response to request for additional …response to request for additional information apr1400 design...

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

APR1400 Design Certification

Korea Electric Power Corporation / Korea Hydro & Nuclear Power Co., LTD

Docket No. 52-046

RAI No.: 92-8068

SRP Section: 03.09.05 – Reactor Pressure Vessel Internals

Application Section: 3.9.5

Date of RAI Issue: 07/21/2015

Question No. 03.09.05-1

GDC 1 and 10 CFR 50.55a require that reactor internals be designed to quality standards commensurate with the importance of the safety functions performed.

A public meeting was held on June 23, 2015, and the applicant provided written material to support the meeting discussion, formally documented in a letter dated July 6, 2015.

The applicant stated that the hold-down ring provides axial force on the flanges of the UGS assembly and the core support barrel assembly in order to prevent movement of the structures under hydraulic forces. The hold-down ring is designed to accommodate the differential thermal expansion between the reactor vessel and the reactor internals in the vessel ledge region. The UGS assembly and core support barrel assembly including the hold-down ring are supported on the reactor vessel ledge. Therefore, the hold-down ring is classified as an internal structure since it is not a major component which provides direct support or restraint of the core within the reactor vessel.

The applicant also stated that loss of preload may occur due to loss of deflection of the hold-down ring as a result of wear on contact surface and stress relaxation during operation. This loss of preload will decrease an axial load on the core support barrel and the UGS flange surface and then induce relative motion between the core support barrel flange and the UGS flange under service and accident loadings. Considering loss of preload, the hold-down ring is designed to have enough preload to prevent relative motion of reactor internal components. The applicant further stated that the function of the hold-down ring is inspected via the Comprehensive Vibration Assessment Program (CVAP).

The staff reviewed this information and found the applicant’s explanation to classify the hold-down ring as an internal structure, and the explanation for the potential loss of preload of the hold-down ring, to be insufficient. Specifically, the information the applicant provided regarding the function of the hold-down ring (i.e., it provides axial force on the flanges of the

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UGS assembly and the core support barrel assembly in order to prevent movement of the structures under hydraulic forces) is the same information already provided in DCD Tier 2, Section 3.9.5.1.2. Simply stating that the hold-down ring is classified as an internal structure because it is not a major component is not sufficient. In addition, as described above, the applicant stated that loss of preload may occur and induce relative motion between the core support barrel flange and the UGS flange under service and accident loadings. However, the applicant further stated that the hold-down ring is designed to have enough preload to prevent relative motion of reactor internals components. The staff found the applicant’s explanation regarding loss of preload of the hold-down ring to be contradictory.

Therefore, the applicant is requested to provide a detailed explanation as to why the hold-down ring is classified as an internal structure. In addition, the applicant is requested to further explain the consequences of a loss of preload of the hold-down ring during all normal and accident conditions. The applicant is also requested to provide information for any analyses performed to confirm the functional and structural integrity of the core support barrel assembly and UGS assembly due to a loss of preload of the hold-down ring. A summary of this information should be included in the DCD.

Response

The hold-down ring is not required to directly support the fuel assemblies because the analysis of the reactor internals models the load path of the hold-down ring connected in parallel rather than in series with the fuel assemblies. The vertical loads from the fuel assemblies will be transferred through the hold-down ring to the upper guide structure and core support barrel upper flanges and then to the reactor vessel ledge. During normal operation, the core support barrel flange would not lift since the hydraulic lift force is smaller than the downward loads such as the dead weight and hold-down spring force of the fuel assemblies. In seismic/ accident conditions, the core support barrel may lift from the reactor vessel ledge, but the lift displacement is limited by the small stroke of the hold-down ring. These functions are not considered to be core support requirements, but rather functional requirements. Therefore, the hold-down ring is classified as an internal structure and is not considered to be a core support structure. The design specification for reactor internals require that ASME Code, Section III, NG requirements be applied for the design and manufacturing of the internal structure. Although the hold-down ring is classified as an internal structure, it is constructed under the same requirements for a core support structure.

The current hold-down ring is designed to have sufficient preload to cover a loss of preload due to wear-in and stress relaxation during normal and accident condition. The configuration of the hold-down ring will remain unchanged even though the hold-down ring loses its preload from stress relaxation. It is confirmed by design analysis and CVAP inspection that there is no relative motion of the reactor internals components during Level A, level B, level C and level D services conditions. Also, there is no adverse effect on function and structural integrity of the CSB Assembly and UGS Assembly due to the design basis loss of preload described previously. The information above is considered too detailed to be specifically included in the DCD.

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Impact on DCD

There is no impact on the DCD.

Impact on PRA

There is no impact on the PRA.

Impact on Technical Specifications

There is no impact on the Technical Specifications.

Impact on Technical/Topical/Environmental Reports

There is no impact on any Technical, Topical or Environmental Reports.

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Docket No. 52-046

RAI No.: 92-8068





GDC 1 and 10 CFR 50.55a require that reactor internals be designed to quality standards commensurate with the importance of the safety functions performed. Standard Review Plan Section 3.9.5, “Reactor Pressure Vessel internals,” Areas of Review 1 includes the physical and design arrangements of all reactor internals structures, components, assemblies, and systems, including the positioning and securing of such items within the RPV, the provision for axial and lateral retention and support of the internals assemblies and components, and the accommodation of dimensional changes due to thermal and other effects.


The applicant indicated that threaded structural fasteners are used in the fuel locating pins, which are attached to the top of the lower support structure beams to provide orientation for the lower ends of the fuel assemblies. The fuel locating pins are secured by tack weld to the lock-bar. The fuel locating pins are designed in accordance with ASME BPV Code, Section III, Subsection NG.

The design of the fuel locating pins, however, is not completely clear. Specifically, the applicant is requested to provide the detailed design of the fuel locating pin, its classification, and its design code for both the fuel pin portion and the threaded structural fastener portion. A summary of this information should be included in the DCD.

Response

Fuel locating pins (i.e., fuel insert pins) are designed to guide and laterally support the bottom of the fuel assembly. The pins are evaluated at various locations including the shank, flange, fuel lower end fitting to pin contact, pin to beam contact, thread bearing part, thread relief

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part, and the thread shear part of the fuel locating pin. The fuel locating pins have large margins enough to support the fuel assembly.

All of the fuel locating pins are part of the lower support structure and are classified as a core support structure. The fuel locating pins including the threaded portions are also classified as a core support structure in the following figure which is included in the design specification for reactor internals. Therefore, the fuel locating pins, including the threaded portions, are evaluated in accordance with ASME Code, Section III, Subsection NG.

Code Classification of LSS/ICI Nozzle Assembly

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The information provided is considered excessive to be specifically included in the DCD.

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APR1400 Design Certification Korea Electric Power Corporation / Korea Hydro & Nuclear Power Co., LTD

Docket No. 52-046

RAI No.: 92-8068





The applicant stated that the plates of the lower support structure consist of a bottom plate and a raised bottom plate. The plates contain flow holes to provide a uniform distribution at the core inlet. Upon entering the inlet nozzles and into the downcomer region, the flow turns upward through the lower support structure plate and through the core. The plates are divided into various flow hole patterns. The patterns are determined by factors such as the ICI locations, the intersection of the support beams with the bottom plate, and the boundary between the raised (peripheral region) and the bottom (central region) portions of the bottom plate. The bottom plate is welded to the lower end of the main support beam, while the raised bottom plate is welded to the main support beam and the lower portions of the lower support structure cylinder. Based on the response provided by the applicant, the staff is unclear to the design of the bottom plate, as well as its function, classification, and design standard. Therefore, the applicant is requested to provide further explanation of the bottom plate for both the raised and bottom portions. The applicant is also requested to provide a drawing of the bottom plate.

Response

The Lower Support Structure (LSS) bottom plate is termed “Bottom Plate” or “Raised Bottom Plate” based on their location. The bottom plates are comprised of seven individual sections and the sections are of four different types. The raised bottom plates are comprised of twenty-four individual sections and the sections are of five different types as shown in Figure 3-1.

The LSS bottom plate is full penetration welded to the LSS main support beam shown in section B-B of Figure 3-1. The LSS raised bottom plate is full penetration welded to the LSS main support beam and cylinder as shown in section C-C of Figure 3-1.

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The pattern of the LSS bottom plates are designed to distribute the reactor coolant flow as uniformly as possible over the pattern cross-section. The LSS bottom plates which are welded to LSS support beam and cylinder provide axial support for the core. Therefore, the LSS bottom plates are classified as core support structures according to NG-1120 and are designed according to the ASME NG Code. The classification of the core support structures including the LSS bottom plate is listed in Table 3-1.

For the detailed configuration of the LSS bottom plate, see the drawing in Figure 3-2.

Item Safety Class Quality Group Seismic Category Core Support Structure SC-3 C I

Table 3-1 Classification of Core Support Structure

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Figure 3-1 LSS Bottom Plate and Raised Bottom Plate

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Figure 3-2 LSS Bottom Plate and Raised Bottom Plate Drawing

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Figure 3-2 LSS Bottom Plate and Raised Bottom Plate Drawing (Continued)

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Figure 3-2 LSS Bottom Plate and Raised Bottom Plate Drawing (Continued)

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Impact on DCD


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Impact on Technical/Topical/Environmental Report

There is no impact on any Technical, Topical and Environmental Reports.

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Docket No. 52-046

RAI No.: 92-8068







The applicant stated that the lower support structure assembly has four support column assemblies to support the bottom plate and the ICI nozzle support plate. Each support column assembly consists of one column boss and three support columns. The upper part of the support column is welded to the lower support structure bottom plate, and the lower part of the support column is welded to the column boss attached to the ICI nozzle support plate. The applicant also provided a detailed drawing showing the support column assembly.

The applicant is requested to further explain the means by which the column boss is attached to the ICI nozzle support plate, as well as the function, classification and design code used for the design of the support column assembly, including the support columns and column boss. The applicant is also requested to include Figure 6-2 of the written response in DCD Tier 2, Section 3.9.5 such that the DCD depicts the complete design arrangement of the reactor internals.

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Response

The column boss is full penetration welded circumferentially to the ICI nozzle support plate and the column is full penetration welded on the surface of the column boss as shown in Figure 4-1.

The column and column boss connect the ICI support plate to the lower support structure and provide axial support for the ICI support plate. The column and column boss are not designed to provide direct support or restraint of the core. Therefore, the column and column boss are classified as internal structures according to NG-1120. Even if they were classified as an internal structure, the column and column boss are designed according to the ASME NG Code.

Figure 4-1 Column to column boss welding

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Impact on DCD

There is no impact on the DCD

Impact on PRA






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Docket No. 52-046

RAI No.: 92-8068







Based on this information, it remains unclear how the core shroud is secured to the lower support structure and the core barrel. Therefore, the applicant is requested to provide this explanation. The applicant is also requested to describe any core bypass flows. A summary of this information should be included in the DCD.

Response

The core shroud (CS) is secured to the lower support structure (LSS) by a welded flexural connection between the LSS cylinder and the CS bottom plate at the manufacturing facility. The assembled CS/LSS is then secured to the core support barrel (CSB) by a welded flexural connection between the CSB lower flange and the LSS cylinder at the site. The configuration and geometry of the flexural welded connection are shown in Detail A of the following figure. A small gap is provided between the core shroud outer perimeter and the core support barrel to limit the amount of bypass flow.

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The description of the reactor core bypass flows is provided in Section 4.4.2.6.1 of the DCD Tier 2. The reactor core bypass flow path is through the outlet nozzle clearances, alignment keyways, core shroud annulus and guide tubes. These bypass flow paths and the percentages of the total vessel flow are given in Table 4.4-3 and shown schematically in Figure 4.4-6 in the DCD Tier 2.

Flexure Weld of CS to LSS and CS/LSS to CSB

The level of detail provided in the above response is considered unnecessary for including in the DCD.

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Docket No. 52-046

RAI No.: 92-8068





1. DCD Tier 2, Section 3.9.5.1.2 states that the control element guide tubes bear the upwardforce on the fuel assembly hold down devices. The staff needs additional information to make a finding regarding the structural integrity of the control element guide tubes in both the normal operating condition and accident conditions, such that insertability of the CEA is not compromised. The applicant is requested to provide a discussion of the following:

o How the structural integrity of the control element guide tubes is maintained due tothe upward force induced from the fuel assemblies through its stated design life of 60years, including in events such as an SSE.

o The mechanism to prevent the control element guide tubes from buckling during bothnormal operating conditions and other postulated conditions such as an SSE.

2. The staff also requests additional information about design provisions that would preventmisalignment from the fuel assembly guide posts and its impact on the control element guide tubes and insert tubes after each refueling outage. According to the letter referenced above, after the core is defueled and refueled, a total of 964 tubes (both the control element guide tubes and insert tubes) need to fit into the fuel assembly guide posts when the UGS assembly is

lifted and put back into the reactor vessel, on top of the fuel assemblies. If there is any misalignment, the bottom end of these control element guide tubes or insert tubes could be pitched or crimped without any indication. This not only could potentially damage the fuel assemblies due to excess compressive force exerted on them, but in the case of a fuel assembly with control element guide tubes, this could also prevent the CEA from inserting into the fuel assembly if a control element guide tube is pitched or crimped. Operating experience, documented in PNO-IV-96-016, “Damaged Fuel Assembly Found During Core Defueling,” dated

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March 28, 1996, and its supplements detail an event that took place during an refueling outage on March 24-25, 1996 at Palo Verde Unit 2. A fuel assembly could not be removed and was found to be damaged. Damage was also found to the upper guide structure in the area where the damaged fuel assembly was located. The applicant is requested to provide a discussion of the following:

o Analyses performed for the control element guide tubes and insert tubes in terms ofhow the structural integrity can be maintained throughout its design life

o Design provisions to address any misalignment issue during refueling outages whenthe UGS assembly is put back into the reactor vessel

o Design provisions to ensure that a similar incident to the Palo Verde event statedabove, or other significant operating experience related to reactor internals, will notoccur in the APR1400 design

o Inspection results from similar operating plants that address these issues

Response

The component designer evaluates the stress of control element guide tubes and insert tubes (or control element guide tube extensions) due to the fuel holddown springs, in-water weights and fluid-induced axial and lateral load for Level A, B, C and D conditions. For the Level D condition, the stress as a result of an SSE is considered through the SRSS with other stresses. The deflection limit between the control element assembly and the control element guide tube is evaluated to ensure the CEA insertability during accident conditions. Additionally, the cumulative fatigue usage factor corresponding with a design life of 60 years is calculated taking account normal operation and accident conditions, including an SSE.

The critical buckling stress of the CEA guide tube at the design temperature is evaluated according to ASME NG-3211 and NG-3133. According to NG-3133.6 (a) & (b), the maximum allowable compressive stress shall be taken to meet the minimum buckling stress at the design temperature.

The structural integrity of the control element guide tubes is verified by the same manner as discussed in response #1.

To ensure that the UGS CEA guide tubes will properly engage the fuel assembly guide post during installation of the UGS with the core in place, the true position tolerance for the fuel assembly guide post is tightly maintained. The true position tolerance should be within the allowable offset between the CEA guide tube and the fuel assembly guide post.

The APR1400 is new type reactor for domestic plants without any operational history. The visual inspection results of baseline and post hot functional testing will be used to show that

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there will be no indications of damage at the outermost tubes, which are subjected to the highest loads and stress by the cross flow on the CEA guide tubes.

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Docket No. 52-046

RAI No.: 92-8068







More information is needed for the staff to make its finding for the UGS assembly. Based on the information the applicant provided in the letter dated July 10, 2015, the staff is still unclear how the shroud tubes and web plates inside the IBA are attached to the bottom of the UGS plate. It also appears to the staff that, from DCD Tier 2, Figure 3.9-13, the UGS assembly consists of two major barrels: the IBA and UGS barrel. No information is provided in terms of how these two barrels are attached to each other and how they are attached to the UGS support plate, and how the shroud tubes and web plates are connected to the IBA. In addition, the applicant is requested to provide justification on why the IBA is classified as internal structure as indicated in DCD Tier 2, Section 3.9.5.1, rather than as a core support structure. Lastly, DCD Tier 2, Figure 3.9-13 shows an inverted top plate at the top of the UGS assembly. No information is provided in the DCD for this inverted top plate. Therefore, the applicant is requested to provide information for the aforementioned requests and to provide a detailed drawing of how the UGS assembly is assembled. A summary of this information should be included in the DCD.

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Response

The UGS support plate is not attached to the IBA. The IBA flange is attached to the UGS upper flange only. The IBA upper flange is circumferentially welded with a full penetration weld on the upper surface of the UGS upper flange as shown in Detail-A of the following figure. The shroud tubes and web plates inside the IBA are not attached to the bottom of the UGS support plate as shown in Detail-B of the following figure. The CEA shroud is attached to the inside of the IBA cylinder only. The outer shroud tubes are welded with full penetration welds on the inside surface of the IBA cylinder axially as shown in Detail-A of the following figure.

The IBA is designed to guide the vertical movement of the CEA, preventing interaction between adjacent CEAs and to provide guidance for the CEA extension shafts and lateral support for the CEAs. Because the IBA is not a major component which provides direct support or restraint of the core within the reactor vessel, it is classified as an internal structure.

The top plate of the IBA provides guidance for the CEA extension shafts into the closure head nozzles when the closure head is being lowered onto the reactor vessel.

Upper Guide Structure Assembly

Since the above response contains a level of detail that is not appropriate to be included in the DCD, it is considered unnecessary to incorporate it into the DCD.

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Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Section 3.9.5.1.3 provides a description of the flow skirt. The flow skirt is a right circular cylinder, perforated with flow holes, and reinforced with two stiffening rings. It is supported by nine equally spaced machined sections that are welded to the bottom head of the reactor vessel. The function of the flow skirt is to reduce inequalities in core inlet flow distribution and to prevent formation of large vortices in the lower plenum.

Based on the information provided in the DCD, additional information is needed to understand the design arrangement of the flow skirt. Therefore, the applicant is requested to provide a detailed drawing of the flow skirt and the location and method at which it is attached to the bottom head of the reactor vessel. In addition, the applicant is requested to clarify the classification of the flow skirt and the structural integrity of the flow skirt under all service level conditions. A summary of this information should be included in the DCD.

Response

Even though the flow skirt is located inside the reactor vessel, it is classified as an attachment of the reactor vessel in accordance with NB-1130 of the ASME Code, Section III. Therefore, the flow skirt is classified as ASME Code Class 1 and designed to Subsection NB of the ASME Code, Section III.

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The flow skirt is welded to the weld build-up pad of the bottom head. The design arrangement and configuration of the flow skirt are shown in the following figures:

Flow Skirt

Flow Skirt Arrangement

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The design meets ASME Code, Section III, Subsection NB requirements. The evaluated locations and significant results are shown below.

*This is the cumulative usage factor.

CONDITION CODE CRITERION

LOCATION STRESS (KSI) ALLOWABLE

(KSI)

DESIGN

PL + PB FLEX TAB

RING PERFORATED REGION

TRIAXIAL FLEX TAB


FAULTED PL + PB FLEX TAB


SERVICE

RANGE OF S.I.

FLEX TAB


FATIGUE FLEX TAB


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Based on the level of detail specified in the above response, it is considered unnecessary to include this information in the DCD.

Impact on DCD There is no impact on the DCD. Impact on PRA There is no impact on the PRA. Impact on Technical Specifications There is no impact on the Technical Specifications. Impact on Technical/Topical/Environmental Reports There is no impact on any Technical, Topical or Environmental Reports.

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Docket No. 52-046

RAI No.: 92-8068


Application Section: SRP 3.9.5



GDC 1 and 10 CFR 50.55a require that reactor internals be designed to quality standards commensurate with the importance of the safety functions performed. Standard Review Plan Section 3.9.5, “Reactor Pressure Vessel Internals,” Areas of Review 1 includes the physical and design arrangements of all reactor internals structures, components, assemblies, and systems, including the positioning and securing of such items within the RPV, the provision for axial and lateral retention and support of the internals assemblies and components, and the accommodation of dimensional changes due to thermal and other effects.

No information is provided for the static O-ring seal at the seal table for the In-Core Instrumentation system, which forms the reactor coolant pressure boundary. Therefore, the applicant is requested to provide information for the static O-ring seal, including its classification and design requirement. A summary of this information should be included in the DCD.

Response

The static O-ring seals are classified as safety related and are designed to the requirements as follows:

The static O-ring seal shall be capable of leak tight operation for a design life of two years under the following design conditions:

Pressure difference from 0 to 2500 psi. Temperature range from 40 ℉ to 150 ℉.

Radiation level of 2E04 Gy (This value includes gamma and neutron radiation).

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The O-ring seals generally used for the In-Core Instrumentation system are replaced every refueling outage according to the supplier’s recommendation. The function and design requirements for the O-ring seals are included in the existing DCD explanation (see 3.9.5.1.4).

Impact on DCD

DCD Tier 2, Section 3.9.5.1.4 will be revised as indicated in the attachment.

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APR1400 DCD TIER 2

3.9-84

3.9.5.1.4 In-Core Instrumentation Support System

The in-core neutron flux monitoring system includes self-powered, in-core detector assemblies, supporting structures and guide paths and an amplifier system to process detector signals. The self-powered in-core detector assemblies and the amplifier system are described in Section 7.7. The instrumentation supporting structures and guide paths are described in this section and shown in Figure 3.9-14.

The support system begins outside the reactor vessel, penetrates the bottom of the vessel boundary and terminates in the upper end of the fuel assembly. Each in-core instrument is guided over the full length by the external guidance conduit, ICI guide tube nozzles of the reactor vessel, the lower support structure, and the guidepost of the fuel assembly. Figure 3.9-14 shows the in-core instrument support structure. The in-core instrumentation support system routes the instruments so that detectors are located in selected fuel assemblies throughout the core. An equal instrument has the same length for all locations. The guide tube routing outside the reactor vessel is a simple 180-degree bend to the seal table. The pressure boundaries for the individual instruments are at the out-of-reactor seal table, where the external electrical connections to the in-core instruments are made. Each instrument has an integral seal plug which forms a seal at the instrument seal table and through which the signal cables pass. Static O-ring seals are used to seal against operating pressure.

Loading Conditions 3.9.5.2

The following loading conditions are considered in the design of the reactor internals:

a. Normal operating temperature differences

b. Normal operating pressure differences

c. Flow loads

d. Weights, reactions, and superimposed loads

e. Vibration loads

Rev. 0

RAI 92-8068 - Question 03.09.05-9 Attachment (1/1)

31749

설명선

The static O-ring seals are classified as safety related, and shall be capable of leak tight operation for a design life of two years under the following design conditions; Pressure difference from 0 to 2500 psi. Temperature range from 40 oF to 150 oF. Radiation level of 2E04 Gy (This value includes gamma and neutron radiation).

40696

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40696

타원

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Docket No. 52-046

RAI No.: 92-8068







The applicant stated that the surveillance capsule assembly is not classified as reactor internals, but as part of the reactor vessel since it is attached to the inside wall of the reactor vessel. The applicant also stated that DCD Tier 2, Section 5.3.1.6 “Material Surveillance” provides information on the surveillance capsule assembly.

The staff understands that the surveillance capsule assembly is attached to the reactor vessel, but disagrees that by virtue of its location, that it is not classified as part of the reactor internals components. SRP Section 3.9.5 defines reactor pressure vessel internals as all structural and mechanical elements inside the reactor vessel, but does not specify a component’s mounting location. The review scope of SRP Section 3.9.5 is not the functionality of the surveillance capsules, but rather the structural integrity of the surveillance capsule assemblies and the method by which the surveillance capsule assemblies are mounted on the reactor vessel. Specifically, the DCD should demonstrate how the surveillance capsule assemblies are mounted to the reactor vessel walls so that their structural integrity can be maintained throughout its design life, and not become loose parts upon a high flow or severe seismic event. Therefore, the applicant is requested to explain the

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design requirement and mounting mechanism of the surveillance capsule assemblies. In addition, if there are other components that are mounted either on the reactor vessel wall or on the core support barrel (e.g., a neutron shield), the applicant is requested to provide similar information for such components. A summary of this information should be included in the DCD.

Response

The surveillance capsule assemblies (SCAs) provide a leak tight barrier to protect the surveillance test specimens and monitors from the impacts of the reactor coolant. The surveillance capsule assemblies fit within the confines of the surveillance capsule holder tubes attached to the reactor vessel wall such that relative vibratory motion is prevented.

The following figure shows the manner in which the SCA is installed in the SCA holder. As shown in the figure, the SCA is supported by pins and the latches of the SCA are inserted into the holes of the SCA holder.

SCA & SCA Holder

The structural integrity of the surveillance capsule holder tubes was provided in design details and substantiated in a stress report and pertinent weld and inspection reports. The SCA structural integrity and holder required wall thickness were determined by a structural evaluation based on design loads defined in the reactor vessel design specification and the requirements of Subsection NB of the ASME Code, Section III. Six (6) SCA holders are attached to the inside of the reactor vessel by full penetration welds.

Other components that are mounted to the reactor vessel inside wall are the core stabilizing lug, core stop, and flow skirts. Information for the flow skirts and core stop are provided in the response to the questions 03.09.05-8 and 03.09.05-12. Information on the core stabilizing lug was provided in the written response to Issue #4 in the material submitted as a result of the public meeting held on June 23, 2015, (refer to the formally docketed letter of July 6, 2015, MKD/NW-15-0035L). The level of detail provided in this response is considered not appropriate for inclusion into the DCD.

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There is no impact on the Technical Specification.



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Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Figure 3.9-8 provides an overview of the reactor internals arrangement. A DVI nozzle is shown in the figure, but no description is provided in DCD Tier 2, Section 3.9.5. The applicant is requested to provide information on the impact of the core support barrel due to DVI nozzle injection, specifically, the temperature difference between the injection water and the normal operating coolant temperature, its impact to the core support barrel stresses, and which operating transients listed in DCD Tier 2, Table 3.9-1 would cause a DVI nozzle injection.

Response

For a DVI nozzle injection signal initiation occurring as a result of an operating transient specified in DCD Tier 2 Table 3.9-1 (i.e., non-Level D service conditions), the reactor coolant system (RCS) pressure will be higher than the DVI pipe line pressure. Therefore, the DVI line water will not be injected into the RPV and the core support barrel (CSB) will not be affected.

In the event of a Level D service condition, including LOCA and steam line break (SLB), the DVI nozzle injection into the RPV may be possible. For Level D conditions, the ASME Code, Section III, NG-3225 and Appendix F, requires that only the primary stress intensity be evaluated; the impact of the temperature difference between the injection water and the


normal operating coolant temperature on the CSB is not considered in Level D conditions according to the corresponding requirements of the ASME Code. Therefore, even if a DVI nozzle injection occurs, the integrity of the CSB is maintained.

Impact on DCD There is no impact on the DCD. Impact on PRA There is no impact on the PRA. Impact on Technical Specifications There is no impact on the Technical Specifications. Impact on Technical/Topical/Environmental Reports There is no impact on any Technical, Topical or Environmental Reports.

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Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Figure 3.9-8 provides an overview of the reactor internals arrangement. A core stop is shown in the figure, but no description is provided in DCD Tier 2, Section 3.9.5. The applicant is requested to provide information for the core stop, its design and intended function, its quantity and location, and the means by which the core stop is attached. In addition, the applicant is requested to identify under what accident conditions or operating transients is the core stop expected to function. A summary of this information should be included in the DCD.

Response

The intended function of the core stop is to sustain the load of the core and internals in the event of a hypothetical failure of the normal core support, to maintain the core cooling capability and provide sufficient reactivity control by limiting the displacement of the core and internals. Therefore, the core stop is designed to sustain an impact load resulting from a drop of the internals along with the static weight. Six (6) core stops are evenly located around the inside circumference, 60° apart from each other, of the reactor vessel bottom head. Each core stop is attached to the reactor vessel bottom head inside surface by a full penetration weld. The core stops are expected to function as a suitable redundancy for core support during

Non-Proprietary

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any accident involving the hypothetical catastrophic failure of the core support. The explanation provided is considered excessive to be specifically included in the DCD.

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03.09.05-13 - 1 / 2 KEPCO/KHNP




Docket No. 52-046

RAI No.: 92-8068





GDC 1 and 10 CFR 50.55a require that reactor internals be designed to quality standards commensurate with the importance of the safety functions performed. Standard Review Plan Section 3.9.5, “Reactor Pressure Vessel internals,” Areas of Review 2 includes the basis for the design of the reactor internals, loading conditions of normal operation, anticipated operational occurrences, potential adverse flow effects of flow-excited vibrations and acoustic resonances, postulated accidents, and seismic events.

The applicant is requested to provide information regarding any method different from DCD Tier 2, Section 3.9.3 that are used to determine the loads listed in DCD Tier 2, Section 3.9.5.2 for reactor internals components under design loading and Service Levels A, B, C, and D loading conditions. A summary of this information should be included in the DCD.

Response

The method used to determine the loads listed in DCD Tier 2, Section 3.9.5.2 for the reactor internal components under design loading and Service Levels A, B, C, and D loading conditions is the same as the method used in DCD Tier 2, Section 3.9.3.

ANSI/ANS 51.1, Section 3.4.1 provides the nuclear safety criteria for Safety Class 1, 2, and 3 mechanical equipment. An example of a limiting set of normal operations and events is given in ANSI/ANS 51.1, Table 3-3. The correlation between plant condition categories, nuclear safety functions, and ASME Section III service limits is given in ANSI/ANS 51.1 Table 3-6.

The guidance in Appendix A of SRP 3.9.3, which is also applicable to core support structure, provides the appropriate loading combinations under Service Levels A, B, C, and D loading conditions.

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Therefore, the method of combining the various loads is based on ANSI/ANS 51.1 and Appendix A of SRP 3.9.3. Since there are no differences between Sections 3.9.3 and 3.9.5.2 in the methods for determining the loads, no additional information needs to be included in the DCD.

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03.09.05-14 - 1 / 2 KEPCO/KHNP




Docket No. 52-046

RAI No.: 92-8068






Based on the information provided in DCD Tier 2, Section 3.9.5.2.3 for Level B service loading, it is unclear to the staff whether the loss of external load with turbine control system failure is an upset event or emergency event. According to DCD Tier 2, Table 3.9-1, loss of external load is upset event 2; however, Section 3.9.5.2.3 states that the loss of external load is an emergency event. Therefore, the applicant is requested to clarify this discrepancy. A summary of this information should be included in the DCD.

Response

In DCD Tier 2, Section 3.9.5.2.3, the loss of external load with a turbine control system failure is evaluated to occur once during the plant lifetime which is considered an emergency event based on the frequency of occurrence (e.g., one cycle). However, for conservatism, this event is considered as a Level B service loading only to evaluate fatigue with this combination of loadings.

DCD Tier 2, Table 3.9-1 describes the events commonly applicable to all of the major components. The upset event 2 of DCD Tier 2, Table 3.9-1 includes the loss of external load with a turbine control system failure event and is conservatively applied to the fatigue evaluations for the reactor internals. The loss of external load for an upset event 2 (total of 20 cycles) can be divided into two categories consisting of: 1) the loss of external load with turbine control function (19 cycles), and 2) the loss of external load with a turbine control

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system failure (1 cycle). The detailed information is provided in the design specification for reactor internals. The explanation provided is considered excessive to be specifically included in the DCD.

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Non-Proprietary

03.09.05-15 - 1 / 2 KEPCO/KHNP




Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Section 3.9.5.2.4, is inconsistent with DCD Tier 2, Table 3.9-1, which states that there are no events classified as a Service Level C condition. DCD Tier 2, Section 3.9.1.1.2, item (r) addresses the failure of small lines outside containment, and DCD Tier 2, Table 3.9-1 includes Upset Event-6 that includes this item. These events appear to be similar to the DBPB load but are categorized as a Service Level B event. Therefore, the applicant is requested to address the apparent inconsistency in the DCD between Tier 2, Section 3.9.1 and Tier 2, Sections 3.9.4 and 3.9.5, which both describe DBPB loads. A summary of this information should be included in the DCD.

Response

The event classification of DCD Tier 2, Table 3.9-1 depends on the design characteristics of each plant and is not related with the requirements of ASME Code Section III, NCA-2142.1. However, DCD Tier 2, Section 3.9.5.2.4 includes DBPB loads as a Service Level C condition.

The DBPB event considered in Level C service loading in DCD Tier 2, Section 3.9.5.2.4 is categorized conservatively as a Level B service condition (Decrease in Reactor Coolant System Inventory) in Table 3.9-1. Therefore, DCD Tier 2, Table 3.9-1 does not include DBPB events as a Level C service condition. However, according to Appendix A of SRP 3.9.3, which is also applicable to the core support structure, it is required that DBPB loads be considered in Level C service conditions. Thus, for the ASME Code Class CS components in DCD Tier 2,

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Section 3.9.5.2, DBPB loads are considered as a Level C service condition as a mechanical load. The information above is considered too detailed to be specifically included in the DCD.

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Non-Proprietary

03.09.05-16 - 1 / 2 KEPCO/KHNP




Docket No. 52-046

RAI No.: 92-8068






The staff reviewed the service loads described in DCD Tier 2, Section 3.9.5 in comparison to the transients presented in DCD Tier 2, Table 3.9-1 and found several apparent discrepancies in the description and categorization of these events. For example, based on the information provided in DCD Tier 2, Table 3.9-1, a main steam pipe break and a main feedwater pipe break are included in faulted events 1 and 2. It is, however, unclear to the staff whether the other faulted events listed in DCD Tier 2, Table 3.9-1 are included in the Service Level D loads for reactor internals. Therefore, the applicant is requested to clarify this discrepancy and describe and justify any differences between Table 3.9-1 and the loads applied to the reactor internals for all service levels. A summary of this information should be included in the DCD.

Response

For Level A service loading, DCD Tier 2, Section 3.9.5.2.2 includes all the system operating transient loads from normal events presented in Table 3.9-1.

For Level B service loading, DCD Tier 2, Section 3.9.5.2.3 includes all the system operating transient loads from upset events presented in Table 3.9-1. Additionally, the loss of external load with turbine control failure is considered a Level B service loading to evaluate fatigue with this combination of loadings for conservatism. (See the response to Question 03.09.05-14.)

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For Level C service loading, there are no system operating transient loads from emergency events in Table 3.9-1. However, DCD Tier 2, Section 3.9.5.2.4 considers DBPB loads as a mechanical load for conservatism. (See the response to Question 03.09.05-15.)

For Level D service loading, DCD Tier 2, Section 3.9.5.2.5 includes all the system operating transient loads from faulted events presented in Table 3.9-1. Specifically, the main steam/feed water pipe break (MS/FWPB) or LOCA loads are larger than the loads from faulted event-3, faulted event-4, and faulted event-6 in Level D service loadings. Additionally, IRWST discharge loads are considered as a faulted event-5. The explanation provided is considered excessive to be specifically included in the DCD.

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03.09.05-17 - 1 / 3 KEPCO/KHNP



Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Section 3.9.5.3 states that the reactor internals are designed to meet interface cold gaps between reactor internals and the reactor vessel and between the main parts of the reactor internals. The applicant is requested to provide information on the design of the reactor internals to accommodate a hot gap at their operating temperature and pressure. A summary of this information should be included in the DCD.

Response

The reactor internals are designed to have no interferences between the reactor internals and the reactor vessel. The minimum hot gap between the core support barrel outlet nozzle and the reactor vessel outlet nozzles is summarized below.

No. Operating Condition Minimum Maximum 1 During installation(cold gap) 0.104 inch 0.136 inch 2 During normal operation(hot gap) 0.027 inch 0.090 inch 3 During heat-up(hot gap) 0.006 inch 0.069 inch 4 During cool-down(hot gap) 0.159 inch 0.222 inch

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The cold gap between the core support barrel outlet nozzle and the reactor vessel outlet nozzles is calculated considering dimension and positional tolerances according to the requirements for alignment.

The hot gap between the core support barrel outlet nozzle and the reactor vessel outlet nozzles is calculated considering radial growth and contraction of the reactor vessel and reactor vessel internals according to the temperature and pressure as provided below.

(1) Radial clearance between the core support barrel and the reactor vessel outlet nozzle during installation

(2) Radial growth of the reactor vessel due to pressure

a = Inner radius of reactor vessel

b = Inner radius of reactor vessel + reactor vessel thickness(including clad thickness of 0.125 inch)

PRV = Pressure of reactor vessel at normal operating condition

ERV = Moduli of elasticity for reactor vessel

ν = Position ratio

(3) Radial growth of the reactor vessel due to temperature

RVIDnoz= Reactor vessel inside diameter at Nozzle

αRV = Reactor vessel coefficient of thermal expansion

TRV = Temperature of reactor vessel in nozzle shell region

(4) Radial contraction of the core support barrel due to pressure

PCSB = Pressure of core support barrel at normal operating condition

CSBIR = Inner radius of core support barrel

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TS

TS

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ECSB = Core support barrel moduli of elasticity

CSBt =Thickness of core support barrel

(5) Radial growth of the core support barrel due to temperature

CSBODnoz= Core support barrel inside diameter at Nozzle

αCSB = Core support barrel coefficient of thermal expension

TCSB = Temperature of core support barrel in nozzle shell region

(6) Effect of the alignment key to keyway gap = 0.016 inch

The nozzle clearance between the core support barrel and the reactor vessel outlet nozzle during operation is determined as follows:

Hot gap = (1) + (2) + (3) + (4) – (5) – (6)

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03.09.05-18 - 1 / 2 KEPCO/KHNP




Docket No. 52-046

RAI No.: 92-8068





GDC 1 and 10 CFR 50.55a require that SSCs important to safety be designed to quality standards commensurate with the importance of the safety functions performed. DCD Tier 2, Table 3.2.1 (page 56 of 86) lists the core support structures as safety Class-3, Quality Group C, seismic Category I with full compliance of 10 CFR 50 Appendix B quality assurance requirement. Note N-2 states that only those core support structures necessary to support and restrain the core and to maintain safe shutdown capability are classified as seismic Category I.

It is unclear to the staff that whether note N-2 from DCD Tier 2 Table 3.2-1 encompasses all core support structures, and if there are any core support structures that are not within the scope of note N-2, that is, that are not classified as seismic Category I. In addition, DCD Tier 2, Table 3.9-12 and Table 3.2.1 do not provide any safety class, quality group, seismic category classification, or quality assurance requirement for reactor internal structures other than core support structures. Therefore, the applicant is requested to provide information for these issues. A summary of this information should be included in the DCD.

Response

DCD Tier 2, Section 3.9.5 states that reactor pressure vessel internals, described as reactor internals in this subsection, refers to the core support structures and internal structures.

All of the core support structures within the scope of note N-2 are classified as seismic Category I. DCD Tier 2, Table 3.9-12 provides some classification for all of the reactor internals including internal structures, but is not a complete classification list. The complete detailed classification is provided in DCD Tier 2, Table 3.2-1. Currently, Table 3.2-1 provides the safety class, quality group, seismic category classification, and quality assurance requirement for core support structures. The NRC issued RAI 30-7927on June 15, 2015 pertaining to issues with Table 3.2-1 and System Quality Group Classification related to Reg.

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Guide 1.26. Efforts are currently being taken to address the noted classification issues and to provide an update to the table which will include core support structures and internal structures.

Impact on DCD

The safety class, quality group, seismic category classification, or quality assurance requirement for internal structures will be added in DCD Tier 2, Table 3.2-1 revised as the issues for RAI 30-7927(System Quality Group Classification) are resolved.

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03.09.05-19 - 1 / 3 KEPCO/KHNP

Non-Proprietary




Docket No. 52-046

RAI No.: 92-8068






DCD Tier 2, Section 3.9.5.3 states that to properly perform their functions, the reactor internals are designed to meet the following deformation limits:

1. Under Level A, Level B and Level C service loadings, the core is held in place, anddeflections are limited so that the CEAs can be inserted under their own weight as theonly driving force.

2. Under Level D service loadings that require CEA insertability, deflections are limited sothat the core is held in place, adequate core cooling is preserved, and all CEAs can beinserted. Those deflections that would influence CEA movement are limited to less than80 percent of the deflections required to prevent CEA insertion.

The applicant establishes allowable deformation limits as 80 percent of the loss-of-function deflection limits. The applicant is requested to explain and justify why the use of 80 percent as the loss-of-function deflection limit is acceptable. The applicant is also requested to provide references for this justification. In addition, the applicant is requested to clarify to which reactor internals components this deflection limit is applicable. A summary of this information should be included in the DCD.

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Non-Proprietary

Response

The allowable deflections are based on the deflection that causes a loss of function or loss of insertability. The insertion criteria are based on the reactor internals providing a clear straight path for the CEA.

This allowable deflection is conservatively assumed to be 80% of the deflection required to prevent CEA insertion which is based on dimensions considering tolerance, true position, etc. For service level D conditions, the resulting calculated deflections due to various loads are compared to allowable deflections equal to 80% of the deflection permitting a free path for CEA insertion. Justification for allowing 80% of the loss-of-function deflection limit is that it provides adequate margin consistent with that for structural failures when considering the uncertainty in the analytical process.

Therefore, the allowable deflection limit is established as 80% of the loss-of-function deflection limit under Level D conditions and is a conservative requirement for successful CEA insertion, in addition to meeting the ASME Code, Section III, Subsection NG stress limits. This deflection limit is applicable to the IBA Top plate, CEA shroud tube, UGS support plate, CEA guide tube, and fuel alignment plate related to the lateral deflection of the CEA guide path.

The loss-of-function deflection limits of the reactor internal components are the same as the minimum gaps which are calculated using the worst case dimension and tolerances from manufacturing and installation.

The calculated deflection amount is provided below. Analysis of the results shows that the calculated deflection of the reactor internals components is less than the allowable deflection limit. Therefore, 80% is to allow for adequate margin.

The information provided is considered excessive to be specifically included in the DCD.

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03.09.05-20- 1 / 2 KEPCO/KHNP

Non-Proprietary



Docket No. 52-046

RAI No.: 92-8068





DCD Tier 2, Section 3.9.5.3 states that in the design of critical reactor internals that are subject to fatigue, stress analysis is performed using the design fatigue curve of Figure I-9.2 of ASME BPV Code, Section III. It is unclear which components are considered “critical” reactor internals and are subjected to fatigue analysis. Therefore, the applicant is requested to clarify in the DCD which reactor internals components are analyzed for fatigue.

Response

The critical reactor internals that are subjected to fatigue, stress analysis are the core support structures stated in DCD Tier 2, Section 3.9.5.1 and also any specific location where welding and structural discontinuities may create stress concentrations.

The core support structures include the following components:

- Core support barrel

- Lower support structure

- Upper guide structure barrel assembly

Impact on DCD There is no impact on the DCD.

Impact on PRA There is no impact on the PRA.

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Non-Proprietary

Impact on Technical/Topical/Environmental Report There is no impact on any Technical, Topical and Environmental Reports.

Impact on Technical Specifications There is no impact on the Technical Specifications.

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