Licensing Issues and the PIRT

Transcription

1 Licensing Issues and the PIRT Frederik Reitsma IAEA Course on High temperature Gas Cooled Reactor Technology Oct 22-26, 2012

2 Content / Overview A few ideas to stimulate discussions: Safety assessment criteria Safety analysis Treatment of uncertainties NRC Advanced Reactor Policy Evaluation models PIRT Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 2

3 Safety assessment Combination of both deterministic and probabilistic methods South African National Nuclear Regulator has set the following limits: Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 3

4 Safety analysis Combination of both best estimate and conservative deterministic models and analysis best estimate analysis, the best estimate material properties and plant parameters are used as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion. Best estimate analyses are used: to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational Occurrences (AOOs) and Design Basis Accident (DBAs). to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs) to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones. Conservative analyses are used: To demonstrate compliance with regulatory limits for AOOs and DBAs Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 4

5 Treatment of uncertainties Combination of both best estimate and conservative deterministic models and analysis best estimate analysis, the best estimate material properties and plant parameters are used as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion. Best estimate analyses are used: to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational Occurrences (AOOs) and Design Basis Accident (DBAs). to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs) to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones. Conservative analyses are used: To demonstrate compliance with regulatory limits for AOOs and DBAs Methodology based on the Code Scaling, Applicability, and Uncertainty (CSAU) process has been adopted to quantify uncertainties in best-estimate calculations Can by used to show the margin of conservative models and analysis The GRS-SUSA code to be used as point of departure. A new IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis is being performed. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 5

6 NRC Advanced Reactor Policy Statement (1/2) Among the attributes which could assist in establishing the acceptability or licensability of a proposed advanced reactor design, and which therefore should be considered in advanced designs, are: Highly reliable and less complex shutdown and decay heat removal systems. The use of inherent or passive means to accomplish this objective is encouraged (negative temperature coefficient, natural circulation). Longer time constants and sufficient instrumentation to allow for more diagnosis and management prior to reaching safety system challenge and/or exposure of vital equipment to adverse conditions. Simplified safety systems which, where possible, reduce required operator actions, equipment subjected to severe environmental conditions, and components needed for maintaining safe shutdown conditions. Such simplified systems should facilitate operator comprehension, reliable system function, and more straight-forward engineering analysis for analysis. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 6

7 NRC Advanced Reactor Policy Statement (2/2) Designs which minimize the potential for severe accidents and their consequences by providing sufficient inherent safety, reliability, redundancy, diversity, and independence in safety systems. Designs that provide reliable equipment in the balance of plant, (or safety-system independence from balance of plant) to reduce the number of challenges to safety systems. Designs that provide easily maintainable equipment and components. Designs that reduce radiation exposure to plant personnel. Designs that incorporate defense-in-depth philosophy by maintaining multiple barriers against radiation release, and by reducing the potential for consequences of severe accidents. Design features that can be proven by citation of existing technology or which can be satisfactorily established by commitment to a suitable technology development program. FR Vol 73 No. 199, pg , Oct. 14, 2008 Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 7

8 Evaluation Models Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 8

9 PBMR Evaluation Models Pre-Break Focuses on the expected pre-break conditions, just before a break in the helium pressure boundary occurs. The time phase for this Evaluation Model ends when such a break occurs. It does not include any of the phenomena that might occur during a pressure boundary break or later. Software used: VSOP, MCNP, FLOWNEX, Fluent, NobleG, FIPREX/GETTER, RADAX. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 9

10 PBMR Evaluation Models Reactivity transients Calculates the transient reactor response for reactivity transient scenarios. Number of calculation models: 2 Software used: VSOP, TINTE. Thermal transients Calculates the transient reactor temperatures for forced cooling and loss of forced cooling scenarios. Number of calculation models: 2. Software used: TINTE, FLOWNEX. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 10

11 PBMR Evaluation Models Normal operation release Calculates the release of activity during normal operation as a result of Helium Pressure Boundary leakage. Number of calculation models: 22 Software used: VSOP, MCNP, FLOWNEX, ASTEC, Fluent, NobleG, FIPREX/GETTER, RADAX. Maintenance Dose Calculates the dose received by maintenance workers during maintenance periods due to the dust in the MPS and the activation of components that took place during normal operation. Number of calculation models: 15 Software used: MCNP, RADAX, SCALE, FISPACT, MicroShield. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 11

12 PBMR Evaluation Models Overall Worker and Public Dose Calculates the doses received by the public and worker for Category A, B and Beyond B accidents that involve a breach in the pressure boundary. Provides the source term for each of these accidents as well as the resulting doses. Provides source terms that are used as an input into the Probabilistic Risk Assessment (PRA) for Category C events. Number of calculation models: 28 Software used: VSOP, TINTE, MCNP, ASTEC, FLOWNEX, Fluent, PC COSYMA, NobleG, FIPREX/GETTER, RADAX. Main components: Pre-break, Initial release, Delayed release, Air ingress, Confinement, Atmospheric dispersion Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 12

13 PIRT Phenomena Identification Ranking Tables Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 13

14 PIRT background Phenomena Identification and Ranking Tables are increasingly used in the nuclear industry. PIRT were initially used to identify the thermal-hydraulic processes that were most important for a safety analysis computer code to simulate with acceptable accuracy, so that a limited set of sensitivity analysis could be performed to help quantify the uncertainty in the safety analysis results. PIRT are now recognized as a valuable tool to help prioritise efforts associated with safety analysis, development and assessment of codes and models, and specification of scaling or other requirements for tests and experiments. There is limited industry knowledge and experience with HTGR accident analysis, relative to that for LWRs. Since operating history is not available to provide the same valuable data, an accepted and auditable method of making early decisions related to analysis is needed. The PIRT process can be used as: a tool and a guide to help prioritise software and model V&V efforts, to agree on the necessary degree of conservatism to include in analysis assumptions and initial conditions, and to focus the available resources on the phenomena believed to be most important to the safety analysis process. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 14

15 PIRT - Process Process: Step 1. Define the PIRT objectives and plant design Step 2. Define the accident or transient scenario Step 3. Define figures of merit Step 4. PIRT team review available data Step 5. Partition scenario into convenient time phases Step 6. Identify involved and affected SSC Step 7. Identify phenomena by time phase and SSC Step 8. Rank importance of components and phenomena with confidence levels Step 9. Finalize and document PIRT for subject scenarios and plant designs Repeat periodically Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 15

16 PIRT Status Decision Chart Confidence Confidence Rank in Rank in Value Status Symptom Action required (Sure / (Sure / (High / Low) Unsure) Unsure) 8 High Unsure Unsure Phenomenon is perceived as significant but is not well known. High priority requirement for analysis and validation. 7 High Sure Unsure Phenomenon is significant and confidence in value is low. High priority requirement for validation. 6 High Unsure Sure 5 High Sure Sure 4 Low Unsure Unsure 3 Low Sure Unsure 2 Low Unsure Sure 1 Low Sure Sure Phenomenon is significant and the confidence in rank is low. Phenomenon is significant and well known. Phenomenon is not significant but not well known. Phenomenon is not significant and the confidence in value is low. Phenomenon is not significant and the confidence in rank is low. Phenomenon is well known and is not significant. High priority requirement for analysis. Should be well represented in the model. Should be readily validated. Requires analysis and validation to determine rank and value. Low priority requirement for validation. Low priority requirement for analysis. May be modelled without validation. Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 16

17 PIRT - examples Source Term (Pre-Break) Reactivity transients Thermal transients Overall Worker and Public Dose PIRTs Initial release Delayed release Air Ingress Confinement Atmospheric Dispersion Non-MPS leaks and spills Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 17

18 PIRT example Transport phenomena (from pre-break PIRT) Radionuclide plateout: Adsorption/ desorption processes Penetration/evaporation Diffusion into material Chemical characteristics of radionuclide Laminar or turbulent flow Dust deposition and lift-off: Agglomeration of dust particles Brownian diffusion Electrostatic forces Inertial separation Laminar or turbulent flow Saffman lift force Sedimentation Thermal gradient (thermophoresis) Plant operational transients Plant vibration Monolayer or multilayer resuspension Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 18

19 Ref Number RT-RU-001 RU RT-RU-002 RU Oct 22-26, 2012 Reactivity Transient PIRT Example for the Reactor Unit System subsystem component Process Phenomena Pebble Bed Fuel Kernel Fission Fission Power production in the fuel Pebble Bed Fuel Kernel Fission Temperature influence on fission (doppler effect) RT-RU-003 RU Pebble Bed Fuel Pebble Fission Moderation - temperature dependence RT-RU-004 RU Pebble Bed Coolant Fission Nitrogen inventory reduction RT-RU-005 RU Pebble Bed Fuel Kernel Fission Non-local Power Production Rank (3 - high, 2 - medium, 1 - low) Confidenc e (Sure/ Unsure) state of knowledg e Rationale IAEA Course on High temperature Gas Cooled Reactor Technology Notes High Sure High The fission process provides the This is due to fission, primary source of power production in a reactor. High Sure High Kernel does get hot. Flux dependence on group structure is important. Temperature dependence of resonance cross sections important. High Sure High Moderator feedback effect is well known and temperature interaction with fission is well modelled High Sure Med Nitrogen is an absorber. The reduction in the nitrogen inventory could contribute to reactivity addition. Med Sure High The negative coefficient of reactivity dependence on temperature is assisted by nonlocal heating which makes this phenomenon important. RT-RU-006 RU Pebble Bed Fuel Kernel Fission Fuel Burn-up Low Sure High Low for perturbations in burn up from the anticipated core state. RT-RU-007 RU RT-RU-008 RU Pebble Bed Fuel Kernel Fission Spatial distribution of burn-up Low Sure High Low for perturbations in burn up from the anticipated core state. Pebble Bed Reactor core Fission Nuclei Breeding Low Sure High The total effect on the FOM is incl graphite minimal SSC Consider Multi-group vs 2-group neutron flux representation. The current 2 group misses out on resonance treatment. The physics is understood, but it comes down to modelling issues and complexity. The moderator feedback effect is known and temp interaction with fission is currently well modelled. This is particularly applicable to the start-up sequence, the the reduction in nitrogen could contribute to reduced absortion. The effect of SAS removal with nitrogen in the core needs to be checked to see if the core approaches criticality. The modelling of the scenario has never been attempted but physics capability is available in TINTE. The non-local power is as a result of absorption of gamma radiation, fast neutrons. Currently when modelling, the fission process is adequately captured, but gamma modelling is approximated. Refers to the average value of Fuel Burn-up (there will be differences when considering the start-up or equilibrium core) Currently the axial distribution within the core is modelled. Nuclei Breeding is captured in the characterisation of power production. The effect is predominantly considered over the whole core and not 19 to the kernel level.

20 PIRT Iterative process Positive outcomes: Good basis for identifying EM development requirements Easier to justify modelling assumptions Challenges: In a new technology experts are not readily available Spent a lot of time on irrelevant phenomena Ranking may be incorrect If nobody knows then typically have to spend a lot of time to find out but in the end this is positive Include external experts Improves credibility (but also complexity ) Perform hierarchical breakdown Revised ranking bins Difficult to decide what to do with medium bins Too many combinations of uncertainties available to adequately action resolution Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 20

21 Source material used: HTGR Technology Course for the Nuclear Regulatory Commission, May 24 27, 2010 HTR/ECS 2002 High temperature Reactor School, 2002 MUA 784: Reactor Physics, F Reitsma, Mechanical Engineering Post-Graduate: Nuclear Theme, University of Pretoria, 2012 Workshop at PHYSOR 2010 Advances in Reactor Physics to Power the Nuclear Renaissance: The Pebble Bed Modular Reactor: From V.S.O.P. (Very Superior Old Product) to Generation IV candidate. Safety Analysis Software Development and V&V, Peter Robinson, Workshop on Safety Aspects of Modular HTGRs, October 2007, Beijing China Radionuclide transport during normal operation conditions, Lize Stassen, Pieter Goede, Gen-IV CMVB Chemistry and Transport Workshop, Centurion, South Africa, January 2009 Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 21