INPRO ASSESSMENT OF INDONESIA'S FUEL FABRICATION FACILITY IN AREA OF SAFETY

Transcription

1 IAEA INPRO TM to Review the Updating of the INPRO Methodology November 2016, Vienna, Austria INPRO ASSESSMENT OF INDONESIA'S FUEL FABRICATION FACILITY IN AREA OF SAFETY

2 NESA OF INDONESIA 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 What has been done Training of NESA (2011) NESA for one large LWR ( ). Report submitted in 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) 11/22/2016 Badan Tenaga Nuklir Nasional 2

3 Assessment of Nuclear Fuel Cycle Facilities: Indonesia s Experience 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) 11/22/2016 Badan Tenaga Nuklir Nasional 3

4 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 11/22/2016 Badan Tenaga Nuklir Nasional 4

5 Assessment of Nuclear Fuel Cycle Facilities: Indonesia s Experience 11/22/2016 Badan Tenaga Nuklir Nasional 5

6 NESA Programme in Indonesia Supporting National Program Fuel Cycle Roadmap Rev 0 Limited NESA for existing NFCF Fuel Cycle Roadmap Rev 1 Revised Fuel Cycle Roadmap as necessary NES #3: SMR2 (RDE) + NFCF NESA for SMR2 (RDE) NESA for fuel fabrication facility NES #2: SMR1 + NFCF Limited scope NESA for SMR1 (Eco, Inf) Full scope NESA for SMR1 NES #1: Large Reactor LR1 + NFCF 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 11/22/2016 Badan Tenaga Nuklir Nasional 6

7 Assessment of Nuclear Fuel Cycle Facilities DATABASE for reference fuel cycle facilities Limited Assessment of Existing Nuclear Fuel Cycle Pilot Facilities Safety Assessment of Fuel Fabrication Facility Development of Fuel Cycle Roadmap 11/22/2016 Badan Tenaga Nuklir Nasional 7

8 Safety Assessment of Fuel Fabrication Facility: 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 UO 2 /y To adopt international standards, guides and codes To adopt state of the art equipment, instrumentation and control Planned, Existing and Reference Facilities Existing Facilities in Serpong (Established in 1980s) Conversion of UF 6 to UO 2 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 Safety assessment is performed on existing facility (hence called assessed facility) against reference facilities 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) 11/22/2016 Badan Tenaga Nuklir Nasional 8

9 Nuclear Fuel Cycle Facilities in Indonesia Mining and Milling, Fuel Fabrication, Research Reactors, Waste Management Mining exploration being performed in Kalimantan, Sulawesi and Irian. Milling at lab scale. Mining excavation Pilot Plant for Conversion, Fabrication, PIE 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. 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 Reactor technology and nuclear safety Management of radwaste and research reactor spent fuel 11/22/2016 Badan Tenaga Nuklir Nasional 9

10 Fuel fabrication facility in serpong The facility is to produce fuel bundle containing UO 2 pellets in zircaloy. The facility is designed to handle enriched U-235 up to 5%. UO 2 Powder UO 2 Pellet s Fuel bundle Process among others: mixing, sieving, compacting, sintering, grinding, fuel filling, welding for pin and bundle, passivation 11/22/2016 Badan Tenaga Nuklir Nasional 10

11 Database of Reference Facility: JSC Mashinostroitelny Zavod 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 11/22/2016 Badan Tenaga Nuklir Nasional 11

12 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 H 2 gas for sintering 11/22/2016 Badan Tenaga Nuklir Nasional 12

13 Fuel assembly - JSC Mashinostroitelny Zavod Visit to production unit PWR fuel assembly Process is automatic, including for quality control. 11/22/2016 Badan Tenaga Nuklir Nasional 13

14 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 11/22/2016 Badan Tenaga Nuklir Nasional 14

15 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 11/22/2016 Badan Tenaga Nuklir Nasional 15

16 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. 11/22/2016 Badan Tenaga Nuklir Nasional 16

17 CR1.2 subcriticality margins IN1.2: Sub criticality margins. AL1.2: Sufficient to cover uncertainties and avoid criticality. The INPRO methodology requires that the K eff value be calculated for all possible configurations and the value should be < 0.90 [Tecdoc 1575 vol 9]. The calculated K eff values for all possible configurations at assessed facility range from Conclusion: The subcriticality margins are sufficient to cover uncertainties. CR1.2 is met. 11/22/2016 Badan Tenaga Nuklir Nasional 17

18 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 11/22/2016 Badan Tenaga Nuklir Nasional 18

19 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. 11/22/2016 Badan Tenaga Nuklir Nasional 19

20 CR1.5 failures and AOO IN1.5: Expected frequency of failures and AOO. AL1.5: Superior to a 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. 11/22/2016 Badan Tenaga Nuklir Nasional 20

21 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 peryear 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 11/22/2016 Badan Tenaga Nuklir Nasional 21

22 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. 11/22/2016 Badan Tenaga Nuklir Nasional 22

23 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. 11/22/2016 Badan Tenaga Nuklir Nasional 23

24 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. Conclusion: CR 2.2 is not met. 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. 11/22/2016 Badan Tenaga Nuklir Nasional 24

25 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. Consist of 5 criteria 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 11/22/2016 Badan Tenaga Nuklir Nasional 25

26 CR3.1 frequency of DBA IN3.1: Calculated frequency of occurrence of DBA. AL3.1: Superior to a 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. 11/22/2016 Badan Tenaga Nuklir Nasional 26

27 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. Conclusion: CR 3.2 is conditionally met. 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. 11/22/2016 Badan Tenaga Nuklir Nasional 27

28 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. 11/22/2016 Badan Tenaga Nuklir Nasional 28

29 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. 11/22/2016 Badan Tenaga Nuklir Nasional 29

30 CR3.5 robustness of containment design IN3.5: Containment loads covered by design of NFCF assessed. AL3.5: Superior to a 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. 11/22/2016 Badan Tenaga Nuklir Nasional 30

31 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. Consist of 3 criteria CR4.1: in-plant severe accident management CR4.2: frequency ofemergencyrelease intoenvironment CR4.3: source term of emergency release into environment 11/22/2016 Badan Tenaga Nuklir Nasional 31

32 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. 11/22/2016 Badan Tenaga Nuklir Nasional 32

33 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. 11/22/2016 Badan Tenaga Nuklir Nasional 33

34 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 g uranium through the stack. The result shows radiologcal intake as low as 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 11/22/2016 Badan Tenaga Nuklir Nasional 34

35 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. Consist of 2 criteria CR5.1 independence of DID level CR5.2 minimization of hazards 11/22/2016 Badan Tenaga Nuklir Nasional 35

36 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 11/22/2016 Badan Tenaga Nuklir Nasional 36

37 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 11/22/2016 Badan Tenaga Nuklir Nasional 37

38 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 Consist of 2 criteria CR6.1 human factors CR6.2 attitude to safety 11/22/2016 Badan Tenaga Nuklir Nasional 38

39 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 11/22/2016 Badan Tenaga Nuklir Nasional 39

40 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 11/22/2016 Badan Tenaga Nuklir Nasional 40

41 UR7 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 Consist of 2 criteria CR7.1 RD&D CR7.2 safety assessmen 11/22/2016 Badan Tenaga Nuklir Nasional 41

42 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 11/22/2016 Badan Tenaga Nuklir Nasional 42

43 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 There is no safety assessment submitted to the regulator for design of planned facility. Conclusion. CR7.2 not met 11/22/2016 Badan Tenaga Nuklir Nasional 43

44 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. 11/22/2016 Badan Tenaga Nuklir Nasional 44

45 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 11/22/2016 Badan Tenaga Nuklir Nasional 45

46 THANK YOU 11/22/2016 Badan Tenaga Nuklir Nasional 46