Studies on the Recriticality Potential during 1F3 Reflooding

Transcription

1 8 TH CONFERENCE ON SEVERE ACCIDENT RESEARCH ERMSAR 2017 Studies on the Recriticality Potential during 1F3 Reflooding Piotr Darnowski, Kacper Potapczyk, Konrad Świrski Institute of Heat Engineering, Warsaw University of Technology May16-18, 2017 Warsaw, Poland

2 Agenda 1. Introduction 2. Methodology 3. Results 4. Discussion 5. Conclusions 6. References 2

3 Introduction - problem statement Recriticality in BWRs B4C and steel subject to eutectic reaction with Liquidous Temp. ~1250 C Fuel has much higher degradation temperature Absorber may relocate before the fuel. Short time window. In the case of unborated water injection recriticality is possible: Ref: [1]-[8] Core spraying ECCS Feedwater restart Water slug due to steam explosion 3

4 Introduction - motivation On January 2015 Finnish VTT published MELCOR2 Fukushima Unit 3 input deck [1]. We started to study the MELCOR model and we observed the following: During the core the fuel was partially intact. Majority of control blades mass was relocated to the lower plenum. Core was reflooded with seawater [2]. Absorber mass Fuel mass 4

5 Methodology Main purpose: Assessment of the risk of recriticality SERPENT 1 & 2 - Monte Carlo neutron transport solver. MELCOR severe accident integral computer code. Data exchange framework between codes was developed to transfer MELCOR results to SERPENT models. EDF - External Data Files 5

6 Methodology - MELCOR VTT MELCOR [1] inputdeck was applied to obtain core state during the accident: water levels, fuel, temperatures, integrity. CVH core nodalization was modified. Calculations with MELCOR Absorber mass Fuel mass original VTT model (4 CVs) Control rod poison (CRP) and fuel masses distribution at the reflooding initiation time. COR package model Intact active core axial levels 6-15, radial rings 1-4. modified core (CVH) model (16 CVs). 6

7 Methodology - SERPENT 3D Full core simulations were beyond our abilities - simplified model was applied. Neutronic modelling approach was based on the research by Mostellar and Rahn [7] and research by Bassam and Witt [8]. 3D core unit cell generated using 2D burnup results. 3D models were composed of four assemblies with different burnup: 15, 22.5 and 30 GWd/tHM. Average burnup 22.5 GWd/tHM (close to 1F3 burnup). Periodic BC on horizontal surfaces and black on the top and bottom. Reference cases HFP, Cold Unrodded and Cold Rodded to assess results. Vertical cross-sections Horizontal cross-section - intact rodded FA 7

8 Methodology - SERPENT 2D SERPENT was as applied as a 2D burnup code to estimate the core inventory. Calculations for different void fractions at Hot Full Power conditions (density history). The STEP3 fuel assembly design described in OECD/NEA Criticality Benchmark Phase IIIC was applied [10]. For other data: 1F3 and PB-2 available data [11],[12],[13],[14] Decay calculations - 48h delay. Changes in reactivity due to the Iodine Pit. 2D data transferred to the 3D models. Axial burnup distribution was modelled in 3D models. 8

9 Results - reference cases Reference cases results - frame of reference for accident results. Hot Full Power, Cold Rodded and Cold Unrodded STATE Comment SERPENT 1F3 State, Temperatures, Void eigenvalue SD HFP 3D, variable T and density CR ColdRodded, 300K, 0% CUR ColdUnRodded, 300K, 0% Approach was based on the research by Mostellar and Rahn [7]. Sub-critical state for k-eff < (khfp-2sd) Super-critical state for k-eff > (khfp+2sd) Substantially sub-critical for k-eff < (kcr-2sd) Substantially super-critical for k-eff > (kcur+2sd). 9

10 Results - base case Absorber mass Fuel mass MELCOR modified model Conservative case. No boric acid No control blades Recriticality - predicted But it is not realistic case 10

11 Results - boric acid sensitivity Different concntrations: 0, 100, 250, 500, 750, 1000, 1500, 2000 ppm of nat. boron. 750 ppm definitly prevents recriticality (substantially sub-critical). 250 ppm is enought for sub-critical state. Ring 3 Ring 4 11

12 Results - control blades sensitivity Ring 3 Ring 4 No boric acid. Control blades with fixed length. Different CBSF - control blade survival fractions. CBSF = Length/Intact Length. ~2 nodes without control blades (~80 cm) may lead to recriticality. 12

13 Discussion - main issues Highly unlikely that 100% of boron will leave the core. Small fraction of boron is enough to maintain sub-criticality. Leakage in the degraded core should be very high. Boric acid accumulation in the RPV - it was transferred during earlier reflooding. In LWRs geometry is optimized and any deviations reduce reactivity Rodless core time window Max flowrate was ~11 kg/s - it is low for recriticality. It is not full core 3D model. Only axial leakage, no radial. Only one type of fuel. In 1F3 there was MOX. Simplified degradation - based on the MELCOR - no absorbers nor structures left in the core. No re-solidification and other geometry variations. Intact fuel in the core assumed to maintain its original geometry and composition. Seawater was injected it may act as a neutron poison. No kinetic/dynamics simulation - SERPENT - only k-eff for the given core state. US EPRI performed similar study with the MAAP5 and the SERPENT for the 1F2 case [15]. They used SERPENT full core 3D model and detailed core inventory calculations with PARCS/HELIOS. no significant differences between 1F2 and 1F3, at least our knowledge about the core status and accident progression doesn t allow distinguishing with respect to re-criticality 13

14 Discussion Comparison of the core water level measured by TEPCO [14] with original MELCOR model and modified model. Uncertain measurements due to severe accident conditions. Rate of water injection with fire engines (Ref [1]). It is uncertain how much water was transferred into the core. Max flowrate was ~11 kg/s. 14

15 Conclusions Recriticality was very unlikely. There are no evidence of the recriticality. Recrticiality posible only when: Water is unborated. No boric acid accumulation in the RPV and no boron (from control blades) accumulation. core is unrodded at least partially. fuel is at least partially intact. Enought water transferred with proper flowrate. Achivments: MELCOR applied with SERPENT Parametric studies of boron and control blades. Older methodology dedicated to assess recriticality was modified. It takes into account the 3D nature of the system. Fast running models - sensitivity and unceratnity studies. 15

16 References [1] Sevón, T., 2015a. A MELCOR model of Fukushima Daiichi Unit 3 Accident, Nuclear Engineering and Design, 284, [2] TEPCO, Timeline of major events at Fukushima Daiichi Nuclear Power Station Unit 3from the occurrence of the earthquake until March 13, page 110, [3] W. Frid, F. Hojerup. Severe accident recriticality analyses. Nuclear Engineering and Design 209 (2001) [4] W. Frid, F. Hojerup, I. Lindholm, J. Miettinen, L. Nilsson, E. K. Puska, H. Sjovall, 1999, Severe Accident Recriticality Analyses, SKI Report 99:32, SKI - Swedish Nuclear Power Inspectorate [5] L. Nilsson; D. Kropaczek; W. Frid, 2000, Studies of Recriticality Transients in BWRs during Reflooding: Simulate-3k Development and Analyses, PHYSOR-2000, American Nuclear Society. Advances In Reactor Physics, and Mathematics and Computation Into the Next Millennium. Pittsburgh, PA, May 7-11, 2000 [6] Hojerup, F., Miettinen, J., Nilsson, L., Puska, E., K., Sjoval, H., Anttila, M., Lindholm, I., 1997, On Recriticality during RefReflood of a Degraded Boiling Water Reactor Core, Report NKS/RAK2(97)TR-A3 [7] Mosteller, R. D., Rahn, F. J., "Monte Carlo Calculations For Recriticality During The Reflood Phase Of A Severe Accident In A Boiling Water Reactor". Nuclear Technology , [8] Bassam, I., Witt, R., Parametric Study of Recriticality in a Boiling Water Reactor Severe Accident, Nuclear Technology 107 [9] SNL, MELCOR Best Modelling Practices - State-of-the-Art Reactor Consequence Analyses Project, Sandia National Laboratories, NUREG/CR [10] OECD/NEA, 2016, Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel - Burn-up Credit Criticality Safety Benchmark Phase III-C Final Report NEA/NSC/R/(2015)6, [11] Moore, R. S., Notz, K. J., 1989, Oak Ridge National Laboratory Report ORNL/TM-10902, Physical Characteristics of GE BWR Fuel Assemblies [12] Larsen, N. H., 1978, EPRI Topical Report by General Electric, Core Design and Operating Data for Cycles 1 and 2 of Peach Bottom 2. [13] TEPCO, 2013a. Plant Specifications of Unit 3, _analysis/ps-unit3-01.pdf ( ) [14] TEPCO, 2013b. Measured Data, Unit 3, Water Level in RPV, 30 November analysis/md-unit3-02.xls [15] EPRI, 2016., Technical Evaluation of Fukushima Accidents: Phase 2 Potential for Recriticality during Degraded Core Reflood. EPRI- Technical Report , Final Report April 2016 [16] Huffer, J., 2004, BWR Axial Profile, Oak Ridge National Laboratory Report, Engineered Systems Project, CAL-DSU-NU REV 00A, %20REV%200A.pdf [17] J. Leppänen. Serpent a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code. VTT Technical Research Centre of Finland. (June 18, 2015) [18] SNL, 2011a. MELCOR Computer Code Manual, Vol. 1: Primer and Users Guide, Version 2.1., Sandia National Laboratories, NUREG/CR-6119, Vol. 1, Rev [19] SNL, 2011b. MELCOR Computer Code Manual, Vol. 2: Reference Manual, Version 2.1., Sandia National Laboratories, NUREG/CR-6119, Vol. 2, Rev [20] Scott, W. B., D. G. Harrison, R. A. Libby, R. D. Tokarz, R. D. Wooton, R. S. Denning, R. W. Tayloe, Jr., 1990, Recriticality in a BWR Following a Core. [21] P. Darnowski, K. Potapczyk, and S. Konrad, Investigation of the recriticality potential during reflooding phase of Fukushima Daiichi Unit-3 accident, Ann. Nucl. Energy, vol. 99, pp , [22] ORNL, Physical Characteristics of GE BWR Fuel Assemblies, ORNL/TM-10902, [23] US EPRI, Core Design and Operating Data for Cycle 1 and 2 of Peach Bottom 2, [24] T. Yokoyama, F. Taiki, and N. Hisashi, Study on Particle and Absorber Effects on Multiplication Factors of Debris Beds with MVP, in Data and Analysis in Nuclear Criticality Safety-I, ANS Winter Meeting 2011, [25] G. Alonso, et. Al.. Annals of Nuclear Energy, Impact of the moderation ratio over the performance of different BWR fuel assemblies [26] Huffer, J., BWR Axial Profile Oak Ridge National Laboratory Report. Engineered Systems Project, CAL-DSU-NU REV 00A %20REV%200A.pdf. 16

17 Thank you for your attention. Thank you for your attention. The research was financed by Faculty of Power and Aeronautical Engineering Dean Grant number PR 504/02052 in

18 Discussion - other issues Neutron leakage in the real 3D core ~ pcms Core shuffling patterns, different fuel, MOX fuel impact Less than 4 bundles per one blade - lower reactivity Simplified upper and lower structures. Not all FPs and higher actinides were taken into account. Noble FPs leave the core during early in-vessel it was not taken into account - Xe-135 up to pcms; Samarium up to -500 pcms and it increased in comparison to the HFP (~17%). Impurities in the cladding and structure ~-100 pcm No kinetic/dynamics simulation - SERPENT - only k-eff for the given core state. No evaporation of water due to the recriticality! 18

19 SNL MELCOR results - MELCOR 2.1 Jeffrey Cardoni Randall Gauntt, Donald Kalinich Jesse Phillips, MELCOR SIMULATIONS OF THE SEVERE ACCIDENT AT FUKUSHIMA DAIICHI UNIT 3, NUCLEAR TECHNOLOGY VOL. 186 MAY

20 Fernandez-Moguel, L. Birchley, J., Analysis of the accident in the Fukushima Daiichi nuclear power station Unit 3 with MELCOR_2.1, Annals of Nuclear Energy 83 (2015)

21 VTT results Sevón, T., 2015a. A MELCOR model of Fukushima Daiichi Unit 3 Accident, Nuclear Engineering and Design, 284,

22 SNL Report 2012 Fukushima Daiichi Accident Study, SAND , August

23 SNL Report ORNL Results - melcor Fukushima Daiichi Accident Study, SAND , August

24 EPRI - MAAP5 Report EPRI, Fukushima Technical Evaluation Phase 1 MAAP5 Analysis

25 EPRI - MAAP5 Report 25

26 EPRI - MAAP5- MELCOR Crosswalk report - 1F1 Modular Accident Analysis Program (MAAP) MELCOR Crosswalk Phase 1 Study

27 Additional slides Ref: [25] Axial burnup profiles for three assemblies used in the core unit cell model. 27

28 Additional slides 28

29 Additional slides 29

30 Additional slides Axial variation of the temperatures (left) and moderator density (right) for the reference Hot-Full-Power 3D calculations. 30

31 Modified model fuel temperaturę before the reflooding COR-TFU Ring 3 48h COR-TFU Ring 4 31

32 Additiona slides - modified core 53h state 32

33 Mod. Model fuel temperatures after reflooding 33

34 Additiona slides - orginal core 53h state 34

35 Orginal model temperatures after reflooding 35

36 Additional slides EPRI - full core results 36

37 Additional slides EPRI - full core results 37

38 Additional slides Isotopes traced in the 2D assembly burnup calculations. Bold underlined isotopes are used in the 3D criticality calculations 38

39 Additional slides Coolant axial density profile used in the burn-up full power calculations. 39

40 Additional slides OECD/NEA Phase IIIC Benchmark Results for STEP-3 assembly with 40% void and our calculations with SERPENT. 40

41 Additional slides Comparison of the average Benchmark results with void fractions 40%,70%,0% with SERPENT burnup results used in the burnup calculations. 40% void fraction case corresponds to the water density g/cc and Level 3 density is equal to g/cc. 41