content of slides MOTIVATION

- BACKGROUND OBJECTIVE

- THE ELEMENTS FOR THE PROPOSAL

- ALARA

- BEPU

- E-SM

- A

- ERT

- THE ADDITIONAL SAFETY BARRIER

- FINAL REMARKS

- APPENDICES (nuclear fuel & supporting references

D'Auria, F.

English

content of slides MOTIVATION
- BACKGROUND OBJECTIVE
- THE ELEMENTS FOR THE PROPOSAL
- ALARA
- BEPU
- E-SM
- A
- ERT
- THE ADDITIONAL SAFETY BARRIER
- FINAL REMARKS
- APPENDICES (nuclear fuel & supporting references

D'Auria F.

Archivio della Ricerca - Università di Pisa

AN ADDITIONAL SAFETY BARRIER FOR EXISTING AND NEW NPP F. D’Auria (with contribution by N. Debrecin  and H. Glaeser)  University of Pisa DESTEC-GRNSPG Nuclear Research Group in San Piero a Grado (Pisa) - Italy  LIST OF CONTENT      MOTIVATION    BACKGROUND              OBJECTIVE    THE ELEMENTS FOR THE PROPOSAL  ALARA  BEPU  E-SM  IA  ERT    THE ADDITIONAL SAFETY BARRIER    FINAL REMARKS    APPENDICES  (nuclear fuel & supporting references) 2/35 MOTIVATION – I, slide 1 of 3  THE NUCLEAR FUEL WEAKNESS - USNRC (draft) RG after 2014  Characterizing LOCA breakaway oxidation – RG 1.222 (17)  H2 that enters the clad promotes rapid embrittlement  (alloy composition  dependent).  Determining Post Quench Ductility (PQD) – RG 1.223 (18)  High PCT (1200 °C) and high local clad oxidation (17%) may not ensure PQD. High  Bu makes the situation worse.  Establishing (new) limits for Zr-Alloy clad material – RG 1.224 (19)  An alloy-specific cladding H2 uptake model is required. Accounting for  oxygen ingress on the ID is needed for Bu > 30.            Zircaloy-2 H2 ppm vs Bu Eq. Clad React. Vs H2 ppm  3/35 THE NUCLEAR FUEL WEAKNESS  Literature overview – Support information in Appendix 1  [FAILURES WITH PARAMETER RANGES WITHIN DBA BOUNDARIES]  • BALLOONING:  burst, pressure, temperature and time (recently measured) showing clad temperature values (at burst)  as low as 500 °C – 600 °C.      • OXIDE: oxide and oxide thickness (a function of Bu) ‘enlarge fuel failure region’, inducing spalling, hydride formation and embrittlement   • HOOP STRESS: individual nuclear fuel rod stress calculation now possible during the in-core fuel cycles: wide range results are predicted   • PCMI & PCI/SCC: complex failure mechanisms better understood, brittle rupture (frequently) possible.   • SNF: Clad weakness also affects releases from SNF and consequent radiological impact upon the environment (a function of Bu and LHGR)            MOTIVATION – I, slide 2 of 3  4/35 FURTHERMORE – Literature overview   ATR (Accident Tolerant Fuel), i.e. new material intended to prolong cladding life, the coated cladding supporting extended exposure to high temperature (1300 – 1400 °C) during LOCA, RIA and BDBA, the SiC (Silicon Carbide) cladding characterized by high melting point and minimal reaction with water (expected for commercial use in 2022), will not change the provided picture and the inevitable weakness of B1 (see below).   Fuel integrity following mechanical loads generated by the pressure wave dependent upon the Break Opening Time (BOT) in case of LBLOCA is not ensured: no experimental evidence available, current computational capability questionable.                         MOTIVATION – I, slide 3 of 3  5/35 6/35 THE NUCLEAR DISASTERS          TMI-2 - 1979 CHERNOBYL-4 - 1986 FUKUSHIMA-1-4 - 2011 (2) WE MUST WORK TO MINIMIZE THE POSSIBILITY OF OCCURRENCE OF THOSE EVENTS AND THEIR CONSEQUENCES MOTIVATION – II, slide 1 of 1     THE ELEMENTS FOR THE PROPOSAL (20 & 21)  ALARA (As-Low-As-Reasonably-Achievable) is an early principle, adopted for Radioprotection & disconnected from DSA.  IA (Independent Assessment) is a requirement, pursued only in principle: a wish rather than an achievement .   BEPU (Best Estimate Plus Uncertainty) is a key approach [origin of the term: nuclear thermal-hydraulics and AA during the ‘90s], not commonly accepted.  E-SM (Extended Safety Margin) is derived from SM, i.e. an established concept in nuclear reactor safety  ERT (Emergency  Rescue Team) is a virtual entity: it shall be mandatory after Fukushima.      BACKGROUND 7/35 THE VISION FOR NRS    8/35 Role for  BEPU Region for IA SM / E-SM Making ‘concrete’ the role of ALARA BACKGROUND BACKGROUND 9/35    THE VISION FOR SAFETY BARRIERS            RADIOACTIVE MATERIAL ELECTRONIC / COMPUTATION - ERT ASSISTED          CONCRETE & MECHANICAL                  FLUID-DYNAMIC / MECHANICAL                       MECHANICAL                            MECHANICAL [WEAK] X           NEW NO MORE SAFETY DRIVEN DESIGN DRIVEN B1 B2 B3 B4 B5 NOT A PHYSICAL BOUNDARY Current NPP OBJECTIVE  10/35    TO PROPOSE   THE BASES  FOR CONSTITUTING AN  ADDITIONAL SAFETY BARRIER   AGAINST THE RELEASE OF FISSION PRODUCTS    11/35 NRC – Interim Acceptance Criteria NRC – Appendix K to 10 CFR 50.46 NRC – RG 1.157 NRC – RG 1.203 Angra-2 BE LB-LOCA Atucha-2 BEPU Chapter 15 1970 2010  A HISTORIC OUTLINE THE BEPU FEATURES 2020 BEPU-FSAR 1990 SAFETY ANALYSIS / LICENSING – IAEA SSG-2, 2010 12/35 BEPU  THE BEPU FEATURES – Accident Analysis THE BEPU FEATURES – BEPU & ALARA WHAT IS BEPU?  The BEPU is a logical process which connects the understanding in NRS (and licensing) with nuclear TH.   The starting point for BEPU are the physical phenomena. This implies the DBA envelope.    BEPU implies the existence of qualified computational tools dealing with different disciplines, input decks or nodalizations and a method to evaluate the uncertainty.   BEPU needs the existence of qualified procedures for the application of the computational tools.   BEPU needs the existence of qualified code users and of maven capable of evaluating the acceptability of analysis.   BEPU needs the existence of ’legal’ acceptance criteria.   The application of BEPU implies the knowledge of the licensing process.   The structure of the FSAR must be adapted to BEPU including the design of the core, the experimental data drawn during the commissioning, the design of EOP, etc.   Any BEPU report should be a living document. 13/35 14/35  Computational tools / SYS TH codes – design and development  Computational tools / SYS TH codes – V & V procedures  Computational tools / SYS TH codes – procedures for application  Computational tools / nodalizations (or input decks) – development  Computational tools /nodalizations – V & V  procedures  Computational tools / code-coupling software – design and development  Uncertainty methods / design and development  Uncertainty methods / qualification procedures  NPP parameters database  Postulated Initiating Events (PIE)    Phenomena / physical aspects which characterize PIE  Databases for code and nodalization qualification  Scaling demonstration / procedures and database  Users of computational tools / qualification  DSA – PSA integration  Instrumentation and Control (I & C) modeling  Documentation requirements for each elements  Licensing framework – acceptance criteria, safety margins, procedures, etc.   V & V;  UNC;   COUPLING; DB;  SCALING   WHAT IS BEPU?-  CONSTITUTIVE ELEMENTS THE BEPU FEATURES 15/35 Code Verification ActivitiesFunctional Requirements of the CodeCode Verification Plan• Objectives• Approach• Schedule• Plan for testingNumerical solution• Verification matrix• Comparison with - Manufactured solutions- Analytical solution- Highly accurate numerical solution - ExperimentSource code (Software quality engineering )• Configuration management• Software quality analysis and testingNumerical algorithm• Verification matrix• Tests for   - Conservation- Flow transitions- Convergence• Robustness• Versatility VALIDATION:  STARTING FROM SETF & ITF CCVM + FFTBM FOR ACCURACY QUANTIFICATION RECENT CONCEPTS:   ASSESSMENT OF VALIDATION  (MINIMUM) INDEPENDENT ASSESSMENT FOR CODE USER V & V KEY ELEMENTS OF BEPU  DRAFT  IAEA-SRS-xxxx    Verification and Validation of Thermal-Hydraulic System Codes for  Nuclear Safety Analyses   DRAFT-IAEA V&V – 2017  VERIFICATION  ESTABLISHED QA PRACTICE 11/35 KEY ELEMENTS OF BEPU  16/35 HIERARCHY & KNOWLEDGE MANAGEMENT   SCALING ‘BRIDGES’ & ACHIEVEMENTS OECD/NEA/CSNI  S-SOAR – 2017  UNCERTAINTY KEY ELEMENTS OF BEPU  Multiple Inputn ~ 105BIC         CODE                INPUTDECK12nSelection of inputuncertain parametersn* < 102ID of range & PDFper each n* Multiple Outputm ~ 103(typical, uninfluent)12mPredictedNPP transientscenario UNCERTAINTYPROPAGATION1 - PROPAGATION OF CODE INPUT “UNCERTAINTIES”Multiple Inputn ~ 105BIC         CODE                INPUTDECK12nMultiple Outputm ~ 103(typical, uninfluent)12mPredictedNPP transientscenario UNCERTAINTYPROPAGATIONRelevant experimentaldata Accuracy quantification& criteria for accuracyextrapolation 2 - PROPAGATION OF CALCULATION OUTPUT “ERRORS”17/35 IAEA SRS-52 – 2008  The crack shape is assumed elliptical (two-dimensional crack) and the dimensions are taken from the ASME XI: ½ (102)¼ (102)< 1023½ (305)¼ (305)> 3052½ Wall thickness¼ Wall thickness102 – 3051Crack length (mm)Crack depth (mm)Wall thickness (mm)No. Crack positionRPV wall thickness = 290 mma = Crack Depth ~ 7 mmFracture Mechanics 1001502002503000 50 100 150 200 250 300 350 400Time [s][°C]7mmTemperature at crack tipThermal Shock0200040000 50 100 150 200 250 300Temperature [°C][N/mm^1.5]KIc  (RTNDT=50°C)7mmt = 50 st = 100 st = 150 st = 200 st = 250 st = 300 st = 400 sKEYRESULT  18/35 CODE COUPLING KEY ELEMENTS OF BEPU  EC-NEA CRISSUE-S – 2004  5/35 THE PROCESS  DATABASICSETFITFNPPSYS TH CODE DEVELOPMENTV & V -SCALINGUNCERTAINTYACCIDENT ANALYSIS  /  FSAR – CHAPT. 15We are here!LICENSING BEPU  Other Disciplines + PSA19/35 (TH) DATABASE / APPLICATION KEY ELEMENTS OF BEPU  NA-SA / UNIPI ARN (R GULATORY BODY) - APPROVED 2012 20/35 ENHANCING THE SM CONCEPT E-SM CONTRIBUTED BY  BEPU-FSAR  A1) Safety Principles, i.e. SP-1 to SP-10; A2) DID Levels, i.e. DL-1 to DL-5; A3) Safety Barriers, i.e. SB-1 to SB-6; A4) Safety Functions, i.e.SF-1 to SF-19; A5) PSA Elements, i.e. PE-1 to PE-n; A6) DSA Elements, i.e. DE-1 to DE-m.   NRST : THE KEY ELEMENTSTHE ‘A’ LIST – 6 topics –KEY ELEMENTSNRST – THE VIRTUAL SPACEKEY ELEMENTSTECHNOLOGICALSECTORSSSCSYSTEMS-STRUCTURES-COMPONENTSTHE LIFETIME……SMABCDA ‘FEW’ 10E4 E-SM DEFINITIONS  E-SM 21/35 ENHANCING THE SM CONCEPT               1 ACCEPTANCE CRITERIA / SAFETY LIMITS VALUES SAFETY BARRIER 1 TO 6 SAFETY FUNCTION 1 TO 19 Safety Margin No Close to the limit Acceptable Safe 2 1 ….  n … Average SM …. SF 1 to 19 Average SM SB 1 to 6 E-SM CONTRIBUTED BY  BEPU-FSAR  AROUND 10E4 E-SM   DETECTION & CONTROL  23/35 ISSUES WITH CURRENT IA ISSUES  NPP COMPLEXITY (efforts needed for IA ‘too large’ out of industry)   SAFETY DEPENDING UPON DETAILS  (details un-known out of industry; issue is proprietary information)   INDUSTRY ENGAGED IN CONTINUOUS CHANGES / IMPROVEMENTS (changes not necessarily qualified, e.g. passive systems)   IA ONLY POSSIBLE WITH LATEST BE TECHNIQUES  (expertise may not be available out of industry)   EXPERT ANALYSTS NOT NECESSARILY AWARE OF LICENSING DETAILS (the licensing framework is complex, too)  2ERT EMERGENCY RESCUE TEAM (20) 23/35 1. ERT ELEMENTS  • Helicopter  • DG + pump • Team (of around 5) 2. ERT OPERATION RANGE  About 500 km in a Country or geographical region targeting (typically) 10 NPP 3. ERT WORKING MODE  Plug-in pump delivery side to known  (3-5)  NPP  ERT-connectors 24/35 Integrating ALARA, BEPU, E-SM, IA, ERT  ALARA  NRS REQUIREMENTS BEPU [BEPU-FSAR] SM [E-SM]  INDEPENDENT ASSESSMENT LEGEND  Philosophy     Software  Hardware  Writing    NPP UNIT ERT ADDITIONAL SAFETY BARRIER     One may state that:   a principle (ALARA taken from fundamentals of the technology) +  a requirement  (IA, becoming actual) +  an approach (BEPU, becoming practical) +   a concept  (E-SM established in nuclear reactor safety, now expanded)  +  a virtual entity (ERT, becoming physical)  =  ADDITIONAL SAFETY BARRIER a ‘new’  SAFETY BARRIER 25/35 • NOT-RECOGNIZING  the weakness of the barrier constituted by clad,  • DELETING LBLOCA from the list of DBA (equals admitting no control of the technology),    HAVE A CONSEQUENCE:  LOW (TECHNICIANS AND) PUBLIC ACCEPTANCE! 26/35 FINAL REMARKS  -  slide 1 of 3 - FURTHER EVALUATION OF CURRENT STATUS -  1) ALARA at the origin of BEPU. 2) BEPU based on V&V, Scaling, Code Coupling, Uncertainty, and Database. BEPU extended to the entire FSAR (analytical parts). 3) E-SM (comprehensive and systematic set of) derivable with support from BEPU. 4) IA based on BEPU and making possible BEPU  5) ERT  a (very) simple product of current technology SUMMARY BEPU: must be pursued. Any further delay  is not justifiable for NRS. Safety Assessment (Licensing) must be independent of Vendor-Owner  BEPU-based I-FSAR & E-SM.  27/35 FINAL REMARKS  - slide 2 of 3 - FIVE ELEMENTS FOR THE NEW SAFETY BARRIER -  Reporting (again) the words of Australian-Chinese colleagues who analyzed the framework of the Fukushima event:  “… upgrading and strengthening a nuclear regulatory system is not optional but imperative to prevent the next core meltdown.”  1) … STRENGTHENING REGULATORY FRAMEWORK  2) RISK OF CORE MELT LOWERED  (to be demonstrated ) FOR A FACTOR 10 - 1000.  3) PROBABILITY OF CORE MELT  (target) TO THE LEVEL OF METEORITE FALL ON THE NPP.  4) COST OF NEW BARRIER  ≈ 1% CURRENT NPP. 28/35 FINAL REMARKS  -  slide 3 of 3 - THE NEW NPP  -  THE NUCLEAR FUEL WEAKNESS – Literature overview  BALLOONING:  pressure and temperature during experiments  w/o (15), (11) and w relocation (14) considered; azimuthal temperature asymmetry (6)                               Burst pressure, temperature, time APPENDIX  1 – (SUPPORT TO) MOTIVATION  1 OF 5 29/35 APPENDIX  1 – (SUPPORT TO) MOTIVATION 2 OF 5 THE NUCLEAR FUEL WEAKNESS – Literature overview  OXIDE THICKNESS (11) vs Bu and failures (3). Hydride rim depth vs oxide thickness and burst failures (7) and sketch of spalling and hydriding (8)-(9)                       30/35 THE NUCLEAR FUEL WEAKNESS – Literature overview  • HOOP STRESS: detailed NPP (Watts, US) calculations  are possible showing widely changing conditions, including Bu effect (10)                       APPENDIX  1 – (SUPPORT TO) MOTIVATION 3 OF 5 Cycle 1 axial maximum  hoop stress (MPa) Cycle 3 axial maximum  hoop stress (MPa) 31/35 32/35 THE NUCLEAR FUEL WEAKNESS – Literature overview  PCMI failures – RIA (7) and PCI/SCC failures, IR project for BWR, (5).                     APPENDIX  1 – (SUPPORT TO) MOTIVATION 4 OF 5 The model assumes that SCC failures begin as an inter-granular fracture due to cesium–iodine chemical attack (in presence of oxygen potential) and independent of applied stress, … that leads to cracks propagation. Clads fail with small diameter changes at relatively low values of LHGR (final ramp). Early (2010) RIA code results showing Bu effect upon clad failures (see also effect of oxide thickness-  this appendix, slide 2 of 5). THE NUCLEAR FUEL WEAKNESS – Literature overview  SNF: Instant Release Fraction at high Bu (40 -60) sharply increases when LHGR > 20 kw/m  (1)            APPENDIX  1 – (SUPPORT TO) MOTIVATION 5 OF 5 Part 1 - When the SNF is disposed of in an underground repository, the radionuclides may gradually be released after failure of the canister and subsequent water ingress. The release rate of radionuclides differs depending on their chemical properties, their chemical speciation in the fuel, as well as the location where they are segregated within the SNF. Part 2 - The release of soluble segregated elements from the accessible gap, cracks and grain boundaries is fast. Most of their inventory is released within a few months or even days. The quantity of these rapidly released inventories normalized to the total nuclide inventories is commonly called the Instant Release Fraction (IRF) … they can significantly contribute to or even dominate the calculated dose exposure. 33/35  1. J. NM 2017, vol 484 - Instant release of fission products in leaching experiments with high burnup nuclear fuels in the framework of the Euratom project FIRST, K. Lemmens, *, E. Gonzalez-Robles, B. Kienzler, E. Curti , D. Serrano-Purroy, R. Sureda , A. Martínez-Torrents, O. Roth, E. Slonszki, T. Mennecart , I. Günther-Leopold,Z. Hozer 2. J. RSER 2013, vol 17 - Accident like the Fukushima unlikely in a country with effective nuclear regulation: Literature review and proposed guidelines, Q. Wang , Xi Chen, Xu Yi-chong 3. J. NM 2011, vol 418 - The effect of fuel rod oxidation on PCMI-induced fuel failure, K-T. Kim 4. J. EFA 2011, vol 17 - Investigation of failure behavior of two different types of Zircaloy clad tubes used as nuclear reactor fuel pins, M.K. Samal, G. Sanyal, J.K. Chakravartty 5. J. ANE 2012, vol 50 - Modeling of BWR Inter-Ramp Project experiments by means of TRANSURANUS code, D. Rozzia, A. Del Nevo, M. Adorni, F. D’Auria 6. J. NED 2014, vol 268 - Dynamic ballooning analysis of a generic PWR fuel assembly using the multi-rod coupled MATARE code, L. Ammirabile, S. P. Walker 7. J. NED 2010, vol 240 - A consistent approach to assess safety criteria for reactivity initiated accidents, C. Sartoris, A. Taisne, M. Petit, F. Barré, O. Marchand 8. IAEA 2015, Conf. Fuel - IAEA–CN–226 ID122, - On the impact of the fuel assembly design evolution in the spent fuel management , J. M. García de la Infanta 9. J. NST 2006, vol 43 - Influence of Outer Zirconia Transient Cracking and Spalling on Thermomechanical Behavior of High Burnup Fuel Rod Submitted to RIA, V. Georgenthum, J. Desquines,V. Bessiron 10. J. NED 2018, vol 327 - Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3 ,S. Stimpson, J. Powers, K. Clarno, R. Pawlowski, R. Gardner, S. Novascone, K. Gamble, R. Williamson         APPENDIX  2 – REFERENCES  1 of 2 34/35  11. J. NED 2011, vol 241 - Oxide thickness-dependent transient cladding hoop stress, K-T. Kim, D. W. Jerng 12. J. CALPHAD, 2016, vol 55 - Application of thermochemical modeling to assessment/evaluation of nuclear fuel behavior, T. M. Besmann, J.W. McMurray, S. Simunovic 13. J. NM 2018, vol 500 - Phenomenology of BWR fuel assembly degradation, M. Kurata, M. Barrachin, T. Haste, M. Steinbrueck 14. J. NED 2017, vol 312 - Effects of fuel relocation on reflood in a partially-blocked rod bundle, B. J. Kim, J. Kim, K. Kim, S. W. Bae, S-K. Moon 15. J. NM 2017 - Study of clad ballooning and rupture behavior of Indian PHWR fuel pins under transient heating condition in steam environment,  T. K. Sawarn, S. Banerjee, S. S. Sheelvantra, J.L. Singh, V. Bhasin 16. J. NED 2014, vol 280 - Study of clad ballooning and rupture behavior of fuel pins of Indian PHWR under simulated LOCA condition, T. K. Sawarn, S. Banerjee, K.M. Pandit, S. Anantharaman 17. USNRC, 2018, RG 1.222, draft 18. USNRC, 2018, RG 1.223, draft 19. USNRC, 2018, RG 1.224, draft 20. J. STNI 2012 -. The Fukushima event: the outline and the technological background, F. D’Auria, G. Galassi, P. Pla,  M. Adorni, 2012. 21. J. NED 2017, vol 324 - Strengthening nuclear reactor safety and analysis, F. D'Auria, N. Debrecin, H. Glaeser   … other called in the text            APPENDIX  2 – REFERENCES 2 of 2 35/35 

AN ADDITIONAL SAFETY BARRIER FOR EXISTING AND NEW NPP

https://arpi.unipi.it/bitstream/11568/941345/1/6-NISLA-2018.pdf

AN ADDITIONAL SAFETY BARRIER FOR EXISTING AND NEW NPP

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Archivio della Ricerca - Università di Pisa