.

( l ) The plant is in a s tandard l ineup governed by operat ing procedures , opera t ing modes , or administrat ive controls at the onse t o f the fire.

(2) Properly or iented check valves funct ion to prevent reverse flow in process systems.

(3) Normally closed manua l valves (hand-opera ted only) will remain u n d a m a g e d by a fire and can be relied upon for system boundary isolation.

(4) Ins t ruments located in a fire effected area (e.g., RTDs, thermocouples , pressure transmitters , flow transmitters, and mechanical ly l inked r emote / loca l indications) are a s sumed to be damaged unless it can be demons t ra ted othmwise. The in s t rumen t fluid boundary associated with these devices, with the except ion o f soldered fittings, is a s sumed to remain intact.

(5) Piping, check valves, strainers, tanks, mamml valves, heat exchangers , safety relief valves, and pressure vessels are assumed to remain funct ional dur ing and after a fire. The integrity of i n s t rumen t tubing, with the except ion o f soldered fittings, is also expected to be mainta ined, t hough the accuracy of the ins t rument reading can be affected due to heat ing of the process fluid.

B-2.2 Considerations for the Selection of Nuclear Safety Systems and Equipment.

Step 1: System Identification. Based u p o n documen ta t i on of p lant design, risk insights and operat ion, p lant systems required to achieve each of the nuclear safety criteria should be identified.

Step 2: System Inter-Relationships. The selection of systems and the documen ta t i on o f how these systems fulfill the nuclear safety pe r fo rmance criteria should shou ld - be depicted in system-level logic diagrams, fault trees, or some o the r m e t h o d that shows e q u i p m e n t dependenc ies . The documen ta t i on should consider not only the required process systems but also the essential m e c h a n i c a l / e n v i r o n m e n t a l suppor t and essential electrical systems requi red to suppor t the nuclear safety per formance criteria.

Step 3: Equipment Identification. (a) P&IDs/flow diagrams shou ld be used to identify the

e q u i p m e n t in the flowpath and the boundary e q u i p m e n t within the systems that are required to achieve the nuclear safety objectives.

(b) Equ ipmen t that is no t directly in a required system flowpath, bu t whose spur ious opera t ion (undes i red opera t ion) could )revent achieving the nuclear safety objectives should be identified (e.g., boundary valve c o m p o n e n t whose spur ious op en in g could divert flow away from critical equ ipmen t ) . The potential for spur ious opera t ions of e q u i p m e n t should be considered when de t e rmin ing boundary valves and e q u i p m e n t selection.

Loops or bypasses within a system where spur ious operat ion would not result in a loss o f flow or inadequate flow to nuclear safety success paths need not be-considered.

For tanks, all out let lines should be considered for their funct ional requi rements . For lines not required to be fimctional, a m e a n s o f isolation should be included when necessary to prevent unnecessary drawdown of the tank. Tank fill lines should also be cons idered .

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For example, if two normally closed valves in series must spuriously open to result in an unrecoverable condition, then both valves should be identified on the nuclear safety equipment list (NSEL). If positive means is provided to preclude spurious operation of one valve/component for non-high low pressure interface component [such as removing power to one of the two motor-operated valves (MOVs) during normal operation], then consideration of the additional component (the other series valve) is not required.

(c) Careful consideration should be given to equipment that could result in a fire-induced plant transient. The following is guidance on considerations that should be given in the identification of equipment that could result in a fire induced plant transient:

(1) Fire-induced plant initiating events [transients and loss of coolant accidents (LOCAs)]. Transients are defined as anticipated operational occurrences (e.g., inadvertent safety injection actuation, loss of off-site power, over cooling, over filling of steam generators, spurious closure of containment isolation valves, significant loss of safety systems, station blackout, rapid cooldown, etc.) that initiate as a result of fire-induced circuit failures.

(a) Loss of primary sy.stem inventory. The potential for fire initiated spurious actuauon at reactor coolant pressure boundaries that could cause an uncontrolled loss of reactor coolant inventory [e.g., spurious actuation of primary coolant interfaces such as at the reactor head vents, normal and excess letdown at a pressurized water reactor (PWR), main steam relief valves (BWRs)] should be considered.

(b) Rapid cooldown. Transients that could result in an uncontrolled plant cooldown due to spurious operation of boundary valves should be considered. Interaction of plant systems such as steam generator (PWR) atmospheric dump valves, power. operated relief valves, safety relief valves (BWR) feedwater, reactor trip, turbine trip, and main steam isolation should be considered as well.

(c) Uncontrolled primary injection. Transients that could potentially result in an undesired or uncontrolled injection into the reactor coolant system should he assessed. This can include spurious actuation of high-pressure injection sources (i.e., HPCS, RCIC, HPCI, feedwater for BWRs, high-head ECCS pumps for PWRs).

(d ) Electric po~r transients. Transients that could result in a loss of any ac power supplies should be considered. This loss can include spurious breaker actuations, onsite generating capability spurious, starts or failures, or inadvertent paralleling of ac sources due to fire-induced circuit failures.

(d) Equipment that requires support such as cooling water, instrument air, HVAC, motive, and control power should be considered in order to understand component and system inter- relationships and sequential equipment loss impact.

(e) Off-site power can be used as a source of power for nuclear safety equipment. Allequipment required to support the portion of off-site power, relied upon to achieve the. nuclear safety performance criteria shodld also be idenUfied.

Off-site power should conservatively be considered available for those cases where availability of off-site power could adversely impact nuclear safety (i.e., reliance cannot be placed on fire causing a loss of off-site power if the consequences of off-site power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of off-site power to prevent spurious operations.

(f) Instrument sensing lines should be considered for potential inaccurate instrument indications and/or spurious equipment actuations that could occur asa result of an instrument sensing line being exposed to a fire and increased temperatures. Any instrument sensing lines that could prevent the fulfillment of the nuclear safety performance criteria should be identified, associated with the equipment that it could impact, and included in the nuclear safety assessment for review on a fire area basis.

(g) Instrument air piping and components (e.g., accumulators) should be considered for viability during and after the fire in providing the motive force for credited components.

(h) Power supplies, including alternate power supplies, for nuclear safety equipment should be identified . Inter-relationships between power supplies (such as bus-tie capability and alternate power supplies) should also be identified. This information is essential m determining nuclear safety equipment losses due to loss of a power supply.

Step 4: Equipment Interrelationships, The necessary relationships between individual nuclear safety equipment and systems should be understood and documented.

Step 5: Documentation.

(a) The bases for selection and exclusion of nuclear safety Stems and equipment should be documented and maintained. Iculations and analyses that have been previously performed in

support of other nuclear safety objectives (i.e., station blackout, seismic qualification) can be utilized provided the results of these analyses have properly considered the applicability to post-fire nuclear safety.

(b) To develop a nuclear safety equipment list (NSEL) in a consistent and reproducible manner, the following should be considered:

(1) Valves/dampers constituting system boundaries should be included in the NSEL. Normally closed manual valves and properly oriented check valves credited as system boundaries are not required to be listed in the NSEL.

(2) Manual drain, vent, and instrument root valves should not be included in the NSEL.

(3) Valves/dampers in the towpath whose spurious operation could adversely afffct system operation should be included in the NSEL. Manual valves~dampers requiring repositioning during the post-fire shutdown should also be included. Manual valves/dampers/check valves that do not require manual actions during the post-fire shutdown should not be included in the NSEL.

(4) Safety/relief valves provided for equipment and piping protection should not be included. However safety/relief valves providing an active nuclear safety function, such as the steam generator relief valves are exceptions.

(5) Pilot solenoid valves should be listed as separate components in the NSEL The cabling associated with the solenoid valves should he listed either under the solenoid valve or the associated process valve (e.g., the air operated valve) as dictated by the ]~roject implementation procedure.

(6) Pumps, tans, turbines, tanks, heat exchangers, and other equipment should be included on the NSEL.

(7) Instrumentation required for, process monitoring of the nuclear safety systems should be identified.

B-3 Nuclear Safety Circuit Analysis. Circuit analysis for required nuclear safety equipment should be performed and documented. Circuit selection criteria include the following:

(1) Identification of all circuits that can adversely affect nuclear safety, including circuits with auxiliary contacts in the control circuits of nuclear safety equipment

(2) Exclusion of circuits that cannot prevent nuclear safety equipment from performing its nuclear safety function

Other risk informed - performance based methods acceptable to the AHJ.can be used to refine the circuit analysis or its assumptions on circuit failure modes. This may include methods that quantify the credibility or likelihood of circuit failure modes and their correlation to spurious component operation.

The equipment listed in the nuclear safety equipment list should be reviewed to determine all circuits (cables) that could prevent the equipment from performing their nuclear safety performance criteria. This analysis is focused on cables, typically either directly associated with component operation or associated with the component via auxiliary relays and contacts.

Other required circuits are defined as cables whose failure due to fire damage can affect nuclear safety capability. The following are three types of other required circuits:

(a) $tmrious Actuations. The concern is fire damage to a cable whose ~ail~re could result in the spurious operation or mal- , operation of equipment that could affect nuclear safety. Identification of cables that could cause spurious component operation is an integral part of the circuit analysis methodology described in Section B-3.1

(b) Common Power Supply. The concern is fire damage to a cable that could result in a loss of power to a power source required to support nuclear safety equipment. This damage could occur if the upstream breaker/fuse for a nuclear safety power supply is not properly coordinated with a downstream breaker/fuse for the cable that is damaged. This is discussed in Section B-3.2.

(c) Common Enclosure. The concern is fire damage to a cable whose failure could propagate to other nuclear safety cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) or the fire could somehow propagate along the cable and impact nuclear safety equipment. This situation is discussed in Sectton B-3.3.

13-3.1 Deflnidotm. The following circuit failure modes should be considered in the evaluation.

(a) Open Circuit. A condition that is experienced when an individual conductor within a cable loses electrical continuity.

(b) Shorts-to-ground. A condition that is experienced when an individual conductor comes in electrical contact with a grounded conducting device, such as a cable tray, conduit, or metal housing.

303


(c) Hot Short. A condi t ion that is exper ienced when individual conductors o f the same or different cables come in contact with each other.

B-3.2 Considerations for Performing Nuclear Safety Circuit Analysis. The following describes a process and the guidel ines for pe r fo rming the nuclear safety circuit analysis.

Step I: Equipment Parameter l'np~Jt. All e q u i p m e n t requi r ing circuit analysis and their funct ional '

pa ramete rs (see Step 5, Section B-2) should be identif ied in the NSEL (see Section B-2). The funct ional parameters include normal and desired positions and failure pos i t ion / s ta tus upon loss o f power and air (if applicable). These parameters establish the framework for circuit analysis.

Step 2: Electrical D'rawi~lg Review. For each c o m p o n e n t to be analyzed, applicable e lementary

diagrams, single-line drawings, connec t ion diagrams, i n s t r u m e n t loop drawings, and vendor drawings should be reviewed as necessary to identify all circuits connec t ed to the componen t , including control cables associated with the c o m p o n e n t via auxiliary contacts in the c o m p o n e n t ' s control circuit.

Step 3: Considerations for Circuit Identification. Each circuit should be evaluated to de te rmine which cables are

necessary to suppor t the nuclear safety funct ion of the componen t . If a f ire-induced circuit failure of the cable can place the c o m p o n e n t in a pos i t ion /cond i t ion o the r than any desired pos i t ion /condi t ion necessary by that c o m p o n e n t in any mode of operat ion required to achieve the nuclear safety pe r fo rmance criteria, the cable should be identif ied as a requi red cable. If a f ire-induced circuit failure o f the cable canno t spuriously reposit ion or prevent the desired opera t ion o f the componen t , it is no t a required cable.

The following provides general evaluation guidel ines for identification of required circuits:

(a) For spur ious actuat ions and signals, all possible failure states, inc luding all possible consequent ia l damage , should be evaluated - - that is, the c o m p o n e n t could be energized or de-energized by one or more f i re- induced circuit failure modes (e.g., ho t shorts, shorts- to-ground, and open circuits). Therefore , valves could fail open or closed; p u m p s could fail r u n n i n g or not runn ing ; and electrical dis tr ibut ion breakers could fail open or c l o s e d . Multiple spur ious actuat ions or signals or ig inat ing f rom fire- induced circuit failures could occur as the result of a fire in a given fire area.

(b ) Dur ing the process o f identifying required nuclear safety circuits, a hot shor t circuit failure on the appropr ia te conductor (s ) should conservatively be a s sumed to cause spur ious e q u i p m e n t operat ion. This will ensure that a comprehens ive populat ion of circuitry is evaluated for the impact of fire on mee t ing the nuclear safety pe r fo rmance criteria.

(c) A cable fault is a ma in ta ined condi t ion that exists for the dura t ion o f the fire until an action has been taken to isolate the given circuit f rom the fire area or o ther actions have been taken to negate the effects of the spur ious actuation. It c anno t be a s sumed that the fault condi t ion should te rminate to a g r o u n d or o p e n circuit condi t ion that should stop the spur ious actuation. Generally upon clearing o f a fault, the e q u i p m e n t would re turn to its desired state (if the de-energized state is the safe state o f the safe shutdown c o m p o n e n t ) . An example is dc solenoid valve or air-

o p e r a t e d valve (AOV) with dc solenoid pilot that would fail to its safe posit ion on a loss o f control power to the solenoid. This would not be applicable to an MOV, which, if spuriously operated, would remain in that posit ion following a loss of control power.

(d) When evaluat ing a c o m p o n e n t failure (e.g., required nuclear safety c o m p o n e n t or a related c o m p o n e n t ) , all circuits that are associated with the c o m p o n e n t ' s opera t ion and all t hose circuits whose failure could affect the c o m p o n e n t ' s opera t ion should be considered as required.

(e) A single ho t shor t f rom an external source could affect any and all conductors , one at a t ime, within a cable that is required or is associated with a requi red nuclear safety c o m p o n e n t . In addit ion to evaluating the affects o f a hot shor t circuit failure, these failures could occur, in con junc t ion with o the r s imul taneous individual or mult iple f i re- induced circuit failures (i.e., shorts-to- g round , open circuits, internal conduc to r to conduc t o r shorts) on o the r conductors associated with the circuit or cable o f concern . For h igh / low pressure interfaces, properly al igned three phase and proper polarity dc ho t shorts mus t be postulated to occur and these circuits mus t be included in the analysis.

(f) For mul t i -conduc tor cables, energized conduc tor ( s ) could shor t circuit individual or mult iple conductors within the cable that is required or is associated with a required nuclear safety c o m p o n e n t .

(g) Plant-specific design features can preclude certain circuit failures f rom occurring. For example , the use o f g rounded , metallic, a rmored cable or dedicated condui t would preclude external ho t shorts from fur ther considerat ion. However, mult iple g r o u n d faults migh t still energize conductors within a g r o u n d e d condui t , shield or a rmor if those conductors are associated with u n g r o u n d e d circuits. See Figure B-3.2(g) for an example involving mult iple shorts to ground. For ease o f analysis when analyzing an u n g r o u n d e d DC circuit for the effects o f a shor t to g round , it should be assumed that an existing g r o u n d fault from the same power source is present .

Step 4: Documentation. The circuit analysis for each c o m p o n e n t analyzed should clearly

d o c u m e n t the des ignated nuclear safety cables and, more important ly, d o c u m e n t and justify excluded cables and circuits. The rationale for selection or exclusion of a cable should also be d o c u m e n t e d . The circuit analysis should list required power suppl ies and identify those cables associated that mus t remain funct ional following a control t ransfer (i.e., valve control cables that remain funct ional following control t ransfer from a r emote shutdown panel) . Any use o f r isk-informed me thods to reduce the scope o f circuit analysis should be adequately documen ted .

Specific Circuit Failure Scenarios The following are examples of typical circuit failure modes

(simplified circuits figure have been used for illustrative purposes only).

(a) Low-Voltage dc (1V dc-48V de). These circuits are typically i n s t rumen t signal cables for moni tor ing , protect ion systems, or control valves circuits (sensor to I / P converter) . Shielded, g r o u n d e d signal are cables typically used. Note that a conductor- to-conductor shor t on i n s t rumen t cable or an in termedia te resistance g r o u n d can result in a false in s t rumen t signal (same failure consequences as cable ho t short , a f ire-induced high signal). See Figure B-3.2(a) for an example of this failure mode .

( ~ Indicator

Protect ion*n-L~ wJ' i " ,24V Curren~ A ~ Short

Transmitter

Note: An internal short at point A could result in an erroneous instrument signal, which could cause erratic indication or related equipment mal- function such as protection system actuation or actuation of system interlocks.

Figure B-3.2(a) Instrumentation circuits.

Many plant protect ion circuits have logic circuitry that require mult iple inpu t signals in o rder to actuate (i.e., one out of two taken twice, two ou t of three, etc.). In these instances multil~le failures on i n s t rumen t inpu t signal cables (or in some specinc designs, mult iple conduc tors within the same cable) could cause a f i re- induced protect ion system actuat ion.

(b) Medium Voltage Single Phase ac (l l SV ac-220V ac). These circuits are typically used for secondary side control t ransformers o f Motor Control Cente r (MCC) loads (e.g., MOV control, fan and small p u m p controls, ac solenoid opera ted valves and dampe r s ) .

For g r o u n d e d ac circuits, a ho t shor t "source" can be from any energized circuit internal or external . For u n g r o u n d e d ac, external and internal ho t short "source" mus t be f rom same the same power supply.

See Figures B-3.2(b) t h rough B-3.2(e) for examples o f open circuits, shorts- to-ground, and hot shorts on typical g r o u n d e d 120V ac control circuit.

304


Valve open/close

- I I . / y - i i - A

o

> 0

±


Notes: 1. An open circuit at points A and B will not affect the valve position if the control switch is open. However, it will not permit the operation of the valve.

2. An open circuit at points A or B will de-energize the valve and change its position if the control switch is closed. In addition, the open circuit will not permit the operation of the valve.

Figure B-3.2(b) Open circuit.

Valve open/close

- I I . - I I -

o l

;I o,I Solenoid valve

(energize to open)

A

>-

. B

Notes: 1. Short-to-ground at point A will not affect the valve position if the control switch is open.

2. Short-to-ground at point A will affect the valve position if the control switch is closed by de-energizing the solenoid.

3. Short-to-ground at point B will not ~ffect the valve position irrespective of the control switch position.

Figure B-3.2(c) Short-to-ground.

PS-1 " R

I ~ Solenoid valve . ~ . ~ (energize to open)

Notes: 1, A hot short at point A (from internal or external source) and an internal short inside the cable connecting to PS.1 will energize the valve and change its position.

2. A hot short at point B will energize the valve and change its position.

3. A hot short at C and D (from an external source) need not be postulated as fire-induOed circuit failure for this valve, except for high-low pressure interface valves.

Figure B-3.2(d) Hot short example (external cable).

Valve open/close

>

2_

PS-1 :1'"- ,

Closes when [ pressurem / ,


Note: Internal short inside the cable connecting to PS-1 and a short across the valve's control switch at point A will energize the valve and change its position.

Figure B-3.2(e) Hot short example (conductor-to-conductor within same cable).

(c) Medium Voltage dc (125V dc to 250V dc). These circuits are typically used for breaker controls for largepumps, switchgear, diesel generators, de distribution, controls for solenoid-actuated air-operated valves, solenoid valves, dc MOVs, and so forth.

For an external hot short, the hot short "source" (other than a 2 conductor proper polarity hot short) must be from the same dc source (battery) as the dc circuit under evaluation in order to provide a complete circuit path. ."

For high-low pressure interface valves, removing power at dc panel during normal operation does not eliminate two conductor proper polarity dc hot shorts from causing spurious actuation. See Figure B-3.2 (f).

305

N F P A 805 -- N o v e m b e r 2000 R O C

/_ Va,veopen/c,o e A ~ Solenoid

'U 1g25VndCd ) B . ~ ( (energized to operl)

T Valve open/close

Multiple shorts-to-ground on an ungrounded dc system on circuits from the same battery (or on ungrounded ac circuits from the same transformer) could spuriously energize a normally de-energized circuit. See Figures B-3.2(g) and B-3.2(h). In addition, a hot short on an ungrounded dc circuit could also result in a protective device actuation (blowing a fuse or tripping a breaker) if a positive conductor shorts to a negative conductor from the same dc source or vice versa.

Figure B-3.2(f) Two-conductor proper polarity dc hot short example (solenoid valve or solenoid pilot valve for AOV).

Multiple shor ts- to-ground on an u n g r o u n d e d dc system on circuits f rom the same battery (or on u n g r o u n d e d ac circuits f rom the same t ransformer) could spuriously energize a normally de- energized circuit. See Figures B-3.2(g) and B-3.2(h). "In addit ion, a hot shor t on an u n g r o u n d e d dc circuit could also result in" a protective device ac tuat ion (blowing a fuse or t r ipping a breaker) if a positive conduc t o r shorts to a negative conduc to r f rom the same dc source or vice versa.

CK'TI. B KT.A.

(+) " ~ [ I E 2 ]

i Hot short Valve example

°per~cl°se T H}"ff~ ,ho"~'~ A [ ~ ~ sot irce

125V dc '1 ~ ~ (sat1 le dc { (ungrounded) / , . . ~ sot.rce)

~ ' Sotenoia vatve ",-.~.-~, ,g~ (energize to open)

(+)

~ ( '~"{ I~ ~ a.

t v ,ve _ ................ open/close - '

CKT. B cK'r. A,{ 125V dc (ungrounded) A ~ .

"(Se¢/eerng°~d~alopen ) snorts-to- i ,pen) .¢~n~Sun d

I L(_zL_tE~ ~ ~-example

Notes: 1. A short at point A (from an internal or external source) will energize the valve and change its position (top figure).

2. A short-to-ground at point A and a short to ground at point B will energize the valve and change its position (bottom figure),

Figure B-3.2(g) U n g r o u n d e d sys tem -- mult ip le shorts- to-ground.

(+)

C

CKT. B KT.A<

Notes: 1. A short-to-ground at point A and a short-to-ground at point B will not affect the valve position.

2. A short-to-ground at point A and a ground at point B and a simultaneous hot short at point C need not be postulated as a fire-induced failure for this valve except for high-low pressure interface valves.

Figure B-3.2(h)

O1 02 03

Breaker MCC (open or closed)

MCC

O

430 V ac ! 1 I MCC

~- ] control ~" _ ! power

. , I . L . . . . . . . . . J

142

~, ~ "~"43 (TOLS)

02 03

~A 6B

Valve motor

O1

C

Note: Hot shorts at points A, B, and C (from an external source, such as adjacent power cables in the cable tray or raceway) will energize the valve motor and change its position. This type of fire-induced circuit failure needs only be postulated for a high-low pressure interface valve.

Figure B-3.2(i) Three-phase ac hot short example (MOV).

It should be noted that for high-low pressure interface valves, r emoving power at the moto r control cen te r (MCC) du r ing normal power opera t ions does not e l iminate the potential three- p h a s e p r o p e r phase ac hot shorts involving the power cable from the MCC to the valve could cause spur ious operat ion of the valve.

B-3.3 Other Required Circuits (Formerly Known as Associated Circuits). This sect ion conta ins a descr ipt ion o f addi t ional circuits that should be considered for their impact on the ability to achieve the nuclear safety criteria. These circuits are divided into the following two categories: l) c o m m o n power supply and 2) c o m m o n enclosure . These categories are descr ibed in the

306

• N F P A 8 0 5 m N o v e m b e r 2 0 0 0 R O C

following paragraphs. Note that circuits required due to spurious operat ion considerations are addressed in Sec t ion B-2.2.

C om m on Power Supply Issues. The concern is fire damage to a cable that could result in a loss of power to a power source required to support nuclear safety equipment . This situation could occur if the upstream breaker / fuse for a nuclear safety

~ ower supply is not properly coordinated with a downstream reaker / fuse for the cable that is damaged. Circuits associated by

common power supply are circuits that are powered from nuclear safety power supplies whose overcurrent protect ion device is not properly coordinated with the overcurrent protect ion of the power supply feeder [see Figure B-3.3(a)].

~//////~////////////////~

s;utmd~)ile ~ I pureed ~i~re i~Jtiorl E ~ shutd°wrl y pump B

• a r e a I F i r e a r e a II Redundant pumps

Figure B-3.3(a) Common power supply issues.

In Figure B-3.3(a), the postulated fire could render nuclear safety ump B inoperable. Fire-induced electrical faults affecting pump or its power cable could cause breaker 1 to trip, thereby

render ing p u m p A inoperable if the two respective overcurrent isolation devices (breaker 2 and breaker 1) are not properly coordinated. This situation results in the interrupt ion of power to redundan t nuclear safety pumps.

Common Enclosure Issues. Circuits, which share enclosures with nuclear safety equ ipment circuits, must be analyzed to de termine their effect on redundan t nuclear safety equipment . The concern is that the effects o f the fire can extend outside of the immediate area. by means of fire-induced faults on inadequately protected cables and could cause an overcurrent condi t ion resulting in secondary ignition. This sitaution could result in remote damage that could disable the redundant train of equipment .

A type of common enclosure failure mode is that o f secondary ignition. As shown in Figure B-3.3(b), the fire at ins t rument B induces a fault on a cable that is not adequately electrically protected. The fault then results in the ignition of a secondar~ fire

the vicinity of the ins t rument A cables due to an overcurrent condit ion resulting in cable jacket ignition. Again, the result could be the loss of the two redundan t trains.

A special type of common enclosure issue involves current transformers. Current transformers are used throughout the electrical distribution system to moni tor bus current and provide differential protect ion: The current transformers are physically located on the electrical conduc tor (i.e., bus bar) and provide a signal through their secondary winding that is proport ional to the current flowing through the conductor . Current transformers are designed to transform high primary current into low secondary current• Theoretically, the secondary voltage is as high as required to maintain a constant primary-to-secondary current ratio. An open ing in the secondary circuit causes excessively high voltages in the current t ransformer secondary circuit in an at tempt to maintain this ratio, which can result in an ignition of the t ransformer materials.

One type of current t ransformer circuit scheme is provided in Figure B-3.3(c). A fire in the control room could cause a fire- induced open ing in the secondary circuit of the current transformers, causing an electrical fault at the current transformers in both nuclear safety buses A and B. This situation could result in a fire within each bus, thereby disabling both buses and the redundan t nuclear safety pumps powered from them.

Potentia secondary due to cat ignition : i re a r e a II ~

Area of =on enclosure

a t i o n

instruments Figure B~3.3(b) Concern common enclosure issues - - secondary ignition.

307

Switchgear room A

.) A

Control room Fire location

Switchgear room B

Safe shutdown Bus B


~ , ~ shutdown components - -

Figure B-3.3(c) Common enclosure issues--current transformer secondaries.

B-3.3.1 Considera t ions W h e n Evaluating Common Power Supply Issues. The nuclear safety e q u i p m e n t list will be used to identify the electrical distr ibution buses, switchgear, load centers , and panels required to provide electrical power to nuc lea r safety e q m p m e n t . T h e fol lowing steps desc r ibe the process r e c o m m e n d e d to evaluate those circui ts (nuc lea r and non -nuc l ea r safety) that share a corn m o n power supply with nuc lear safety equ ipmen t .

Step 1: ldent!fication of l~quired Power Supplies. A listing of all the electrical distr ibution system buses, switchgear,

load centers, panels, and so forth, should be prepared utilizing the nuclear safety e q u i p m e n t list (Section B-2.2).

Step 2: Evaluation of Electrical Protective Device and CooMination. Each load on the nuclear safety power supplies for circuit

coordinat ion should be reviewed per accepted design practices. A list of the circuits that are not properly coordinated should be prepared.

Step 3: Disposition and Documentatiot~. , The results of the c o m m o n power supply issues should be ' d o c u m e n t e d . For cases in which coord ina t ion between series protective devices canno t be demons t ra ted , a c o m m o n power supply o ther required circuit shonld be a s sumed . to exist. These circuits should be disposi t ioned by one o f the following means:

(1) Demonst ra te coordinat ion by ref ining the available short circuit cu r r en t a n d / o r trip device characteristics.

(2) Identify protective device setpoint changes ( including changes in fuse size a n d / o r clearing characteristics) that should establish coordinat ion .

(3) Incorporate the cables of concern into the nuclear safety analysis as required circuits for the affected power supply. This p rocedure should ensure that, on a fire area basis, the impact of c o m m o n power supply circuits is evahmted.

B-3.3.2 Comideralions When Evaluating Common Enclosure Issues .

Step 1: Common E~closu're - - Seconda U lguitwn. As descr ibed previously, s e c m f l a r y ignition has the potential to

be a concern due to f i re - induced electrical faults on inadequately protec ted cable r e su l f ing in secondary ign i t ion . This conce rn can be addressed by evaluating the plant electrical des ign criteria to ensure t h a t a d e q u a t e overcmxent pro tec t ion for all conduc tors exists t h roughou t the plant .

Step 2: Current 7)ansformer Secondary lguttion. Current t rans formers that a re consm~cted such that an o p e n

secondary circuit could cause ignition o f the t ransform er shou ld

be cons idered o ther requ i red circuits o f the power supply in which they are located. Cur ren t t ransform ers tha t a re suscept ible to igni lion d u e to o p e n secondary windings and have secondary circuits ex tending ou t s ide the fire a rea that are nbt isolated by t ransducers shou ld have their circuits included in the nuclear safety assessm ent.

Step 3: Disposition and Documentation. The results of the c o m m o n enclosure issues should be

d o c u m e n t e d and mainta ined. Calculations and analyses that have been pe r fo rmed in suppor t o f o ther issues can be utilized provided the results o f these analyses have properly considered the applicability to post-fire nuc lear safety.

B-4 Nuclear Safety Equipment and Cable Location. The purpose o f this step is to identify the location of the nuclear safety e q u i p m e n t and cables identified in Sections B-2 and B-3. EcLuipment and cables should be located utilizing available in lormat ion from exist ing databases, plant layout drawings, and, if necessary, p lant walkdowns. Equ ipment and cables should be located by the smallest des ignator (room, fire zone, or fire area) for ease o f analysis.

The cable and raceway database (or the hard copy drawings) establish the relat ionship between cables and their routing. Based on this informat ion, supp lemen ted as necessary by plant walkdowns, the location of each required circuit can be identified. The process usually entails the following process:

(1) Cable-to-raceway relat ionship (what condui t , cable trays, j u n c t i o n boxes, pul lboxes that a cable is routed th rough)

(2) Raceway-to-fire z o n e / r o o m relat ionship (what fire zones, rooms, or fire areas that a raceway is routed th rough)

(3) Cable endpo in t location (fire zones, rooms, or sub-fire areas, where a cable terminates; this can be necessary since cable endpo in t s can exist in plant areas with no raceways conta in ing the cable - - i.e., a cable te rminat ing in the control room, coming up t h rough the floor o f a cable spreading room)

(4) Fire z o n e / r o o m to fire area relat ionship (what sub-fire areas compr ise a fire area)

B-5 Fire Area Assessment . In Sections B-2, B-3, and B-4, the in t e rdependenc ie s o f nuclear safety systems, equ ipmen t , and cables have been established. An assessment for each fire area should be p e r f o r m e d in o rder to demons t ra te ach ievement of the nuclear safety pe r fo rmance criteria. The nuclear safety e q u i p m e n t and their inter-relat ionships should be analyzed for each area to ensu re that success path(s) are available based upon the postulated e q u i p m e n t and cable losses in that fire area. G the r risk in fo rmed - pe r fo rmance based me thods acceptable to the AHJ can be used to refine the fire area assessments and the assumpt ions made regarding the impact on a success path. A success path should be identified that ensures each o f the nuclear safety pe r fo rmance criteria is met.

B-5.1 Methodology. Step 1: Ident!fication of Affected ,S~),stems attd Components. All unpro tec ted cables and e q u i p m e n t within a fire area could be

affected by the fire. This does not imply that the fire ins tantaneously spreads t h roughou t the fire area, but rather is i n t ended as a conservative me thod to address the fact that, for this analysis, ne i ther the fire size nor the fire intensity is rigorously de te rmined . As a m i n i m u m , this should include the following:

(a) Nuclear safety e q u i p m e n t affected by ei ther the equ ipment , its associated cables, or associated i n s t r u m e n t tub ing located in the area.

(b) Equ ipmen t affected by loss o f a suppor t system (cooling water, ventl'lation) or loss o f power supply

Using the e q u i p m e n t inter-relat ionship tool (logic diagrams, fault trees, etc.), identify the e q u i p m e n t affected by a fire in the area and the reason for the e q u i p m e n t loss (i.e., e q u i p m e n t located in the area, cable located in the area, power supply lost, etc.)

Step 2: Success Path Determination. The following guidl ines should be considered in the

de te rmina t ion o f success path (s): (a) Suppor t systems should be reviewed first in order to assess

the impact on the systems being suppor ted . For example , it is r e c o m m e n d e d that the impact on the electric power distr ibution system be reviewed first. This review is pe r fo rmed so that an unnecessary a m o u n t of ' t ime is no t spent analyzing process systems (i.e., charging, auxiliary feedwater, residual heat removal) that migh t not have necessary power to suppor t the specific train of c o m p o n e n t s to suppor t safe shutdown.

(b) Similarly, cool ing systems such as se~wice water and c o m p o n e n t cool ing water should be reviewed for the systems that they support .

(c) The impact o f a part icular fire on process mon i to r ing ins t rumenta t ion should be reviewed prior to assessing o ther system

308

N F P A 8 0 5 m N o v e m b e r 2 0 0 0 R O C

availability, to ensure that process moni tor ing instrumentat ion is available for the equ ipment relied upon to meet the success path.

(d) Note that some equ ipment can perform more than one function, such as RHR isolation valves that are normally closed during plant operation at power and dur ing the initial phases of plant cooidown but are required to be open to support RHR operation and the safety function of long-term decay heat removal.

(e) Exclusionary analysis. (1) An acceptable alternative to the methods described above is

the performance of an exclusionary analysis. This analysis would assure that cables and componen t s for at least one ~uccess path are not located in the area.

(2) In taking the performance based approach, factors discussed in (d) above should again be considered in the event equ ipment (or its cables) required for a success path to achieve and maintain the nuclear safety performance criteria are in the fire area o f concern.

(3) Although an exclusionary analysis uses a different approach than the step 2 success path determinat ion process described in items (a) through (f), all o f the fundamentals , such as componen t inter-relationships, recovery action feasibility, and so forth, should be considered. Potential losses o f equ ipment due to common power supply, common ' enclosure concerns, and spurious actuations and signals should also be addressed. Success paths relied upon for achieving the nuclear safety performance criteria using the exclusionary analysis me thod should be clearly documented for each fire area in order to ensure future changes do not invalidate the results o f the analysis.

(f) The basis for successful demonst ra t ion of the nuclear safety performance criteria for a fire in each fire area should be documented for equ ipment potentially affected by a fire in the area. This documenta t ion is particularly important when redundant nuclear safety equ ipmen t is located in the area and " reliance is placed on physical separation, fire protect ion systems and features, or recovery actions, in o rde r to achieve and maintain the nuclear safety performance criteria.

B-5.2 Methodology Success Path Resolution Considerations. (a) The magnitude, duration, o r complexity of a fire cannot be

foreseen to the extent o f predict ing the t iming and quantity of fire induced failures. Nuclear safety circuit analysis is not in tended to be performed at the level o f a failure modes and effects analysis since it is not conceivable to address every combinat ion of failures. Rather, for all potential spurious operat ions in any fire area, focus should be on assessing each potential spurious operat ion and mitigating the effects o f each individually. Multiple spurious actuations or signals originating from fire-induced circuit failures could occur as the. result o f a gwen fire. The simultaneous equipment or componen t mal-operations resulting from fire induced failures, unless, the circuit failure affects multiple components , is no t expected to initially occur. However, as the fire propagates, any and all spurious equ ipmen t or componen t actuations, if not protected or properly mitigated in a timely manner , could occur. Spurious actuanons or signals that can prevent a required c o m p o n e n t from accomplishing its nuclear safety function must be appropriately mitigated by fire protection features.

(b) An assumption of only a single spurious operat ion without opera tor intervention (i.e., having two normally closed MOVs in series with cables routed through an area, and assuming only one of the valves could spuriously open) should not be relied upon for ensuring a success path remains available. Therefore, in identifying the mitigating action for each potential spurious operat ion in any given fire area, an assumption such as that stated above should not be relied upon to mitigate the effects o f one spurious operat ion while ignoring the effects o f another potential spurious operat ion.

(c) Where a single f ire 'can impact the cables for High-low pressure interface valves in series, the potential for valves to spuriously operate simultaneously should be considered. Removing power to two or more normally 'closed high-low pressure interface valves in series dur ing normal operat ion (which reduces credible spurious operat ions to multiple three-phase ac hot shorts or multiple p roper polarity dc hot shorts on multiple valves) is an acceptable me thod of ensuring RCS integrity wimout additional analysis o r fire protect ion features. This criteria applies to all fire areas, including the control room, and to all circuits regardless of whether or not they can be isolated from the control room by the actuation of an isolation transfer switch.

(d) The performance-based approach should consider the fire protect ion systems and features of the room and what effects the fire scenarios would have on the nuclear safety equ ipment within the area under consideration. \

(e) Recovery actions can be per formed as part pf a performance- based, risk-informed approach subject to the limitations of Chapter 4 of the standard to mitigate a spurious actuation or achieve and maintain a nuclear safety performance criteria. For the equ ipment requiring manual actions, information regarding the fire areas requiring the action, the fire area in which .the action is performed and the time constraints to perform the actions should be obtained to assess the feasibility o f the proposed action.

(1) The proposed recovery actions should be verified in the field to ensure the act ion can be physically per formed under the conditions expected during and after the fire event:

(2) When recoveryactions are necessary in the fire area under consideration, the analysis should demonst ra te that the area is tenable for the actions to be performed and that fire or fire suppressant damage will not prevent the manual action from being performed.

(3) The lighting should be evaluated to ensure sufficient lighting is avadable to perform the in tended action.

(4) Walk-through of operations guidance (modified, as necessary, based on the analysis), should be conducted to de termine if adequate manpower is available to perform the potential recovery actions within the time constraints (before an unrecoverable condit ion is reached).

(5) The communicat ions system should be evaluated to de te rmine the availability o f communicat ion, where required for coordinat ion of recovery actions.

(6) Evaluations for all actions, which require traversing through the fire area or an action in the area of the fire, should be" per formed to de termine acceptability.

(7) Sufficient time to travel to each action location and perform the action should exist. The action should be capable of being identified and performed in the time required to support the associated shutdown function(s) such that an unrecoverable condit ion does not occur. Previous action locations should be :onsidered when sequential actions are required.

(8) There should be a sufficient number of essential personnel o perform all o f the required actions in the times required, based

on the minimum shift staffing. The use of essential personnel to perform actions should not interfere with any collateral fire " brigade or control room duties.

(9) Any tools, equ ipment or keys required for the action should be available and accessible. This includes considerat ion o f SCBA and personnel protective equ ipment if required.

B-6 Special Considerations for Non-Power Operational Modes. In order to assess the impact o f fire originating when the plant is in a shutdown mode, the same basic methodology utilized for the nuclear capability safety assessment is used when assessing the impact o f fire on nuclear safety during non-power operational modes. The set o f systems and equ ipment are those required to support maintaining shutdown conditions. Additionally, the criteria for satisfying the performance criteria while shut down can be more qualitative in nature and have less reliance on pe rmanen t design features. For example, existing licensing basis might have allowed redundan t success paths required for long term cooling to be damaged due to a single fire and subsequently repaired. For a fire originating while in a shutdown mode , this can result in a loss of long-term decay heat removal capability. This insight should be factored into outage planning, by limiting or restricting work activities in areas of vulnerability, ensuring operability of detect ion and suppression systems and control o f transient combustible loading.

Shutdown or fuel pool cooling operat ions are categorized as ei ther low or high risk evolutions. Fire protect ion requirements for equ ipment needed or credited for these operat ions would depend upon the categorization of the evolution the equ ipment supports. The categorization of the various shutdown or fuel pool cooling plant operational states (POSs) should be per formed to de termine whether the POS is considered as a high or low risk evolution. Industry guidance, such as NUMARC 91-06, Guiddines for Industry Actions to Assess Shutdown Management, can be used in this determinat ion.

In general, POSs at or near the ~isk level o f full power operations are considered low risk evolutions. High risk evolutions for shutdown would include all POSs where the fuel in the reactor and residual heat removal (RHR)/shutdown cooling is not being used (i.e., for a PWR this would be modes 3 and 4, when steam genera tor cooling is being used. In addition, high risk evolutions would include RFIR POSs where reactor water level is low and time to boil is short. POSs where the water level is high and time to boil is long are considered low risk evolutions.

An example categorization for a PWR would be: High risk evolutions: All modes 2 through 5

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Mode 6 with water level below reactor flange Low risk evolutions: Mode 6 with water level above the reactor flange fuel in the fuel pool, core loading or un load ing

The following is general gu idance /d i scuss ion on the applicability of the major nuc lear safety capability assessment steps to non- power operat ional modes , shu tdown cooling, or spen t fuel pool cooling.

The same methodology used for fires or iginat ing at power should be used for e q u i p m e n t required in high risk evolutions (see Section B-2). For shu tdown cooling, many of the systems and e q u i p m e n t analyzed to main ta in safe and stable condi t ions (cold shu tdown) for non-power operat ional (fuel coolant t empera tu re <200EF) condi t ions shou ld be sufficient. For spen t fuel pool cooling, any systems, equ ipmen t , and associated ins t rumenta t ion mus t be identified and their inter-relat ionships identified in o rder to prop~erl, y assess their..susceptibility, to fire. damage in. h igh risk . evoluuons. Any a d d m o n a l 6qu lpment , ( r ac ludmg m s t r u m e n t a u o n for process mon i to r ing when the plant is in an abnormal condit ion) should be identified to s u p p l e m e n t the cold shu tdown cool ing systems and equ ipmen t . Power sources necessary to suppor t the shutdown cooling and spen t fuel cool ing should be identified, similar to the m e t h o d used for power operat ions.

Nuclear Safety Capabili ty Circuit Analysis. T h e same methodo logy used to evaluate f ire-induced circuit failure for fires or ig ina t ing at power should be followed (see Section. B-3).

Nuclear Safety Equipment and Cable Location and Identif ication. The same methodology used to evaluate fire- induced circuit failure for fires or iginat ing at power should be followed (see Section B-4).

Fire Area Assessment . Following the identif ication o f systems and equ ipmen t , a review of allowed and actual p lant operat ional modes and allowed outage times and practices shou ld be per formed. This review will help to identify areas o f vulnerability to ensure that the nuclear safety per formance criteria is me t for fires or iginat ing du r ing these modes .

The nuclear capability assessment for non-power operat ional modes will be per formance-based and shou ld clearly demons t r a t e that the nuclear safety pe r fo rmance criteria is adequately satisfied. This capability assessment should consist o f a review of the p lant ' s TS and administrat ive control practices, outage p l ann ing and assessment processes, and discussions with plant outage and operat ions staff. A review of fire protect ion system operabili ty r equ i rements and t ransient combust ib le control p rograms should be pe r fo rmed to identify practices du r ing shutdown modes . Compl iance strategies for achieving the nuclear safety pe r fo rmance criteria can include one or more o f the following:

(1) Verifying vulnerable area free o f in tervening combust ib les while on shu tdown cooling

(2) Providing fire patrols at periodic intervals when in periods o f increased vulnerability due to p,ostulated e q u i p m e n t out of service and physical location o f e q u i p m e n t and cables

(3) Staging o f backup equ ipmen t , repair capabilities, or cont ingency plans to account for increased vulnerability

(4) Prohibi t ion or l imitation o f work in vulnerable areas du r ing periods of increased vulnerability

(5) Verification of operable detect ion a n d / o r suppress ion in the vulnerable p lant areas du r ing periods of increased vulnerability

(6) Verifying that the quanti ty o f combust ible materials in the area remains below t h e h e a t release level that would chal lenge e q u i p m e n t requi red to main ta in shutdown cooling. COMMITTEE STATEMENT: The changes were editorial. N U M B E R OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMM IT T E E ACTION:

AFFIRMATIVE: 23 NEGATIVE: 1 N O T RETURNED: 2 Bigglns, Vieira

EXPLANATION O F NEGATIVE: CONNELL: Note 3 of Section B-3.2 Step 4 (b) for Figure B-3.2(d)

states: "A hot shor t at C and D (from an external source) need not be postula ted as f ire-induced circuit failure for this valve, except for high-low pressure interface valves." This posit ion is no t in accordance with the cu r r en t NRC guidance, and is no t suppor ted by the cur rendy available test data which indicates that this circuit fault should be cons idered for all valves. This s t a tement also appears to conflict with the s ta tements in Section B-3.2 Step 3 (e) and (f) o f the draft s tandard.

(Log #8) 805- 108 - (B-l): Accept in Principle SUBMITTER: James E. Lechner , Nebraska Public Power District C O M M E N T O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text to read as follows:

"...sufficiertt a)'atema and e q u i p m e n t remains. . ." SUBSTANTIATION: Implies that r edundan t systems may be necessary for a given success and is inconsis tent with intent of sect ion. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commi t tee Action on C o m m e n t 805-107 (Log #124). NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #9) 805- 109 - (B-2.2 Step 3(b)): Reject SUBMITTER: James E. Lechner , Nebraska Public Power District C O M M E N T O N PROPOSAL NO: 805404 RECOMMENDATION: Revise text to read as follows:

For example , if two normally closed valves in series must spuriously open to result in an unrecoverable, then both-at least one o f the valves should be identified. S U B S T A N T I A T I O N : To be consis tent with the BWROG Safe Shutdown Methodology documen t . Combina t ions of e q u i p m e n t spuriously opera t ing are cons idered probabilistically remote. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: Both values for e q u i p m e n t identification process need to be included. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #49) 805- 110 - (B-2-2, Step 3, (5),(b) 2nd Paragraph): Reject SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute CO MMENT O N PROPOSAL NO: 805404 R E C O M M E N D A T I O N : Delete the following the sentence beginning , "Off-site power should conservatively..." SUBSTANTIATION: This sentence is inconsis tent with the first sen tence in paragraph (g). If analysis can suppor t shows offsite power is unaffec ted and available to suppor t safety equ ipment , it can also show offsite power is unavailable u n d e r o ther c i rcumstances where its availability would have an adverse impact. Making a conservative assumpt ion o f availability is no t necessary when a realistic analysis can demons t ra te the opposite. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: No analysis can guarantee an off site loss o f power. NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N C O M M r r T E E ACTION:


(Log #81) 805- 111 - (B-2-2 Step 3 (b)) : Reject S U B M I T r F ~ : Robert C. Daley, Entergy Operat ions , Inc. C O M M E N T O N PROPOSAL NO: 805404 R E C O M M E N D A T I O N : The following wording should be deleted:

°t" . . . . . . . z - r en . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . .

SUBSTANTIATION: This type of failure sets a dangerous precedent , and the failure itself is no t really credible. This essentially makes all values in series ana logous to a h i / l o pressure valve due to the way it is being treated. There is no historical evidence or test results that suppor t two valves open ing

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N F P A 805 - - N o v e m b e r 2 0 0 0 R O C

simultaneously. This type of failure is outside the scope o f the analysis for many BWRs. C O M M r r T E E ACTION: R e j e c t . COMMITTEE STATEMENT: Both valves for the equ ipment identification process need to be included. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITrEE ACTION:


(Log #48) 805- 112- (B-2-2, Step 3 (c) (2)): Accept in Principle SUBMITTER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise the sentence beginning "It should be noted..." to read as follows:

"High/ low pressure interfaces are those reactor coolant boundary valves whose spurious open ing could result in ruptured downstream piping, and the loss of coolant inventory which cannot be mitigated in sufficient time to achieve nuclear safety performance requirements . Examples are residual heat removal (PWR), pressurizer power-operated relief valves (PORVs) at PWRs, and shutdown cooling (BWR). These interfaces are subjected to the more restrictive criteria for circuit analysis and-

SUBSTANTIATION: t . Clarify the sentence - it is confusing. 2. Delete the reference to Chapter 3 - it is not clear "what

additional fire protection features documented in Chapter 3" are being referred to. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commit tee Action on Comment 805-107 (Log #124). NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #125) 805- 113- (B-2-2(a)): Accept in Principle SUBMITTER: Elizabeth A. Kleinsorg, Duke Engineering Services COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete Tables B-2.2(a) and B-2.2(b). SUBSTANTIATION: Tables are misleading as to the selection of equipment . They may be limiting or suggest a hierarchy that does not exist. COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Commit tee Action on C ommen t 805-107 (Log #I24). NUMBER OF COMMITFEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #10) 805- 114 - (B-3-1): Accept in Principle. SUBMITFER: James E. Lechner, Nebraska Public Power District COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Delete entire definition of Short Circuit:

SUBSTANTIATION: Bounded by term "Hot Short" not part o f industry vernacular for circuit failure modes. COMMITI'EE ACTION: Accept in Prin<:iple. COMMITTEE STATEMENT: See Commit tee Action on C ommen t 805-107 (Log #124). NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #50) 805- 115 - (B-3-2, Step 3 (d)): Accept in Principle SUBMITrER: Frederick A. Emerson, Nuclear Energy Institute COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Replace ~is postulated to" and ~are

~ ostulated to" with "could". UBSTANTIATION: Use of the word "could" is consistent with

the language in Step 3(a), and 3(e). Requirements for analysis are already stated in 3(a), "all possible failure states must be evaluated". COMMITTEE ACTION: Accept in Principle. COMMITTEE STATEMENT: See Committee Action on Commen t 805-107 (Log #124). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #79) 805- 116 - (B-3-2 Step 3 (d) and (e)): Reject SUBMITTER: Robert C. Daley, Entergy Operations, Inc. COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: Revise Section B-3.2, Step 3(d) and (e) as follows:

(d) should be written as follows: "A single hot short from an external source is postulated to affect any and all conductors, one at a time, within ~:able that is required or is associated with a required nuclear safety component . " The remainder Of (d) should be deleted.

(e) Delete this step in its entirety. SUBSTANTIATION: Performing cable selection as presently written in these sections would cause undue hardship to BWR

c lants. This port ion of the methodology has no t been, and cannot e performed for the control room circuitry configurating at

BWRs. I believe that if this were to be approved as is, no BWR could conceivably use it. COMMITTEE ACTION:" Reject. COMMITrEE STATEMENT: Other failures as stated in the current d o c u m e n t need to be considered. NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


COMMENT ON AFFIRMATIVE: GARRETT: In determinis t ic analysis, there must be separation 6f

initiating events (cable faults) and consequences (spurious o[~erations). The Industry recognizes that various different types of cable faults can occur dur ing a fire, therefore ho t shorts, shorts to ground and open circuits must be analyzed to ensure that a comprehensive populat ion is included in the "database" of safe shutdown cables. In most cases, the superposit ion of single cable faults o f each type onto each conductor of cable in a circuit in a one at a time fashion will result in this comprehensive populat ion o f cables. The standard proposes a more conservative methodology with respect to recognizing the possibility of concur ren t faulted cable conditions. Since spurious equ ipmen t operat ion can only be dealt with deterministically on a one at a time basis, considerat ion of multiple simultaneous cable faults resulting in concur ren t spurious operat ions must be resolved utilizing risk informed techniques.

I~ATCHFORD: In deterministic analysis, there must be separation of initiating events (cable faults) and consequences (spurious operations). The Industry recognizes that various different types o f cable faults can occur during a fire, therefore hot shorts, shorts to ground and open circuits must be analyzed to ensure that a Comprehensive 16opulation is included in the "database" of safe shutdown cables. In most cases, the superposit ion of single cable faults o f each type onto each conductor of cable in a circuit in a one at a time fashion will result in this comprehensive populat ion of cables. The standard proposes a more conservative methodology with respect to recognizing the possibility of concur ren t faul ted cable conditions. S incespur ious equ ipment operat ion can only be dealt with determmistically on a one at a time basis, considerat ion of multiple simultaneous cable faults resulting in concurren t spurious operat ions must be resolved utilizing risk informed techniques.

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(Log #83) 805- 117 - (B-3.3.2, Step 1): Accept in Principle SUBMITTER: Chris topher Pragman, PECO Energy COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Revise text as follows:

As described previously, secondary ignition can be a veal- concern due to fire-induced electrical faults on inadequately protected cable resulting in secondary ignition. This concern can be addressed by evaluating the plant electrical design to ensure that adequate overcurrcnt pr~,tecfion for all conduc to~ exists *~'~"-~5 ' '~1" . . . . . ~"~'~ v ~ " .. . . . . secondary, fires are prevented bv design. SUBSTANTIATION: NFPA 805 does not need to prescribe a solution to preventing secondary fires. Only specify the object ive--that they be prevented. COMMITTEE ACTION: Accept in Principle. COMMITrEE STATEMENT: See Committee Action taken on Comment 805-107 (Log #124).

The commit tee is not specifying a solution. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #80) 805- 118 - (B-5-1 Step 2(e)): Reject S U B M ~ : Robert C. Daley, Entergy Operations, Inc. COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Section B-5.1, Step 2(e) needs to be reworded as follows:

"Instruments exposed to a fire (e.g., RTDs, thermocouples, pressure transmitters, flow transmitters, (an~ mechanicall-" !ir, kcd . . . . . . : ~ : - - ' :~ ' : ^~°~ are assumed to suffer damage t~at results in failure on the instruments."

Delete the wording "mechanically linked remote/ local indications." S U B S T A N T I A T I O N : Damage to mechanically linked remote/ local indications is not a credible failure mechanism. This failure mechanism has not been evaluated at many BWRs and addition o f this requirement in NFPA 805 could cause undue hardship to utilities. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: The commit tee believes mechanically linked remote/ local indications can be damaged by fire.

See Committee Action on Comment 805-107 (Log #124). NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #11) 805- 119 - (B-5-1, Step 3 (b)): Reject SUBMITTER: James E. Lechner, Nebraska Public Power District COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Delete entire paragraph (h).

Reformat remaining paragraphs. S U B S T A N T I A T I O N : Per BWROG Safe Shutdown Methodology, Appendix G. Consideration does not need to be given to comblncd equ ipment spurious operations. This is especially applicable to MOV's where thc appropriate combinat ion of circuit faults needed is highly improbable. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: Multiple failures can occur as a resuh of fire. NUMBER OF C O M M I T r E E MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: ALPERT: See comments by Sinopoli and Garrett. CARLISLE: I concur with Mr. Sinopoli 's commen t dated

8/9/2000. DAVIS: I concur with Mr. Lechner ' s comments (and the

BWROG positions). GARRETT: Spurious equipment operat ion can only be dealt

with deterministically on a one at a time basis. This is consistent with similar industry accident and transient analytical practices of postulating worst case "single failures." Within existing regulations, postulation of multiple simultaneous failures is

beyond the design basis of any plant. This is where risk techniques Provide value to the industry in determining the potential safety impact of multiple failures based on the probability of those failures occurring. Similarly, risk techniques must be utilized to de te rmine the safety impact of multiple simultaneous fire induced spurious operations. This should be based not only on the probability of the cable faults themselves, but also the likelihood of the specific combination of faults that would be required to result in the simultaneous spurious operations, and the likelihood that these faults would occur in the right combination at the same point in time before o ther faulted conditions could produce a different, less, consequential result, The basis o f the commentor ' s original a rgument is that the .probabilities o f the cable faults not only occurr ing but occurring m the proper combination, with the proper t iming are sufficiently low as to be below currently defined limits o f regulatory concern.

HENNEKE: Ongoing industry work has not shown to date any significant multiple spurious operations. The work has not been completed, and since this is still ongoing, the requirement for protect ing for multiple spurious operations is unsubstantiated and unnecessary. The requirements of the standard should be to

~ rotect nuclear and personnel safety on our experience and nowledge. The requirements and exceptions in o ther sections

are based on testing, and experience. The multiple spurious operat ions requirements are based on concern and lack of knowledge. Industry efforts should be allowed to continue to research the problem, and the standard should be updated once the results are de termined. Since initial results show no concern, the Proposal 805-119 should be accepted.

RATCHFORD: Spurious equ ipment operat ion can only be dealt with deterministically on a one at a time basis. This is consistent with similar industry accident and transient analytical practices of postulating worst case "single failures." Within existing regulations, postulation of multiple simultaneous failures is beyond the design basis o f any plant. This is where risk techniques provide value to the industry in determining the potential safety impact of multiple failures based on the probability of those failures occurring. Similarly, risk techniques must be utilized to de termine the safety impact of multiple simultaneous fire induced spurious operations. This should be based not only on the probability of the cable faults themselves, but also the likelihood of the specific combinat ion of faults that would be required to result in the simultaneous spurious operations, and the likelihood that these faults would occur in the right combination at the same point in time before o the r faulted conditions could produce a different, less, consequential result. The basis o f the commentor ' s original a rgument is that the probabilities of the cable faults not only occurring but occurring in the proper combination, with the proper t iming are sufficiently low as to be below currently defined limits of regulatory concern.

SINOPOLI: Spurious equipment operation can only be dealt with determinisfically on a one at a time basis. This is consistent with similar industry accident and transient analytical practices of postulating worst case "single failures." Within existing regulations, postulation of multiple simultaneous failures is beyond the design basis o f any plant. This is where risk techniques provide value to the industry in determining the potential safety impact o f multiple failures based on the probability of those failures occurring. Similarly, risk techniques must be utilized to de termine the safety impact o f multiple simultaneous fire induced spurious operations. This should be based not only on the probability of the cable faults themselves, but also the likelihood of the specific combinat ion of faults that would be required to result in the simultaneous spurious operations, and the likelihood that these faults would occur in the right combination at the same point in time before o ther faulted conditions could produce a different, less, consequential result. The basis o f the commentor ' s original a rgument is that the probabilities o f the cable faults not only occurring but occurring in the proper combination, with the p roper t iming are sufficiently low as to be below currently defined limits o f regulatory concern. (Technical input provided by Jim Lechner.)

(Log #12) 805- 120 - (B-5-1 Step 3 (c)): Reject S U B M r l ~ E R : James E. Lechner, Nebraska Public Power District C O M M E N T O N PROPOSAL NO: 805-404 RECOMMENDATIO N: Delete the following text:

_..v".V~. . . . . . . . . . ~ - T : " \ " . . . . . . . . . . . . . v ' ~ v ' ~ w . v . . . . . . . . . . . . . . . . . . . . .

312


S U B S T A N T I A T I O N : Based on BWROG Safe Shutdown Methodology, Append ix C the H i / L o pressure interface concerns are highly improbable and should not be treated any differently than any o ther MOV. The event is not credible. COMMITTEE ACTION: Reject. COMMITTEE STATEMENT: His substant iat ion does not agree with the proposed r ecommenda t ion . H i / L o pressure intmf'ace concerns should be treated differently than MOV's. NUMBER OF COMMITTEE MEMBI~RS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


EXPLANATION OF NEGATIVE: ALPERT: See c o m m e n t s by Sinopoli and Garrett. CARLISLE: I concur with Mr. Sinopoli 's c o m m e n t dated 8 /9 /2000 . DAVIS: I concur with Mr. Lechne r ' s c o m m e n t s (and the BWROG

posi t ions) . GARRETT: The case of Hi /Low pressure interfaces has historically

been treated as a highly improbable, but highly consequent ia l event with special assumpt ions applied to do determinis t ic analysis. Nearly 25 years o f risk in formed t e c h n i q n e d e v e l o p m e n t have occurred since the fire protect ion rules were promulgated . A sufficient body of knowledge exists such that the same risk techniqttes postulated above are equally applicable to H i /Low pressure interfaces as any o ther occur rence o f mult iple s imul taneous spur ious e q u i p m e n t operat ion.

HENNEKE: See my F.xplanation o f Negative Vote on C o m m e n t 805- 119 (Log #11).

RATCHFORD: The case of H i /Low pressure interfaces has historically been treated_ as a highly improbable , but highly consequent ia l event with special assumpt ions applied to do detet~ninistic analysis. Nearly 25 years o f risk in formed technique deve lopment have occurred since the fire protection rules were promulgated . A sufficient body of knowledge exists such that the same risk techniques postulated above are equally applicable to H i /Low pressure interfaces as any o ther occur rence of nruhiple s imul taneous spur ious e q u i p m e n t operat ion.

SINOPOLI: The case o f H i /Low pressure . interfaces has historically been treated as a highly improbable, but highly consequent ia l event wilh special assumpt ions applied to do determinist ic analysis. Nearly 25 years of risk in formed technique deve lopment have occurred since the fire protection rules were promulgated . A sufficient body of knowledge exists such that the same risk techniques postulated above are equally applicable to Hi /Low pressure interfaces as any o ther occur rence of mult iple s imul taneous spur ious e q u i p m e n t operat ion. (Technical input provided by J im Lechner . )

(Log #51 ) 805- 121 - (B-6): Accept in Principle SUBMITFER: Frederick A. Emerson , Nuclear Energy Institute COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Change "<200EF" to "200°F ". S U B S T A N T I A T I O N : Typographical error. COMMITTEE ACTION: Accept ill Principle. COMMITTEE STATEMENT: See Commi t tee Action on C o m m e n t 805-30 (Log #93). NUMBER OF COMMITTEE MEMBERS ELIGIBLE T O VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #58) 805- 122- (B-6): Accept S U B M r l T E R : Dennis H e n n e k e , Sou the rn California Edison COMMENT O N PROPOSAL NO: 805404 RECOMMENDATION: Revise second parag~-aph, sen tence start ing "In general" change "Low Risk Evolution" to "High Risk Evolution".

] Also change at" to 'ahove". S U B S T A N T I A T I O N : Possibly at or above normal power levels and high risk while possibly below normal power risks are low risk. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #59) 805- 123 - (B-6): Accept SUBMITTER: Dennis I t e n n e k e , Southern California Edison COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: Add to the end of tile first sentence in "Nuclear safety..." section, change to read:

"should be used for equ ipmen t required in high risk evolutions." Add the end of the sen tence start ing "For spent fuel pool

coo ng add "for e q u i p m e n t n h gh r sk evohtt ons. SUBSTANTIATION: The Nuclear Safety capability assessment is only needed for high risk evohttions.

This needs to be clear here. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:

AFFIRMATIVE: 2q NEGATIVE: 1 NOT RETURNED: 2 Biggins, Vieiva

EXPLANATION OF NEGATIVE: CONNELL: This c o m m e n t deleted the gu idance in Section B-6

to identifv the equ ipmen t and cables required for spent fuel pool cooling u'nless tile plant opera t ing state that relies upon the e q n i p m e n t or cables is classified as a "high risk evolution." The definit ion o f a "high risk evolution" provided in the appendix will exclnde the spent fltel pool e q u i p m e n t and cables fi-om considerat ion in almost all cases, as r e c o m m e n d e d in Section D-4. The equ ipmen t and circuits required for spent fuel pool cooling should be identified regardless of the risk cl~issification. The fire protect ion systems or features provided for that e q u i p m e n t and cables, if any, could be based, in part, on valid risk insights. In addition, the reference to NUMARC 91-06 in Sections B-6 and D-4 are inappropr ia te as this d o c u m e n t does not provide a process for de te rn l in ing the classification (high or low) of an evolution as is al luded to in the draft s tandard.

(Log #60) 805- 124 - (B-6): Accept SUBMITTER: Dennis H e n n e k e , Sou thern California Edison CO MMENT ON PROPOSAl . NO: 805404 RECOMMENDATION: Add to tile end o f the first sentence for the third, fourth, and fifth sections of B-6, the s tatement:

"for equ p m e n t required in high risk evolut ons. S U B S T A N T I A T I O N : The Circuit Analysis Cable Identification and Area Assessment is only required for high risk evolutions. This needs to be cleat" here. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


(Log #CC4) 805- 125 - (C-2-2): Accept SUBMITTER: Technical Commi t t ee on Fire Protection for Nuclear Facilities COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: In Section C-2.2, u n d e r the definit ion of "Fire" and Table C-2.2(a) , change "COMPBRN III" to "COMPBRN IIIe". S U B S T A N T I A T I O N : Correction• See reference (2) for Appendix C. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE O N COMMITTEE ACTION:


• (Log #CCI) 805- 126 - (C-6.2(1)): Accept SUBMITTER: Technical Commi t t ee on Fire Protection for Nuclear Facilities CO MMENT O N PROPOSAL NO: 805-404

I RECOMMENDATION: Add "Draft" before "NUREG 1521". S U B S T A N T I A T I O N : This d o c u m e n t is a draft d o c u m e n t at this t ime. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26

313


VOTE ON COMMITTEE ACTION: AFFIRMATIVE: 24 NOT RETURNED: 2 Biggins, Vieira

(Log #2) 805- 127- (D-3-3): Reject SUBMITTER: Ronald Rispoli, Entergy Corp. COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: Add a s tatement that 3 hour Appendix R fire barriers provide acceptable barriers to fire spread and do not need to be explicitly analyzed for fire spread. SUBSTANTIATION: Requiring explicit t reatment of 3 hour fire barriers has no risk-significance. If we confine our analysis to fire areas with 3 hour fire barriers, we should not have to consider

0oread; there is no value in it. MMITrEE ACTION: Reject.

COMMITTEE STATEMENT: The availability of the fire barrier is independen t of the fire resistance rating. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #61) 805- 128- (D-3-5): Accept SUBMITTER: Dennis Henneke , Southern California Edison COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Change "In a typical fire PSA, ...discuss" to "the analysis should consider". SUBSTANTIATION: Change provides more exact wording. COMMITTEE ACTION: Accept. NUMBER OF COMMrIq'EE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #62) 805- 129- (D-3-6): Accept SUBMITTER: Dennis Henneke , Southern California Edison COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Sentence starting "Regulatory guide..." should have the following added "on uncertainty analysis methods" to the end of the sentence. SUBSTANTIATION: Wording is not a complete sentence. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #CC5) 805- 132- (D-5): Accept SUBMITTER: Technical Commit tee on Fire Protection for Nuclear Facilities COMMENT ON PROPOSAL NO: 805-404 RECOMMENDATION: In Section D-5, Application of Fire PSA Methods to Change Analysis, add the following paragraph at the end of section:

"Section D-3 provides general guidance for performing a fire PSA that may be applied to shutdown and low power operations. Ano the r acceptable approach is qualitative examination of the impact o f the p roposedchange to de termine if it results in an increase in risk during shutdown and low power operation. For example if the proposed change in the switchgear room is a new sprinkler system, the post modification fire scenarios (with lower rated ERFBS and automatic suppression) should be demonstra ted to be equivalent to or better than the pre modification (with 1 hour EREBS and no automatic suppression) dur ing shutdown and low power operations." SUBSTANTIATION: Add information on how to use fire PSA during various modes of plant operat ion. COMMITTEE ACTION: Accept. NUMBER OF COMMrrTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #CC6) 805- 133- (D-7): Accept SUBMITTER: Technical Commit tee on Fire Protection for Nuclear Facilities COMMENT ON PROPOSAL NO: 805404 RECOMMENDATION: In section D-7 References:

1) Delete reference 4. 2) Revise reference 9 to "Guidance for Development of Response

to Genetic RAI on Fire IPEEE, EPRI SU-105928, March 2000." SUBSTANTIATION: Correction and clarification. COMMITTEE ACTION: Accept. NUMBER OF COMMITTEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #63) 805- 130 - (D-4): Accept SUBMITrER: Dennis Henneke , Southern California Edison COMMENT ON PROPOSAL NO: 805-404

I RECOMMENDATION: Change reference to Section 4-2 to Section B-6. SUBSTANTIATION: Section 4-2 moved to Appendix B in the last revision. COMMrI ' rEE ACTION: Accept. NUMBER OF COMMITrEE MEMBERS ELIGIBLE TO VOTE: 26 VOTE ON COMMITTEE ACTION:


(Log #64) 805- 131 - (D-4): Accept SUBMrI'rER: Dennis Henneke , Southern California Edison COMMENT O N PROPOSAL NO: 805-404 RECOMMENDATION: 1. Correct formatt ing problems for section (bold heading, indent High Risk Evolutions).

2. Change last word in section - "required" to "used". SUBSTANTIATION: 1. Format issues.

2. Methods acceptable to the AHJ may be used. "Required" is alread implied by the word "acceptable".

Editorial Correction

The Technical Commit tee on Nuclear Facilities notes the following editorial changes to NFPA 805, Performance-Based S tandard for Fire Protection for Light Water Reactor Electric Generat ing Plants.

1. In Section A41 .4 , B-2.2, step 5, (b)(3), B-5.2(e), B-5.2(e)(2) change "manual actions" to "recovery actions". This was intended to be per formed universally throughtout the document , but was missed.

2. In A-l-6 change "Risk Informed" to "Risk Informed Approach". This change is made to be consistent with the definition m Chapter 1.

3. Delete the last sentence of A-2-3.3.3. Commit tee Statement: Total plant risk evaluation was modified and should be deleted in the appendix.

314


4. R e n u m b e r A'2- ~.'2,.~ to A 2 3.~.3 A-2-3.3 to A-2-3.4 A-2-3.3.1 to A-2-3.4.1 A-2-3.3.2 to A-2-3.4.2 A-2-3.3.3 to A-2-3.4.3 A-2-3.4 to A-2-4.4 A-2-3.4.1 to A-2-3.5.1 A-2-3.4.2 to A-2-3.5.2 A-2-3.4.43 to A-2-3.5.3

This matches the append ix material in the text in the body.

5. In B-3.2, Step 3, (cl add the word "recovery" before the word "action" and "actions". This was in tended to be a universal change.

6. In B-3.2, Step 3, (f), change "short circuit individnal" to "short to individual". Editorial.

7. Ill C-6.1, (12) Revise the third sentence to read: "Heat trmtsfer through. . . "

Revise the fourth sentence to read: "There can he several (up to about nine) fires in a ..."

Editorial.

8. In C-6.2(1) the word "Draft" before "NUREG 1521".

9. Revise Table C-2-2(b) as shown on the next page: Correct ions were made to the table accurate.

315

f Program Type Number Wall Heat Lower Heat

CFAST [C-6.1 (1)1 FASTLITE [C-6.1 (5)1 COMP-BRN Ill [G6.1 (2)1

F1YE [C-6.1 (6)]

FL,~dVIME [C.6.1 (lO)] MAGIC [C-6.1 (12)1 FLOW - 3D [C-6.1 (11)]

LE5 [C-6.1 (8)1 FPETOOL [C-6.1 (7)]

ASCOS [c-6.1 (9)1

CONTAM [C-6.1 (3)1

Zone

Zone

Zolle

of Rooms

15

Transfer

Yes Yes

Yes Yes

Yes

Level Gas Targets Temp

No

No

No Yes

Provides initial screen, leads to use of PRAs, look up tables

Multi

Multi

Few

CFD

Zone

Network flow

.°

Few

2 1/2

Multi

Multi

Yes Real

Yes ba~-Yes

Yes Real

Yes Real

No No

No N/A

No N/A

Yes

Yes

Yes

Yes

No

No

No

Fire Gas 0 2 Vertical HVAC Detectors Sprinklers Remarks Concentrations Depletion Connections FAN and

Ducts Specified Yes Yes Yes Yes Yes Yes Fewer rooms if multiple PC Specified Yes Yes Yes Yes Yes Yes Easy input and

run for PC Growth No Yes No No Yes No Input calculation distributions for

Mote-Carlo calculatmns Gathers mfo and keeps records I no computer neeessal)'

Specified Yes Yes No Yes No No French, ISPN multiple

Specified Yes Yes Yes Ye......fis ~.~ No French, EdF multiple Specified Yes Yes Yes Yes Yes i Depends on

user, significant computing time, andacceptable granularity

Specified Yes Yes Yes Yes Yes Yes

Specified Yes Yes No " No Yes No Easy inputs for PC, has "TOOLS"

N/A No N/A Yes No N/A N/A ASHRAE document (for smoke flow)

N/A Yes N/A Yes No N/A N/A Superior nunlerics, front end, and graphics (for smoke flow)

Z

I Z O ,¢

t ' l

dennis j. kubicki, u.s. dept. of energy, md tel€¦ · (log #13) 805- 2 - (1-2): accept in...

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