inpro assessment of indonesia's fuel fabrication facility in area … · indonesia's fuel...

INPRO ASSESSMENT OF

INDONESIA'S FUEL FABRICATION FACILITY

IN AREA OF SAFETY

IAEA INPRO TM to Review the Updating of the INPRO Methodology

15-17 November 2016, Vienna, Austria

11/22/2016 Badan Tenaga Nuklir Nasional 2

Why NESA?

Targeting assessment of NES for both NPP and fuel cycle facilities

Aiming to support Indonesia’s nuclear power program, addressing

sustainability aspect and enhance awareness of long term

commitment to nuclear power, and identifying potential regional /

international arrangement

NESA already contributed to strategic planning 2015-2019

What has been done …

• Training of NESA (2011)

• NESA for one large LWR (2012-2014). Report submitted in 2015.

• NESA for one SMR (2015). Limited scope (Economics and

Infrastructure). Report on going

• Roadmap of Indonesia”s nuclear fuel cycle

• NESA for Fuel Cycle Facilities (2012- onward)

NESA OF INDONESIA


The experience …

• Targeting at assessment of mining and milling, conversion, enrichment, fuel fabrication, interim spent fuel storage, and final disposal facilities,

• No sufficient data on planned facilities: (i) Planned fuel fabrication at basic design stage; (ii) Pilot fuel fabrication in Indonesia does not match that for the assessed large NPP

• Difficulty to seek data of reference fuel cycle facilities, e.g. safety assessment report (SAR)

Assessment of Nuclear Fuel Cycle Facilities:

Indonesia’s Experience


Change of direction …

Develop database for reference LWR-related nuclear fuel cycle

facilities, i.e. technical data and services

Limited assessment of existing pilot facilities in Indonesia (Limited

assessment is focused on economics, waste management,

environment and safety; general overview for others areas)

Develop fuel cycle roadmap showing national nuclear energy

policies, possible scenarios to achieve the targets, milestones of

nuclear fuel cycle facilities development, assessment of

sustainability of the scenarios based on the INPRO Methodology,

and identification of necessity for R&D as well as pursuance of

regional / multilateral collaboration




NES #1: Large Reactor LR1 + NFCF

2012 2013 2014 2015 2016

Limited scope NESA for

SMR1 (Eco, Inf)

Fuel scope NESA for LR1

(Eco, Inf, WM, PR, PP, Env, Safety)

Pre- NESA for PWR NFCF/ Database

•Full scope NESA for LR1-based NFCF

NES #2: SMR1 + NFCF

NES #3: SMR2 (RDE) + NFCF

Fuel Cycle Roadmap Rev 0

Limited NESA for existing NFCF

Supporting National Program

•NESA for fuel fabrication facility

•Full scope NESA for SMR1

•Revised Fuel Cycle Roadmap as necessary

NESA for

SMR2 (RDE)

Fuel Cycle Roadmap

Rev 1

NESA Programme in Indonesia


Assessment of Nuclear Fuel Cycle Facilities

Safety Assessment

of Fuel Fabrication

Facility

DATABASE for

reference fuel cycle

facilities

Limited Assessment of

Existing Nuclear Fuel

Cycle Pilot Facilities

Development of Fuel

Cycle Roadmap


Reference Facilities

(1) Open Source: Columbia

Fuel Fabrication Facility, US

• Design capacity: 1500 t U/y

• Started production in 1969,

renewed license for 20-year

since 2007

(2) Site Visit: JSC

Mashinostroitelny Zavod Fuel

Production Facility, Russia

(part of NESA Workshop by

CICE&T)

Existing Facilities in Serpong

(Established in 1980s)

• Conversion of UF6 to UO2

at Fuel Element Production

Installation (for research

reactor fuel manufacture)

• Pelletization (Nat U) at

Experimental Fuel Element

Installation (for HWR fuel

manufacture). Design

capacity ~65 kg U/day,

able to handle enriched U.

• HWR Fuel assembly

Planned Facility • Designed by Center for

Nuclear Facilities

Engineering-BATAN

• Basic design completed in

2013 based on existing pilot

facility in Serpong

• Sited on Serpong

• Design capacity of 710 t

UO2/y

• To adopt international

standards, guides and codes

• To adopt state of the art

equipment, instrumentation

and control

Safety Assessment of Fuel Fabrication Facility: Planned, Existing and Reference Facilities

• Safety assessment is performed on existing facility (hence called assessed facility) against reference facilities

Mining exploration

being performed in

Kalimantan, Sulawesi

and Irian. Milling at lab

scale.

Indonesia manages RR

fuel production facility and

HWR fuel pilot facility in

Serpong. A pilot facility

for HTGR pebble fuel type

is being established in

Yogyakarta.

Nuclear Fuel Cycle Facilities in Indonesia


Reactor technology and

nuclear safety

Mining and Milling, Fuel Fabrication, Research Reactors, Waste Management

Pilot Plant for Conversion,

Fabrication, PIE

Mining excavation

Management of radwaste

and research reactor

spent fuel

Indonesia operates 3

research reactors, and

plans to deploy an

experimental power

reactor

Waste management centered

in Serpong: Waste treatment,

Interim storage and Planned

short-lived LLW demo

disposal


The facility is to produce fuel bundle

containing UO2 pellets in zircaloy. The facility

is designed to handle enriched U-235 up to

5%.

UO2

Pellet

s

Fuel

bundle

UO2

Powder

Process among others: mixing, sieving,

compacting, sintering, grinding, fuel filling,

welding for pin and bundle, passivation

Fuel fabrication facility in serpong

• Site visit to nuclear fuel production facility at JSC Mashinostroitelny Zavod – as part of NESA Workshop organized by CICE&T in Obninsk, Russia, 8-12 December 2014

• Overview of nuclear fuel production facility: PWR fuel (Russian, AREVA), CANDU fuel

• Production lines:

Fuel pellets

Fuel assembly

Control assembly

• Automatic quality control process in production line


Database of Reference Facility:

JSC Mashinostroitelny Zavod

Fuel Pellets - JSC Mashinostroitelny Zavod


• Process is similar to that in existing fuel fabrication facility in Serpong

• Unit has been undergone revitalization since 1996

• Largely automated included for quality control.

• Allowed production capacity at 204 pellets/min.

• Safety aspects are in place, e.g. in using H2 gas for sintering

Fuel assembly - JSC Mashinostroitelny Zavod


• Visit to

production unit

PWR fuel

assembly

• Process is

automatic,

including for

quality control.

SAFETY BASIC PRINCIPLE (BP)


INPRO basic principle for sustainability assessment in the area

of safety of NFCF:

Safety of planned NFCF should be superior compared against safety of

reference NFCF. In the event of an accident, off-site releases of

radionuclides and/or toxic chemicals should be prevented or mitigated so

that there will be no need for evacuation

• There is only 1 BP on draft compare against 4 on original

• There are only 7 URs on draft compare against 14 on original

• There are 22 CR on draft compare against 33 on original

UR1

• UR1 Robustness of design during normal operation: The

uranium or MOX fuel fabrication facility assessed should be

more robust relative to a reference design regarding system

and component failures as well as operation.

• Consists of 6 criteria

1. CR1.1 design of normal operation systems.

2. CR1.2 sub criticality margins

3. CR1.3 operation

4. CR1.4 inspection

5. CR1.5 failures and AOO.

6. CR1.6 occupational



CR1.1 design of

normal

operation

systems.

IN1.1:

Robustness of

design of normal

operation

systems.

AL1.1: Superior

to a reference

design.

Seismic load. The combined load design of assessed facility is

0.16 g below maximum ground acceleration 0.05 g at the site

meanwhile no earthquake data available for reference facility that

has 2% chance to hit 0.3 g in 50 y.

Flood/Water level. Both facilities have taken flooding into

account in the design.

Fire Prevention. The assessed facility has implemented similar

safety prevention system. Fire Hazard Analysis has been done

based on inspection. A separate more detailed document on Fire

Hazard Analysis needs to be prepared to show robustness.

Explosion Prevention. Both assessed and reference facilities have

necessary measures in place for several parameters. Further

comparative information is needed to prepare a more complete

assessment.

Conclusion: Robustness of the assessed fuel fabrication facility is considered

comparable to that of the reference plant. Assessed facility has an advantage having no

dam upstream compare than reference facility. CR1.1 is conditionally met.


CR1.2 subcriticality

margins

IN1.2: Sub criticality

margins.

AL1.2: Sufficient to cover

uncertainties and avoid

criticality.

The INPRO methodology requires that the Keff value

be calculated for all possible configurations and the

value should be < 0.90 [Tecdoc 1575 vol 9].

The calculated Keff values for all possible

configurations at assessed facility range from 0.85 –

0.87.

Conclusion: The subcriticality margins are sufficient to cover uncertainties. CR1.2 is

met.


CR1.3 operation

IN1.3: Quality

of operation.

AL1.3: Superior

to a reference

design

Degree of automation. The assessed facility (pelletizing) has degree of

automation much less than the reference facility in Russia.

Periodic and intensive training. Experiences gaining from existing

facility on training for operators, supervisors and radiation protection

personnel will ease their implementation on planned facility. Types of

training need to be extended as proposed in the NUREG, and details

such as periodicity are to be indicated.

Availability of manuals. Operations and emergency instructions

manuals conform to those at the reference plant and the IAEA Safety.

Periodic mock-ups. Emergency exercises have been carried out at the

existing facility. The reference plant maintains Emergency Management

Program which are annually reviewed at minimum.

Conclusion: CR1.3 is partially met


CR1.4 inspection

IN1.4: Capability to

inspect.

AL1.4: Superior to a

reference design.

Monitoring of radiation levels as stated on basic design

document of planned facility is comparable to the reference

plant. Such monitoring has been performed at existing facility.

Monitoring of pressure drop across HEPA filters is done once a

week, but no continuous monitoring of pressure drops across

HEPA filters as suggested by the INPRO Methodology. The

HEPA filters are replaced when pressure drops 650 Pa, but in

reference facility the more stringent criteria for replacing filters

are used; a routine schedule, airborne radioactive

concentrations, hood velocity, differential pressure and

particulate penetration

Continous monitoring of cooling water temperature which is

connected to alarm for furnaces is available .

Conclusion: Monitoring of radiation level is acceptable, however, continuous or online

monitoring is limitedly available such as for sintering process, but not for other system

such as HEPA filter. So, CR1.4 is partially met.


CR1.5 failures and

AOO

IN1.5: Expected

frequency of failures

and AOO.


reference design

There is no evidence that probabilistic as well as

deterministic analyses have been done to determine the

frequency of failure and AOO on planned as well as on

existing facility. But in the safety analysis report of existing

facility, the expected frequency of failures and disturbances

in the facility has been evaluated / determined by using the

qualitative method PHA.

The reference facility develops and maintains an Integrated

Safety Analysis (ISA) but there is no clear statement on

probabilistic or deterministic safety analysis. The safety

analysis method used in the facility is qualitative method

PHA.

Conclusion: No information of the frequency of failures and disturbances is

available at the assessed facility and the reference plant. Hence, no comparative

assessment can be made. So CR 1.5 is not met.


CR1.6 occupational

dose

IN1.6: Occupational

dose values during

normal operation and

AOO.

AL1.6: Lower than the

dose constraints

Due to processes characteristic involving fresh uranium on

existing and perhaps planned facility, dose accepted by

worker far less then dose constrain. Normally, dose accepted

by worker in existing facility is not more than 1.5 mSv per-

year far less than 50 mSv dose constrain. Data on

occupational dose values during AOO is not available

Conclusion: Occupational dose during normal operation is lower than the dose contrain

but no data available for the AOO. So CE1.6 is partially met

UR2

• UR2 Detection and interception of AOO and failures: The

fuel fabrication facility assessed should detect and intercept

deviations from normal operational states in order to prevent

AOO from escalating to accident conditions.

• Consist of 2 criteria

1. CR2.1 I&C systems

2. CR2.2 grace period after AOO and failures.



CR2.1 I&C systems

IN2.1: I&C system to

monitor, detect,provide

alarm and together

with operatoractions

intercept and

compensate AOO

andfailures.

AL2.1: Availability of

such systems

and/oroperator actions.

The assessed facility is equipped with control system with

instruments to monitor operation of ventilation, fire alarm,

personnel and work area (video) and audio communication.

I&C at the assessed facility including limits for alarms and

shutdown conditions for process equipment / evacuations

are in place to detect / intercept deviations in order to

deliver safe operations.

At the reference plant, a design philosophy that includes

I&C systems to monitor and control the behavior of Items

Relied on for Safety (IROFS) is implemented.

Conclusion: CR 2.1 is conditionally met.


CR2.2 grace period after

AOO and failures

IN2.2: Grace period until

human actions are

required after AOO and

failures.

AL2.2: Adequate grace

period is defined in

design analyses.

Disturbance in cooling water is critical to sintering

furnaces. In case of loss of power, back-up genset is

available to drive emergency pump to restore the

circulation. There are four chillers to supply cooling

water with one chiller designated for emergency. The

installation also has three water reservoirs.

Emergency back-up genset (1650 kV, 80 hours) will

start automatically in 15 seconds to serve safety

relevant loads.

No grace period has been determined in the safety

analysis report of assessed facility. No data was

available for the reference plant to allow for

comparative assessment.

Conclusion: CR 2.2 is not met.

UR3

• UR3 Design basis accidents: The frequency of occurrence of DBA in the fuel fabrication facility assessed should be reduced. If an accident occurs, engineered safety features and/or operator actions should be able to restore the facility assessed to a controlled state and subsequently to a safe state, and the consequences should be mitigated to ensure the confinement of nuclear and/or toxic chemical material. Reliance on human intervention in the facility assessed should be minimal, and should only be required after a sufficient grace period.


• CR3.1 frequency of DBA

• CR3.2 engineered safety features and operator procedures

• CR3.3 grace period for DBA

• CR 3.4 bariers

• CR 3.5 robustness of containment design



CR3.1 frequency of DBA

IN3.1: Calculated frequency

of occurrence of DBA.


reference design

The information indicates that analysis of calculated

frequency of DBA has not been performed at the

assessed facility, and data is also not available for the

reference plant.

Conclusion: CR 3.1. is not met. Data is inadequate to allow for comparative

assessment. More information on calculated frequency of DBA for fuel fabrication

facility must be sought.


CR3.2 engineered

safety features and

operator

procedures

IN3.2: Reliability

and capability of

engineered safety

features and/or

operator

procedures.

AL3.2: Superior to

a reference design.

The INPRO Methodology refers to engineered safety features

such as temperature control system to shutdown furnaces in

the event of loss of cooling water, and secondary ventilation

systems which would take over in the event of loss of a glove

box barrier .

The assessed facility has in place the followings :

- Redundancy in VAC and chilled water supply,

- Independency in VAC through isolation, physical separation

by distance, barrier, and layout configuration of process

components and equipment

- Diversity for power source for safety & non safety relevant

loads, available emergency power supply

- Fail-safe principle (H2 gas in sintering process)

The features available at the reference plant, e.g. secondary

ventilation system, emergency power supply. Further info is

needed from the reference plant for more detailed assessment.

Conclusion: CR 3.2 is conditionally met.


CR3.3 grace period for

DBA

IN3.3: Grace period for

DBA until human

intervention is necessary.

AL3.3: Increased relative to

a reference design.

Data was not available for the assessed plant. In the

reference plant, the procedure is to escape

immediately. Further info is required on grace

period for DBA.

Conclusion: CR 3.3. is not met. Data is inadequate to allow for comparative

assessment.


CR3.4 barriers

IN3.4: Number of

confinement barriers

maintained (intact) after a

DBA.

AL3.4: At least one.

One barrier, the containment/building with

ventilated system, remains intact in the facility

avoiding an emergency release of radioactivity

and/or toxic chemicals to the outside of the facility

after DBA.

Conclusion: CR3.4 is met.


CR3.5 robustness of

containment design

IN3.5: Containment loads

covered by design of NFCF

assessed.


reference design.

The last barrier is the containment/building which

has been designed to withstand external events like

winds, tornadoes, floods, seismics (0.16 g) and

internal loads like processes and devices loads and

combination loads (combination of permanent,

accidental, winds, thermal and seismic)

Conclusion: CR3.5 is met.

UR4

• UR4 Severe plant conditions: The frequency of occurrence

of emergency release of radioactivity into the environment

from the NFCF should be reduced. Source term of the

emergency release into environment should remain well within

the envelope of reference facility source term and should be so

low that calculated consequences would not require evacuation

of population.


• CR4.1: in-plant severe accident management

• CR4.2: frequency ofemergencyrelease intoenvironment

• CR4.3: source term of emergency release into environment



CR4.1: in-plant severe accident

management

IN4.1: Natural or engineered

processes, equipment, and AM

procedures and training to

prevent an emergency release to

the environment in the case of

accident

AL4.1: Sufficient to prevent an

emergency release to the

environment and regain control of

the NFCF

There is no evidence that procedures,

equipment and training sufficient to prevent

large release outside containment and regain

control of the facility are available. The

available organization, procedures, aquipment

and training are for accident prevention and

emergency preparedness purposes - not to

regain control of the facility.

Conclusion: CR4.1 is not met.


CR4.2: frequency of emergency release

into

environment

IN4.2: Calculated frequency of an

emergency release of radioactive

materials and/or toxic chemicals into the

environment.

AL4.2: Lower than in reference facility.

Calculated frequency of emergency

release into

environment only available for

scenario explosion of hydrogen gas in

the furnace. The such calculated

frequency shows the probability of the

release of uranium to the environment

through the stack is 10E-8 per even

well below 10E-6 per unit.-years.

Conclusion: Estimate of frequency of release of radioactivity to the environment was

available for one DBA only So CR4.2 is partially met.


CR4.3: source term of emergency release into

environment

IN4.3: Calculated inventory and

characteristics (release height, pressure,

temperature, liquids/gas/aerosols, etc) of an

emergency release

AL4.3: Should remain well within the

inventory and characteristics envelope of

reference facility source term and should be so

low that calculated consequences would not

require evacuation of population

Calculation has been performed

on worst scenario, i.e.

instantaneous release of 33.75 g

uranium through the stack . The

result shows radiologcal intake as

low as 0.0878 Bq at 500 m from

the stack.

Conclusion: The result shows that release consequence / dose is sufficiently low to

avoid necessity for evacuation. CR4.3 is met

UR5

• UR5 Independence of DID levels and inherent safety

characteristics: An assessment should be performed for the

fuel fabrication facility to demonstrate that the different

objectives of levels of DID are met and that the levels are more

independent from each other than in existing systems. To excel

in safety and reliability the facility assessed should strive for

incorporating into its design increased emphasis on inherently

safe characteristics.


• CR5.1 independence of DID level

• CR5.2 minimization of hazards



CR5.1 independence of

DID levels characteristics.

IN5.1: Independence of

different levels of DID in the

fuel fabrication facility

AL5.1: More independence

of the DID levels is

demonstrated compared to

the reference design, e.g.

through deterministic and

probabilistic means,

hazards analysis, etc.

Independence principle for engineered safety features

is implemented through isolation, physical separation

by distance, barrier, and layout configuration of

process components or equipments. VAC system is

physically separated for each work area. Normal power

sources and secondary power sources are also located

at different places (separated with walls). However,

there is no probabilistic, deterministic or hazard

analaysis for assessing independency of DID level

avalaible to assessor.

Conclusion: There is inadequate evidence to justify the sufficiency of the level of

independence for DID. CR5.1 is partially met


CR5.2 minimization

of

hazards

IN5.2: Examples of

hazards: fire,

flooding,

release of radioactive

material, radiation

exposure, etc.

AL5.2: Hazards

minimized according

to the

state of art.

The assessed facility employs administrative and operator

controls as well safety features, e.g.:

• On line controls (OLCs), e.g. control of pressure,

temperature, and flowrate

• Detector and burner in the case of H2 use in furnaces

• Use of containment for materials and process equipment:

building, room, glove box, and also ventilation

systemequipped with dampers to prevent spread of

contaminants / isolate the room.

• Storing and processing of fisile material has always

performed in sub-critical condition (Keff < 0.9).

The reference facility employs material possession limits,

strict control of combustible and flammable materials,

constraints on procurement, use, and transfer of nuclear

materials

Conclusion: There is evidence that administrative and safety features at assessed

facility in Serpong are comparable with those at the reference plant, but no justification

is available for superiority, so CR5.2 is not met

UR6

• UR6 Human factors related to safety: Safe operation of the

fuel fabrication facility assessed should be supported by taking

into account human factor requirements into design and

operation of the facility, and by establishing and maintaining

an adequate safety culture in all organizations involved in a

nuclear energy system


• CR6.1 human factors

• CR6.2 attitude to safety



CR6.1 human factors

IN6.1: Human factors

addressed systematically in

the life cycle of the fuel

fabrication facility.

AL6.1: Evidence available.

The followings are evidence of human factors at

existing facility:

• Performing human resources development and

implementing a behaviour based safety system

• Periodic traning to all personel

• Qualification scheme

• Work procedures

• Health check up

Human factors at reference facility is based on he

integrated Behavioral Safety and Human Performance

Program including Behavioral Safety Process, and

Human Performance Process concept and their

implementation.

Conclusion: Human factors have been addressed at Serpong facility, so CR1.6 is met


CR6.2 attitude to safety.

IN6.2: Prevailing safety

culture.

AL6.2: Evidence provided

by periodic safety reviews.

In the assessed facility, safety culture is recognized,

implemented and reviewed. Some evidence on the

implementation:

- Training of supervisor and all personnel on safety

culture

- Sharing implementation of safety culture

- Self assessment on safety culture

- Socialization and strengthening of individual

commitment to safety

Conclusion: CR6.2 is met

• UR7 R&D for innovative designs: The development of

innovative design features of the fuel fabrication facility

assessed should include associated research, development and

demonstration (RD&D) to bring the knowledge of facility

characteristics and the capability of analytical methods used

for design and safety assessment to at least the same

confidence level as for operating facilities


• CR7.1 RD&D

• CR7.2 safety assessmen


UR7


CR7.1 RD&D

IN7.1: RD&D status.

AL7.1: RD&D defined,

performed and

database developed.

There is no RD&D document related to safety

available to assessor. Planned facility is still in basic

design phase, but developer stated that the design will

follow standards, guidelines, prcedures and codes

related to safety. Since the planned facility will use

proven processes, technologies, materials and

components, then a pilot plant is not necessary.

Conclusion. CR7.1 is not met


CR7.2 safety assessment

IN7.2: Adequate safety

assessment.

AL7.2: Approved by a

responsible regulatory

authority

Licensing to operating non-reactor facility require SAR

approved by the regulator, BAPETEN. The existing

facility has received extension for operation licence

since 2013.

There is no safety assessment submitted to the

regulator for design of planned facility.

Conclusion. CR7.2 not met

Conclusions and Recommendations

Indonesia has benefited from NESA in raising awareness of

NES sustainability and following up actions needed to close

the gaps between the existing facilities and the planned

facilities.

The assessment, however, remains a challenge for newcomers

and even Member States with pilot / lab scale facilities:

• Data availability from overseas reference facility is limited in open

sources.

• Assessment cannot be done using data from conceptual design or small

scale facility, i.e. there are many gaps as consequences from nature of

the design and the methodology.

• CR6.2 can only be evaluated on operating facility because the

responsibility to fulfil the criterion is on operator not in the designer.


Conclusions and Recommendations

In terms of draft INPRO methodology, BPs, URs and CR are

simpler than the original version

INPRO may need to develop standardized reference NFC

Facility to improve quality of assessmen in such no reliable

data of reference plant are available


THANK YOU


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