ML20247D082

Summary of 890315 Meeting W/Epri & Contractors in Palo Alto,Ca Re Advanced LWR Issues.List of Attendees & Handouts Encl

ML20247D082
Person / Time
Issue date:	03/21/1989
From:	Long W Office of Nuclear Reactor Regulation
To:	Chris Miller Office of Nuclear Reactor Regulation
References
PROJECT-669A NUDOCS 8903300376
Download: ML20247D082 (100)

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{{#Wiki_filter:. .An .n March 21 1989 0 Project No. 669 HEMORANDUM FOR: Charles L. Miller, Project Director Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects - III, IV V and Special Projects FROM: William O. Long, Senior Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV V and Special Projects

SUBJECT:

MEETING REPORT, NRC-EPRI MEETING AT PALO ALTO, CA. MARCH 15, 1989, TO DISCUSS ADVANCED LIGHT WATER REACTOR ISSUES On March-15, 1989, members of NRR staff met with the Electric Power Research Institute and its contractors to discuss issues related to the staff's review-I of the Advanced Light Water Reactor Requirements Document. The meeting was l held at 3412 Hillview Ave. Palo Alto, CA. in the EPRI conference ~ room. The meeting consisted of presentations prepared by EPRI and its contractors.- The attachments include an attendance list and copies of all handouts. l /s/ William O. Long, Senior Project Manager Standardization and Non-Power Reactor Project Directorate-Division of Reactor Projects - III, IV, Y and Special Projects l Attachments: l As stated DISTRIBUTION: LCentral Filex GHolahan a NRC PDR * " " a MCunningham JSniezek BSheron PDSNP R/F BMorris WLong WTPratt(BNL) l. 0GC TKing l EJordan AThadani BGrimes NRC Participants ACRS(10) RWHouston RWoods LRubenstein y (EPRI TRIP REPORT) {'e [ g PM:PDSNP D:PDSNP M C h<,0 h WLong i CMiller 03/3,/89 03/bl/89 Fd 8903300376 890321 6 A PDC

f V 'o UNITED STATES ! 4.. ' ' 8' -o NUCLEAR REGULATORY COMMISSION l c E. WASHINGTON, D. C. 20555 March 21, 1989 Project No. 669 MEMORANDUM FOR: Charles L. Miller, Project Director-I Standardization and Non-Power 1 Reactor Project Directorate Division of Reactor Projects - III, IV 1 s V and Special Projects i FROM: William 0. Long, Senior Project Manager Standardization and Non-Power { Reactor Project Directorate Division of Reactor Projects III, IV .V and Special Projects

SUBJECT:

MEETING REPORT, NRC-EPRI MEETING AT PALO ALTO, CA. MARCH 15, 1989, TO DISCUSS ADVANCED LIGHT WATER REACTOR ISSUES { l On March 15, 1989, members of NRR staff met with the Electric Power Research i Institute and its contractors to discuss issues related to the staff's review 1 l of the Advanced Light Water Reactor Requirements Document. The' meeting was ] l held at 3412 Hillview Ave. Palo Alto, CA. in the EPRI' conference room. 4 The meeting consisted of presentations prepared by'EPRI and its contractors. l The attachments include an' attendance list and copies of'all handouts. 1 / W William O. Long, Senior Project Mana' er g Standardization and Non-Power j Reactor Project. Directorate Division of Reactor Projects - III, IV, Y and Special Projects Attachments: As stated l e i

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(.s Meeting with NRC to Discuss ALWR Severe Accident Approach March 15,1989 Agenca ) 8:30 a.m. Introductory Remarks NRRIEPRI 8:45 a..m. ALWR Program Overview EPRI 9:00 a.m. ALWR Program Policy Positions ) Design Bases Licensing Design Basis + + Risk Evaluation Basis + Performance Evaluation Basis Severe Accident t ALWR Top Levelinvestment Protection + and Public Safety Criteria - CDF< 1X10( 5)/yr ) -- Frequency < 1X10( 6)/ year for site boundary dose greater than 25 rem whole booy Prevention of Core Damage - margins + and features Plant features provided for other j + purposes but with value in SA mitigation (cavity configuration, RCS depressur!zation capability, ability to flood cavity) Discussion NRR/EPRI 10:30 a.m. Source Term ALWR Soarce T arm Treatment EPRI DiscusEon NRR/EPRI 11:30 a.m. PRA for At.WR ALWR Function-Level PRA Evaluation EPRI Key Assumptions and Groundrules EPRI (KAG) Document Discussion NRR/EPRI i l 12:15 p.m. Working Lunch 12:30 p.m. Treatment of Key Severe Accident issues ALWR Technical Basis for EPRI Hydrogen Control and Generation + Direct Containment Heating + Core Bebris Coolability + Discussion (following each subtopic) NRR/EPRI 4:00 p.m. EPRI Response to 11/22/88 NRC Letter EPRI on Proposed Enhancements 5:00 p.m. Adjourn

4 ALWR PROGRAM OVERVIEW J ] l EPRI ALWR Procram ALWR PROGRAM GOALS Establish utility leadership and effect positive progress toward a revitalized nuclear power opton in the United States Formulate a practical and credible foundation for the design of advanced light water reactors for the next decade EPRI ALWR Procram ,e,,, l

'a e Y ALWR PROGRAM OBJECTIVES In support of tMse goals, the ALWR Program objectives are: . A stabilized regulatory basis, via cooperative effort with NRC to identify and resolve outstanding issues of nuclear plant safety . Development of a set of specitic desi n and performance l requirements for the advanced LWR he " Requirements J Document") l l = An assessment of nuclear plant design concepts which would I incorporate greatly simplified, passive safety systems l EPRI ALWR Procram / l l l l ALWR FUNDAMENTAL ACCEPTANCE CRITERIA = Technicalexcellence The ALWR must be a good power plant - safe, user friendly, efficient, compatible with the environment Economic advantage The ALWR must be economically competitive with other power generation options, considering life cycle cost and first cost . Investment protection The ALWR must provide very high protection of the utility investment, particularly in terms of: very low susceptibility to major accident - assured licensibility predictable 'and controllable construction schedule predictable and controllable plant availability 1 m__________ _ _. _ _ _ _ J

w PROGRAM APPROACH . Ensure utility focus, leadership . Examine experience, build on success . Involve NSSS vendors and A-Es, apply their talents; incorporate their best products and ideas . Work with NRC in a constructive, nonconfrontational environment . Establish a sensible starting point for standardization EPRI ALWR Procram ( 3 ALWR DESIGN PRINCIPLES l ALWR Utility Steering Committee has established the following as key technical pnnciples to guide the design of any ALWR Highest attention to nuclear safety Simplicity - to enhance safety, constructibility, operations and maintenance Margin - a rugged, forgiving plant l Proven techriology reliance on demonstrated success paths Human f actors attention to man-machine interf ace in every aspect of the design EPRI ALWR Procram 4

~- -r 1 l l [ Major Elements and Funding U. S. Utilities U. 5. Taxpayers U. S. ALWR (via Dot) ) (via EPRI) g . cerurication EP.fti -ocaewn e aeguistory -c-e system eo-320N Staoistration e Passive Pl.nt W l e uttlitypec ts - System / component

• 0versti verification /~est
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l EPRI ALWR Procram 1 INTERNATIONAL PARTICIPATION IN THE ALWR PROGRAM Agreements are in place with: Taiwan Power Co. Korea Electric Power Corp. - Kansal Electric Power Co. - KEMA, The Netherlands - ENEL, Italy - EdF, France These agreements provide for: Active participation in ALWR Program Shared technicalinformation Direct funding to the program EPRI ALWR Procram

4 l EVOLUTIONARY AND PASSIVE CONCEPTS In the ALWR Program, two concepts are being developed: Evolutionary ALWR A simple, rugged and reliable advancement of today's LWR designs, using conventional safety system concepts l Passive Plant - A greatly simplified ALWR which employs primarily passive I means for accident prevention and mitigation EPRI ALWR Proaram l i 1 THE EVOLUTIONARY ALWR,IN CONCEPT ) PWR or BWR, higher rated plant (1100-1300 MWe) - Substantialimprovements in safety, simplification, margin, - Extensive application of lessons learned in existing plants Utilizes conventional safety systems mncepts - Employs advanced mntrol systems A direct descendent of today's LWRs EPRI ALWR Procram l 1

l ~ EVOLUTIONARY ALWR - BENEFITS l This concept is attractive because: - The Evolutionary ALWR has very high success potential;its j design is a direct application of 30 years' experience, clearly [ proven technology l - The Evolutionary ALWR can be available in the very near term. It is closely linked to U.S. vendor products (ABWR, APWR, Systems 80+) now being developed and marketed internationally, and being submitted to NRC for certification I EPRI ALWR Procram PASSIVE PLANT-CONCEPT in concept, the ALWR Passive Plant is a design which: . Utilizes primarily passive means (gravity, natural circulation, stored energy) for accident prevention and mitigation . Keeps core protected without operator action for about 3 days !s greatly simplified compared to existing plants . Can be PWR or BWR, of reference size 600 MWe . Can be constructed in three years, with extensive modularization, prefabrication EPRI ALWR Pmaram

y e. .w ~ 1 1 = 1 1 PASSIVE PLANT-BENEFITS The ALWR Passive Plant is an attractive concept because: . It provides a basis for renewed public, government and investor confidence it offers fundamental advances in safety, simplicity, constructibility . Its lower rating: - can better match needs of U.S. utilities with low or uncertain demand growth permits smaller capital investment to first power generated . Smaller, simpler plants have had historically high capacity factors EPRI ALWR Procram I THE UTILITY REQUIREMENTS DOCUMENT l The Requirements Document is the primary work product in this phase of the ALWR Program . It establishes top tier, functional and system / component design requirements for - evolutionary and passive plants PWR and BWR, entire plant it incorporates resolutions of generic safety issues and optimization issues . It reflects industry and NRC consensus the principal safety, performance and design requirements for the ALWR EPRI ALWR Procram s 4 _..____________E__..___

~ l 3 ALWR REQUIREMENTS DOCUMENT STRUCTURE VOLUMEl-ALWR TOP. TIER l REQUIREMENTS

• EXECUTfvE SUMM ARY e ALWR POLICIES
• ALWR KEY REQUIREMENTS I

l VOLUMElil PASSIVE PLANT j VOLUME 11. EVOLUTIONARY PLANT ALWR REQUIRE.kENTS PTER 1: OVERALL PERFORM ANCE i l AND DESON REQUIREMENTS CHAPTER 1: OVERALL PERFORMANCE FOR PASSIVE ALWR AND DESON REQUIREMENTS PLANTS FOR EVOLUTIONARY ALWR PTER 213: REQUIREMENTS FOR PLANTS SYSTE*.1$ AND STRUCTURES CHAPTER 2-13: REQUIREMENTS FOR SYSTEMS AND STRUCTURES e -,, l l ALWR Utility Requirements Document Chapter 5 1 ENGINEERED SAFETY SYSTEMS l l L 1 EPRI ALWR Procram mc m. l l j ^

v l + t KEY REVIEW AREAS FOR CHAPTER 5 NRC confirmation of the adequacy of certain key requirements is necessary for industry confidence to proceed on ALWR designs. EPRI has requested that the NRC Draft SER for Chapter 5 include specific findings on these key areas.

• Defined Optimization Subjects l

- Type C containment leak rate testiinterval(Appendix C.1 and Section 6.3.2) - Source term (Appendix C.2 and Sections 12.3 - 1.2.3.4) Hydrogen control (Appendix C.3 and Sections 2.4.1.7 and 6.5) For ALWRs designed to meet the Chapter 5 requirements, resolution of those Generic Safety Issues contained in Appendix B to Chapter 5 (and provided in the section of the chapter that reference Appendix B) EPRI ALWR Procram KEY REVIEW AREAS FOR CHAPTER 5 Acceptability, for purposes of protection of the public health and safety, of: A COF (mean) s1 x 10-5/ reactor year '1 ' Y 10-6/2fc '9m criterion ANS 5.1 decay heat for calculations other than 10CFR50 Appendix K (Section 2.2.6) 2 divisions for safety systems (Section 2.3.1) + 8 hour SBO survival time on best estimate basis (provided that CDF target is met)(Section 2.3.3) Emergency power source start and load sequencing time, of 40 seconds (objective: 60 seconds) to improve reliability (Section 3.4.5) EPRI ALWR Procram me n ,s

.e KEY REVIEW AREAS FOR CHAPTER 5 Acceptability of: -75% active clad reacted less than 13% by volume (average) as deterministic containment evaluation basis for combustible gas control (Section 2.4.1,7 and 6.5) Severe Accident approach outlined in Sections 1.2.4, 2.4.2.3 through 2.4.2.8, and 6.6, including the following specific SA requirements: - containment margin (Section 6.6.2) - prevention of direct containment heating (Section 6.6.3.1 l and 6.6.4.3) - assurance of core debris coolability of spreading and l covering with water (Section 6.6.3.2 through 6.6.3.4) containment filtered vent not reautred EPRI ALWR Procram CHAPTER 5 POLICY STATEMENTS . ALWR Design Bases: - licensing design basis - risk evaluation basis - performance evaluation basis . Severe Accident Protect'on . ALWR Treatment of Source Term issues

Figure 5.1-2 ALWR Defense in Depth Accident Resistant Dogjgns ALWR Design Features Which Provide intrinsic Safety Design Margins Simplicity l Best Materials l Extended Operator Response Times Chapters 3 & 4 - RCS and Reactor Systems E Core Damaae Prevention Systems Which Prevent Initiating Events from Progressing to the Point of Core Damage l Mitiaation Systems to Contain Fission Products Released as a Result of Core Damage Accidents Chapter 5 - Engineered Safety Systems Page 5.1-3

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l' g SEVERE ACCIDENT PROTECTION 1 ' Three Kev E!ements:

1. Design to meet applicable Regulatory Requirements

- covers licensing design basis provides significant margin

2. Emphasize core damage prevention targetr core damage frequency, s 10 E 5/ reactor year l

- provide additional preventive features to protect utility investment

3. Analyze dominant SA sequences to show sufficient margin:

- conservative design (e.g., containment design pressure) - investment protection features (e.g., increased RCS inventory) - plant features provided for other purposes, but with value in SA mitigation (e.g., RCS vent values) - realistic, mechanistic evaluations including non-safety related l j equipment EPRI ALWR Procram 1 amC Me as l I i ALWR CORE DAMAGE FREQUENCY CRITERION in addition to meeting all other licensing design basis requirements, mean annual CDF s 1 x 10 E -5 Believed to be a factor of 5 to 10 better than most current plants Sufficiently low for protection of utility investment a Function level PRA models are being developed for ALWRs Work to date indicates that the ALWR requirements specified to date will result in a plant that is likely to meet the 1 x 10 E 5 target value EPRI ALWR Procram .me ni m I

~ ALWR PUBLIC SAFETY CRITERION . Additionalcheck bevond meeting other licensing design basis requirements . Site boundary (0.5 mi.) whole body dose shall be less than 25 Rem for accident sequences whose cumulative frequency (mean value) exceeds 1 x 10 E -6 per reactor year . This criterion is viewed as an extremely demanding target worth reaching for EPRI ALWR Procram ( 3 ALWR PUBLIC SAFETY CRITERION The criterion was selected based on a number of considerations a desire on the part of utility sponsors to define an ALWR that is excellent in all respects - an accident frequency less than 1 x 10 E 6 per reactor year is low enough to satisfy this desire for excellence and the public perception - 25 Rom at the site boundary is a very low dose with *no observable health eff acts" (10CFR100) A design that meets this criterion should meet the health effect goals in the NRC Safety Goal Policy with margin . Preliminary estimates based on exisitng plant PRAs indicate that an improved ALWR whose dominant accident sequence frequencies are reduced has good likelihood of meeting this stringent criterion EPRI ALWR Procram

- i i I DEMONSTRABON METHOD i a Perform PRA, obtain Release Categories and mean frequency for each i Perform CRAC-type calculation for each Release Category, obtain f spectrum of whole-body doses at 0.5 mi dependent on weather (and l its conditional probability) Use results to construct CCDF " risk curve

• for dose at 0.5 miles No points on curve allowed to exceed both mean frequency of 10 6/ reactor year and 25 Rem Not 10-5 Allowed i

Mean 10 - l Freq. 'l 10 J 10 15 20 25 Rem EPRI ALWR Procram \\ I Prevention of Core Damage-Margins and Features l . No core damage for 6 inch LOCA 15% core thermal margin i . Larger primary and secondary inventories . Secondary plant requirements to reduce transient frequency l . Improved offsite power connections . No BWR recire loops - RIPS instead . PWR Th limited to 600 F . BWR full 3 division safety systems plus RCIC PWR safety injection with IRWST, no need to change to rec 9eulation, no low head pumps . P'NR improved EFW, safety grade bleed / feed cooling, higher pressure DHR system EPRI ALWR Procram b

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n ALWR PROGRAM Peta ACTIVITIES i In Support of Investment Prctection goals and Severe Accident treatment PRA Key Assumptions and Groundrules Document Define the important assumptions and reach closure with NRC - Use in PRAs of ALWR designs to demonstrate that top-level ALWR investment protection and safety targets are met Function Level PRA Evaluations - Test design capability to meet the overall safety targets - Feedback to design requirements on effectiveness of design alterr atives EPRI ALWR Procram 1 ~ l i l PRA KEY ASSUMPTIONS AND GROUNDRULES (KAG) DOCUMENT l Backcround . Requirements Document specifiies that a PRA be conducted to: - To assure a balanced design from a risk standpoint; 1 - To demonstrate that ALWR CDF and public risk criteria are met; j . USC requested ALWR PRA Subcommittee to define key PRA assumptions compatible with 1 x 10 E 5 CDF in late 1986 i KAG is Appendix A of Chapter 1, Volume 11 of the Requirements Document; KAG will be revised for passive plant later this year . KAG will be submitted to NRC as separate document then incorporated as Appendix A to Chapter 1 during rollup EPRI ALWR Procram l l

to ALWR PRA SUBCOMMITTEE W. R. Sugnet/S. T. Gray EPRI I D. H. Worledge EPRI J. Bickel/D. Dube NUSCO W. RasinM3. Hudson /B. Dolan Duke D. Leaver /R. Summitt DOE /ARSAP R. Jacquith C-E K. Holtzclaw/J. Duncan GE M. Hitchler/K. Vavrek W C. Stepp EPRI M. McCann/J. Reed JBA J. Gaertner EPRI S. Lewis SAROS EPRI ALWR Procram ( l l Definition of Core Damage Two step definition for analyst convenience Uncovering of active fual,and l Realistiethermal calculation shows temperat'ure in any node > 2200 F l - realistically 2700 2800 F no practical difference f or most sequences since recovery unlikely duringcore heatup through the additional 500 600 F considerable precedent surrounds 2200 F EPRI ALWR Procram 4

+ Modeling and Quantification . Event treeMault tree models for sequences, functions, front line sy: stems and supporting systems . Expreit modeling of all hard-wired dependencies . Preservation of logicalidentity of major contributors to CDF . Truncation value 1 x 10 E -8 or less . Point-estimate quantification using mean values . Common cause treatment per joint report EPRI NP-5613 -NUREG/CR 4870 . " Standard" reliability data specified - use more design-specific data only where important to the results EPRI ALWR Procram 1 l l l Uncertainty j . Probabilistic Risk Analyses point out uncertainty in Severe Accident evaluation (The same uncertainties also underly any deterministic evaluatens.) . Decisions must be made on the adequacy of plant designs given these uncertainties . ALWRTreatment Use Eganvalues as point estimates for comparison with the two top-level criteria - compatible with the point-estimate form of theCDF and Public Safety criteria - represent central tendency - also represent spread for right-skewed distributions l Provide qualitative discussion of sources of uncertainty and - l their magnitude l - Perform sensitivity studies where appropriate EPRI ALWR Procram a

i i ANNEX A RELIABILITY DATA BASE FOR ALWR PRAS l Table A13 i SUGGESTED INITIATOR FREQUENCIES FOR ALWRS l Suggested l Event Description Frequency l BWR Ti Turbine trip 2.3 T2 Loss of main condenser 0.49 i T3 Loss of feedwater 0.37 l T4 Loss of normal offsite power

• 0.035 Ts Loss of a major ac power bus 1.5 x 10'3 A

Large loss-of-coolant accident 5.8 x 10" S Small loss-of-coolant accident 5.1 x 10'3 i PWR T1 Reactor / turbine trip 2.8 T2 Loss of main feedwater 0.46 T3 Loss of normal offsite power

• 0.035 T4 Steamline break 1.5 x 10'3 T5 Loss of a major ac power bus 1.5 x 10'3 d

A Large lossef-coolant accident 3.4 x 10 St intermediate loss-of-coolant accident 3.4 x 10" S2 Small loss-of-coolant accident 3.0 x 10'3 R Steam-generator tube rupture 6.1 x 10~3

• For totalloss of off-site power, the co.xiltional unavailability of the reserve supply (0.22) must also be multiplied by this value. In addition, for the advanced PWR the frequency of demand for emergency power must also reflect the conditional unavailability of the full-load rejection capability.

Page A.A-6 l

v i ] I ~ 4 b ANNEX A RELIABILITY DATA BASE FOR ALWR'PRAS 1 l 1 Table A2 2 CUMULATIVE NON-RECOVERY PROBABluTIES Probability of not Probabliny of not Time (hr) recovering power Time (hr) recovering power 0.5 0.61 13 0.013 4 1 0.54 14 9.1 x 10 2 0.32 15 6.1 x 10'3 l 3 0.25 16 4.1 x 10'3 1 4 4 0.18 17 2.7 x 10 4 5 0.14 18 1.8 x 10 1 4 6 0.14 19 1.2 x 10 d 7 0.14 20 7.5 x 10 8 0.11 21 4.8 x 10" 4 9 0.11 22 3.1 x 10 d 10 0.11 23 1.9 x 10 4 11 0.071 24 1.3 x 10 12 0.019 Page A.A 16 l

_m. .. o ANNEX A . l RELIABILITY DATA BASE FOR ALWR PFMS i Table A3-1 COMPONENT FAILURE DATA Survey Component Failure Mode Failure Rate - Entry 4 Motor-operated valve Fails to operate on 4.0 x 10 /d 1 demand Transfers closed 1.4 x 10'7/hr 2 j 4 Air-operated valve Falls to operate on 2.0 x 10 /d 3 demand Transfers ofosed 1.5 x 10~7/hr 4 d Check valve (other Falls to operate on 2.0 x 10 /d 5 than stop) demand Transfers closed 2.0 x 10~7/hr 6 Reverse leakage (gross) 6.0 x 10'7/hr 7 4 Stop-check valve Fails to operate on 1.0 x 10 /d 8 ~ demand i Transfers closed 2.0 x 1'0'7/hr 9' Reverse leakage (gross) 6.0 x 10 7/hr 10 j 4 Check valve Intemal rupture 5.0 x 10 /hr 11 4 Manual valve Plugs / transfers closed 3.7 x 10 /hr 12 4 l Pressurizer safety valve Falls to open on 1.0 x 10 /d 13 (PWR) demand I 4 Falls to reclose 7.0 x 10 /d 14 4 Safety / relief valve Fails to open on 6.0 x 10 /d 15 (BWR) demand 4 Falls to reclose 6.5 x 10 /d 16 4 Pilot-operated relief Falls to open on 7.0 x 10 /d 17 valve demand l Fails to reclose 2.5 x 10 2/d 18 4 Motor-driven pump (all Falls to start on demand 2.0 x 10 /d 19 types) Fails to run 2.5 x 10'5/hr 20 4 Motor-driven pump Fails to start on demand 2.3 x 10 /d 21 (LPl/RHR) Falls to run 1.3 x 10'3/hr 22 4 Motor-driven pump Falls to start on demand 1.0 x 10 /d 23 (safety inj.) Fails to run 5.0 x 10 5/hr 24 Page A.A-19 a i l 4

i a ANNEX A ~ RELIABILITY DATA BASE FOR ALWR PRAS Table A3-2 MAINTENANCE UNAVAILABluTIES FOR THE BWR Train Unavailability System Shoreham PRA NUREG/CR-4550 Value Selected 1 Reactor-core isolation 1.1 x 10 2 3.5 x 10 4.0 x 10'3 4 cooling ) 4 4 High-pressure injection 4.0 x 10 3.5 x 10 4.0 x 10'3 4 4 Low-pressure injection 4.0 x 10 1.9 x 10 2.0 x 10'3 I Emergency service 2.0 x 10 1.9 x 10'3 2.0 x 10'3 4 water Standby-liquid control 2.5 x 10'3 3.5 x 10'8 3.0 x 10'3 4 4 Diesel generator

• 6.0 x 10 6.0 x 10 1

Table A3-3 MAINTENANCE UNAVAILABluTIES FOR THE PWR Train Unavailability Oconee Seabrook Value System PRA PSS NUREG/CR-4550 Selected 4 4 Turbine-driven AFW 3.8 x 10 4.6 x 10 6.0 x 10'.3 5.0 x 10'3 4 4 Motor-driven AFW 1.5 x 10 1.8 x 10 1.9 x 10'3 2.0 x 10'3 d 4 4 Safety injection 6.3 x 10 1.8 x 10 1.9 x 10 2.0 x 10'8 4 4 4 4 Residual-heat removal 2.0 x 10 2.3 x 10 1.9 x 10 2.0 x 10 4 4 4 Containment spray 2.0 x 10 1.8 x 10 1.9 x 10 2.0 x 10'3 4 4 Diesel generator

• 4.6 x 10 6.0 x 10 6.0 x 10'3 1

l The unavailability for diesel generators was taken from NUREG/CR 2989, which was also the source for NUREG/CR-4550. l Page A.A 23

m s c,. ANNEX A RELIABILITY DATA BASE FOR ALWR PRAS Table A3-4 COMMON-CAUSE FACTORS l Number of survey Coniponent Failure Mode Failures Entry Safety injection pump Faus to n, tart 2 of 2 1.4 x 10 2 of 4 4.7 x 10 24 3 of 4 7.6 x 10 4 or 4 3.6 x 10'3 Faus to run 2 of 2 8.0 x 10'3 2 of 4 7.6 x 10'3 3 of 4 1.7 x 10" 4 of 4 7.4 x 10 6 l l Emergency feedwater pump Faus to start 2 of 4 3.0 x 10 2 3 of 4 1'.3 x 10'3 4 of 4 4.1 x 10.s FaHs to run 2 of 4 3.0 x 10'3 3 of 4 2.6 x 10'8 4 of 4 7.1 x 10'7 Low-pressure injection pump Falls to start 2'of 2 1.4 x 10'1 2 of 3 5.4 x 10 2 3 of 3 1.4 x 10-2 Faus to run 2 of 2 3.9 x 10-2 2 of 3 1.9 x 10 2 3 of 3 1.6 x 10'.3 Containment spray pump Falls to start 2 of 2 1.3 x 10 Faus to run 2 of 2 (no evidence) Service wa:er/CCW pump Faus to start 2 of 3 5.6 x 10 2 3 of 3 1.7 x 10 2 2 of 4 3.8 x 10 2 3 of 4 4.9 x 10'3 4 of 4 2.2 x 10'3 Falls to run 2 of 3 3.6 x 10 24 3 of 3 3.9 x 10 2 of 4 2.2 x 10-2 3 of 4 1.1 x 10'3 4 of 4 1.8 x 10" Page A.A-24

e ANNEX A ALWR COMPONENT FAILURE DATA SURVEY 1. Motor-operated valves: failure to operate on demand Generic Sources Failure Rate (/d) 'j NUREG/CR 4550 3.0E-3 NUREG/CR 1363 4.0E 3 Oconae PRA 4.0E-3 Seabrook PSS 4.3E-3 Fhre plants (below) 4.6E 3 Arithmetic Average 4.0E-3 Geometric Average 3.9E 3 Plant-Specific Evidence Failures Demands Failure Rate Oconee 42 6,725 6.2E-3 Zion 31 14,677 2.1E 3 Indian Point 3 1,505 2.0E 3 Millstone 60 11,732 5.1 E-3 PWRX 69 10,052 6 9E 3 Total: 205 44,691 (SE-3 Value selected: .4.0E-3 Rationale: Value is representative of both generic data sources and l plant specific failure rates. 2. Motor-operated valves: transfer closed Generic Sources Failure Rate (/hr) NUREG/CR-4550 1.3E 7 NUREG/CR 1363 5.7E-8 NUREG/CR 2815 2.0E 7 Oconee PRA 2.3E 7 Seabrook PSS 9.3E-8 Fourplants (below) 1.4E 7 Arithmetic Average 1.4E-7 Geometric Average 1.4E-7 Plant Specific Evidence Failures Hours Failure Rate Oconee 0 1,890,000 1.BE 7 Zion 0 3,220,000 1.0E-7 Indian Point 0 1,429,000 2.3E-7 PWR X 1 817,399 1.2E-6 Total 1 7,356,399 1.4E-7 Value selected: 1.4E-7 Rationale: Value is representative of both generic data sources and plant specific fal!ure rates. Page A.A-26

6 g ) 1 1 Human Interactions 1 Employ EPRI SHARP process (EPRI NP-3583) or equrvalent - thorough, well documented process . Use THERP (NUREG/CR 1278)to quantify pre-existing .i test and maintenance errors l Use EPRI HCR correlation (EPRI RP-2487-1) for procedural actions and recovery actions that have a strong time dependence

• Prolonged misdiagnosis prevented by improved MMIS and symptom-based procedures - not modeled

. Human actions that result in initiating events are captured in the 1 E frequency data - not explicitly modeled l ) Y EPRI ALWR Procram ,me n u r External Events . Screening approach based on building design, l arrangement, and separation used to eliminate need for most external events + Bounding Seismic Hazard curve developed based on EPRI Seismicity Owners Group methodology 5 " Standard" fragility values specified, consistent with a areas for special consideration and seismic walkdown c,f design EPRI ALWR Prdaram

3 J q Containment and Offsite Consequence Analysis j l l . CET approach specified with MAAP as primary working tool, supplemented as necessary by more l detailed special purpose tools j . " Standard" site meteorology and demographics defined based on bounding most U. S. sites CRAC 2 specified as standard tool (with MAACS as alternate when sufficiently mature) i I I EPRI ALWR Procram Status of Function Level PRA Models . Modeling of internal events based on ALWR Chapters 1 5 is complete . Results of first cut have been used to steer some decisions on system requirements Offsite power ennections - Combustionturbine - DC power and cooling water system configurations . Modeling of external events in progress . Revision of models for ALWR Chapters 6 13 and "Rollup" changes planned for late 89 - account for decisions on support system configurations (esp. electric power) update for changes to chapters as a result of iterations to resolve comments EPRI ALWR Procram l

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CHAPTER 5 HYDROGEN CONTROL ALWR Requirements: Licensing Design Basis (deterministic case) - accommodate R oxidation of 75% active fuel cladding j - keep average & concentration below 13% ] Risk Evaluation Basis (spectrum of cases) - perform PRA - important SA sequences - realistic H2 generation from all sources provide containment arrangement that ensures mixing employ a hydrogen control system if necessary to control concentration -EPRI ALWR Procram I ALWR TREATMENT OF SOURCE TERM ISSUES Expect that Licensing Design Basis source term will be modified per results of recent research ALWR Requirements contain system configurations that rely on expected source term modifications PWR Containment Spray - delete spray additive - radionuclides removal rates based on realistic aerosol behavior Delete requirement for charcoal in filtration systems EPRI ALWR Procram NRC $99 M +

l l = ) ALWR TREATMENT OF SOURCE TERM ISSUES 4 Credit radionuclides t.crubbing in suppression pool: DF based on: particle size distribution per MAAP or equivalent calculation - scrubbing effectiveness per SUPRA (EPRI), SPARC (NRC), or equivalent calculations l Dose calculations will assume mechanistic release of l fission products into mntainment beginning about 30 minutes after core uncovering Containment leakage decreases with decreasing l containment pressure after an accident EPRI ALWR Procram l l J l' l l 1 s l l l l l l

ALWR LICENSING DESIGN BASIS SOURCE TERM DISCUSSION Presented to Division of Nuclear Reactor Regulation i by the Electric Power Research Institute March 15,1989 David E. Leaver Tenera, L.P. San Jose, CA

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l BASIC LDB SOURCE TERM APPROACH l o Utilize source term R&D of the last decade to define more realistic requirements for the 10CFR100 evaluation; simplify ALWR design where possible Limit changes to areas where there is general industry - o NRC agreement t W l

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in d S k E or a D G ie e ehl )d N N sg e A A ain H eh g r is u RC cwe in O inod l F t F l n O oat S t n o ST oe c AC yt m r ( BE al F san S F svia E E eot G cmne e ota N Ne cr r AH C u r E C o I N T h C nr O R A ee WR I t p T LP t s A A nwe s R R e E DO mlom l P E au ST n lo O ON iatnv P E tne D O M omd g N R c e yi E e R n A PI x U egai nrpm r E Q u a R E s s r r sr n R o A Aauf W D R EC A IT H C g L A dn i RR ex A yi WP am I T L R r N TO pu r s E N no) T ET uhl O RN ac r RE o ei P UM wpp CE t y st R ee( I mmR U ) Q uuW E sl R sop Av( l

A_W7 -YJ 70G E \\ CO \\~70. 7 EQU 7 E V E N-~S l Presented to Division of Nuclear Reactor Regulation by the Electric Power Research Institute and the Advanced Reactor Severe Accident Program March 15,1989 l David E. Leaver Tenera, L.P. San Jose, CA

INTRODUCTION ~ The control of combustible gases, principally hydrogen, is a well known issue for light water reactors The issue is real; hydrogen was generated and burned at 1 i TMI-2 in significant quantitles 1 Much has been learned about hydrogen generation and l combustion since the aftermath of TMI-2; based on this I experience the ALWR Program has devalooed an approach to hydrogen control to meet regulatory requirements and to ~ address severe' accident concerns l l I I s-720400-016 031389089E

B E X X X R N BD X X X X X X O L ITA T N E n S io t E a R P lu a L v E O s t k R is n e T R O m N s s e is v r C iu a s i s as is N s q B N O is s t e n B i y a m l r I B n i a A y ig T E T n o e io l n s G r a t N g D t O E is la a 3 is s R S e u g lu 1 a D R D g i a E n B Y P e s v n E 5 n r g 7 e k H F o e io nl n c is R t O is i r L R nt g W a F O N en o r r L lu o o A 5 3 a O co r f f E T L c d I i v 7 1 y s s e E P R n h t t r r n n h o o k f O O oe t P g y e e f f s r i s e m m o is is s s R C o S icr k f e e a a o d t f r r s t isy o iu iu n B B r eh q q o l la c h y s a t e e r r a co a e ic ic a r r af m R R n n o r r r s h h ai m W W y c c p s ha u L L e e e p CB S A A K T T A 2 3 4 5 6 7. 8 1

~ CHARACTERISTICS OF EPRI ALWR LICENSING DESIGN BASIS vs. RISK EVALUATION BASIS FOR HYDROGEN CONTROL Licensing Design Basis Risk Evaluation Basis I o Meet regulatory criteria o Meet 10-6, 25 rem (and EPRI optimization issues) o Use PRA and best-estimate deterministic o Deterministic analysis analysis with allowance for uncertainties o Degraded core in-vessel with reflood o Consider in-vessel and ex-vessel hydrogen o Must provide including potential for l appropriate level of core concrete attack conservatism o Take credit for the l margin which exists l from the LDB l i I s-720400-016 031389089E

KEY HYDROGEN REGULATORY REQUIREMENTS Amount of Hydrogen Generated: o 10CFR50.44 75% active clad oxidation (Mark Ill and ice condenser containments) o 10CFR50.34(f) 100% active clad oxidation (plants with license applications pending in 1982) (Severe accident policy statement and proposed 10CFR52 makes 10CFR50.34(f) applicable to advanced j plants) l Limit on global hydrogen concentration in containmerit: j a o 10CFR50.44 No requirement stated o 10CFR50.34(f) Uniformly distributed hydrogen < 10% or atmosphere non-l combustible Avoid local detonations of hydrogen which could cause loss i of containment integrity or failure of systems necessary for safe shutdown. s-720400-016 031389089E

l Equipment necessary for maintaining safe shutdown and j ~ I containment integrity shall be able to function during and after a hydrogen burn. Limit containment loads during a hydrogen burn to ASME Section lil, Service Level C. (~ 170% of containment design pressure for typical containment steels) l l 4 S-720400-016 031389D89E

~ ~ Consideration of Degraded Core: o 10CFR5.0.44 Analyze. consequences of hydrogen from up to 75% active clad oxidation during a degraded core accident; degraded defined as localized clad melting but not core meltdown; analysis to include period of recovery from degraded condition. o 10CFR50 App. A Containment design margin shall reflect consideration of 10CFR50.44 clad oxidation that may res' ult from degradation but not total failure of ECCS. o 10CFR50.34(f) No explicit mention of degraded core. 1 1 s-720400-016 03138E8%

EPRI REQUIREMENTS FOR ALWR HYDROGEN CONTROL Licensing Desian Basis Must be capable of handling hydrogen equivalent to o that from 75% active clad oxidation assuming degraded core in-vessel (optimization issue specified in Chapter 5) o Size PWR containment to limit hydrogen gas global concentration to s; 13 mole % in dry air (optimization issue specified in Chapter 5) o BWRs - Igniters or inerting o Containment arranged to promote mixing and avoid DDT geometries o Key equipment designed to survive global burn i o Containment structures must meet Service Level C for global burn o Provide local igniters in PWR where necessary l l l l s-720400-016 031389D89E

\\ [OPTNIZATION SLBACT sENirro \\ i 1 SUBJECT DEFINITION SLEJECT .DE.N.T.F.IC.A. TION. SUBJECT SCREENING AND CATEGORIZATION SUBJECT PAPERS SUBJECT c ASS.E.SS i.1 E E.......... _ _ _ _ _ _ _ _ _ _, \\ DFUT TO REQUIREMENTS DOCUMENT / Nureg 1197 Process for Treating ALWR Optimization Subjects

O, l. 1 t CONTAINMENT REQUIREMENTS FOR MIXING AND AVOIDING DDT GEOMETRIES Specific features to promote mixing and to avoid deflagration-to-detonation transition shall be considered in containment arrangement including: + Avoiding small. enclosed spaces with a hydrogen source (Chapter 6, Paragraph 4.3.2.5) Providing transverse venis along the length of + tubular enclosures (Chapter 6, Paragraph 4.3.2.5) Avoiding obstacles in the flow paths of tubular + enclosures (Chapter 6, Paragraph 4.3.2.5) Use of gratings vs. solid floors (Chapter 6, + Paragraph 4.3.2.5) j + Use of vented compartments (Chapter 6, Paragraphs 4.3.2.5,4.3.4.6) s-7204CO-016 031389D89E

~ Risk Evaluation Basis u o Must consider all sources of combustibles including hydrogen generation in-vessel and hydrogen and CO generation from core-concrete aitack. i o Consider potential for combustible gas generation beyond 75% (or even beyond 100%) active clad oxidation o Consider potential for debris articulation and co-i i dispersal of water and debris at vessel failure o Perform deterministic analysis on a sequence specific, best-estimate basis, and also allow for uncertainties in i key phenomena o Utilize realistic failure threshold for containment i 4 o Meet 10,25 rem (note margin compared to NRC safety goal) o Address any outliers from the plant specific PRA s-720400-016 i 031389D89E

WHY ARE 75% AND 13% THE CRITERIA THAT SHOULD BE USED FOR ALWR? o They yleid designs that will conservatively accommodate hydrogen from the range of degraded core accidents considered for the ALWR Licensing Design Basis l o They afford appropriate margin for the expected range of Risk Evaluation Basis conditions l o They yield the best overali design consistent with ALWR principles of simplicity and an integrated design approach to safety: Avoid an oversize, uneconomical dry containment Avoid a complex, containment wide network of Igniters and associated power sources, controls, and test and maintenance burden [

c. /

V s.,,_, 031389D89E

~ APPROACH TO REB HYDROGEN EVALUATION l l The ongoing evaluation includes both deterministic and, probabilistic analyses ) I The evaluation is addressing: 1 o Amount end location of H generated in-vessel ) 2 o Ex-vc3sel quantity of unoxidized metals (e.g., Zr) and subsequent degree of reaction j in containment at time of vessel rupture j o Likely. Ignition of H2 based on debris / gas temperatures, cavity conditions, and transport from cavity o The potential for global accumulation of detonable concentrations of hydrogen given the actual containment volume, geometry, and the expected steam concentration from cavity flood et'"% .b .y S-720400-016 031389D89E

~ ARSAP's preliminary findings are: o Sequences potentially leading to large quantitles of H2 generation ex-vessel have ignition sources present precluding accumulation of a detonable mixture o Sequences leading to large quantitles of total H2 generation without ignition are of extremely low likelihood l l Any risk-significant conditions (i.e., outliers) will be identified l o and resolved by the plant specific ?RA 1 i l l S-72040} 010 031389D89E

1. .. ~. '{ \\ TECHNICAL SUPPORT FOR THE EPRI HYDROGEN CONTROL REQUIREMENT FOR ADVANCED LIGHT WATER REACTORS MARTIN G. PLYS MANAGER METHODS DEVELOPMENT FAUSKE & ASSOCIATES, INC. 16WO70 WEST 83RD STREET BURR RIDGE, ILLINorS 60521 d E J

u ( h 1 OUTLINE J' 1. HYDROGEN GENERATION

SUMMARY

OF EVALUATIONS WITHOUT I REFLOOD

SUMMARY

OF EVALUATIONS WITH RECOVERY l 2. HYDROGEN COMBUSTION DEFLAGRATIONS PREVENTION OF DETONATIONS ~3.

SUMMARY

AND CONCLUSIONS E

THE PROPOSED ALWR HYDR 0 GEN REQUIREMENT THE CONTAINMENT AND COMBUSTIBLE GAS CONTROL SYSTEMS SHALL HAVE SUFFI-CIENT FREE VOLUME TO PREVENT ACCUMULATION OF GREATER THAN 13 PERCENT HYDROGEN BY VOLUME DRY AIR WHILE THERE IS OXYGEN AVAILABLE TO FORM A DETONABLE' MIXTURE. THE CONTAINMENT SHALL HAVE SUFFI-CIENT STRENGTH AND FREE VOLUME TO MAINTAIN INTEGRITY IN THE EVENT OF COMBUSTION OF HYDROGEN RESULTING l FROM OXIDATION OF 75 PERCENT OF THE l ACTIVE CLAD IN VESSEL, CONSIDERING RECOVERY. ( I sGr ( )

~ ( h, HYDR 0 GEN GENERATION:

SUMMARY

OF EVALUATION MAAP-DOE MODEL VALIDATION HAS BEEN PERFORMED FOR SEVERE FUEL DAMAGE (SFD) TESTS LOFT-FP2 TEST TMI-UNIT 2 MAAP-DOE HAS BEEN APPLIED To ALWR CASES: WITH RECOVERY WITHOUT RECOVERY ~ E

~. ( h HYDROGEN GENERATION:. REACTOR APPLICATION VALUES OF % MWR FOR EXPERIMENTS l CANNOT BE DIRECTLY APPLIED TO A REACTOR CASE - BOUNDARY CONDITIONS FOR FLOW,

POWER, AND HEAT LOSS DIFFER.

u i ANALYSIS OF REACTOR CASES YIELDS ABOUT 30%-35% MWR FOR CE ALWR.DBA LOCA AND BLACKOUT CASES WITH NO j RECOVERY. ANALYSIS OF RECOVERY SEQUENCES SHOWS A RANGE OF 30% TO 70% MWR. TMI REACHED 50% MWR DUE TO THE CORE REFLOOD - AND RECOVERY. d E )

O O l f 1 HYDROGEN DEFLAGRATIONS HYDROGEN-AIR-STEAM FLAMMABILITY LIMITS ARE WELL-KNOWN. THE IDEAL ADIABATIC ISOCHORIC COM-PLETE COMBUSTION PRESSURE HAS BEEN DETERMINED-FOR MIXTURES OF INITIALLY 13% HYDROGEN IN DRY AIR WITH VARIOUS ADDED STEAM CONCENTRATIONS. THE HIGHEST IDEAL POST-COMBUSTION PRESSURE POSSIBLE FOR A FLAMMABLE MIXTURE IS ABOUT 6.6 ATMOSPHERES. HYDROGEN STRATIFICATION (POOR MIXING IN CONTAINMENT) DOES NOT AFFECT THE STATIC PRESSURE RESULT. THIS CAN MEET APPLICABLE SERVICE LEVEL C PRESSURE LIMITS. m Ef

STATIC PRESSURE RISE AND FLAMMABILITY i RESULTS M)R 1hJ IN DRY AIR VITH 7 l STEAM ADDED TO 1004 HUMIDITY 1 2 3 4 T(K) X X, P P g 7 F 300 12.7 2.6 1.2 5.8 325 11.7 9.7 1.4 5.9 l l 350 9.9 24. 1.8 6.0 l j 375 7.4 43. 2.5 6.6* 380 6.9 47. 2.8 1Wet H m le Present. 2 2Steam mole present. 3 Initial pressure, atm. ' Ideal final pressure, atm.

• Probable final pressure 4-5 atm due to incomplete combustion.

/ l l l l l l l

6 o 0 s 1 g __ ~ [ _~l___1 1 x i l l ~I l E E i l l M R l s l l L I U B U s l 1 l MI S E N 1 l 1 l S T O S 1 1 I 1 1 X S E E 1 1 I 1 1 A o 1 T OR L 1 1 sL MPP P S E N 1 1 1 M U TL O 1 1 + 1 1 1 NB O 1 1 M 1 O B 1 I 1 EA 1 I T M l s/ 1 I l l g C P S 4 L 1 l / N O 1 I 1 C U l 1 C A O l I 1 l N B 1 l l 1 QO 1 M I M 0 l f 1 I 4 1 1

1. , I 1

1 1 I 'I g i 1 i I Ii C 1 1 1 1 ?- N 1 1 1 1 1 O sO 1 1 l N 1 a 1 l l I 1 l T L 1 N 4% 1 1 1 A MEE. E O 2 1 I A l l 1 R U l U * l 0 T I '- 9 I l I 3 R T 1 M I AS E T^ P= 1 1 1 I N T X S L S 1 1 1 1 S 1 E A E 1 P U c N 1 1 Tc 1 1 T M R MB N 1 1 5 E 1 1 O 1 P O M 2 1 C 1 1 P 1 C O 1 1 1 N 1 1 l 1 1 N l C 1 O 1 ~ 0 O 1 l 1 C 2 1 1 1 I 1 1 T M 1 1 1 1 1 1 S 1 l N E S 1 l 5 U R l 1 A R N l i 1 B U l l U EO E i l Y M B l l S l i I l T T T l l T S b O l l i I I C l S E gS L i l 0 S t I gO BT 1 R 1 l 1 O I P ~ oP M 1 A l R l P 1 l W l pM MI n. 1 I 1 1 L l O M l l l L 5 C A 1 I 1 I A L I .1 I 1 F l 1 i l 1 1 jd = 2 : _2 = 5=_5 @E353 5525=5 o O 6 4 g o 8 6 4 I 1 1 3 l x N m ZO_I<Cx wO2OO

4 [ HYDROGEN DETONATION: DETONATION LIMITS l l " DETONATION LIMITS" ARE DEPENDENT ON BOTH THERMODYNAMIC CONDITIONS AND I GEOMETRY. DETONABILITY-IS RELATED TO AN IN-TRINSIC MIXTURE PROPERTY: THE DETONATION CELL WIDTH, A. 1 i THE DETONATION CELL WIDTH DEPENDS UPON THE HYDROGEN AND OXYGEN CON-l CENTRATIONS (EQUIVALENCE RATIO), THE STEAM CONCENTRATION, AND TEMPERA-TURE.

6i J

4

v e e A I too 4 l l l Tube staaetse Smeted foils P m swre oscillattees (cm) 5 0 1s O 58 so V V to 3 5 ~ " lo, - E s ad E .O E 1 Ei E g I I t I g 4o to so to to so 5"2 ris. 2 ~ Figure 2-8 Measured values (McGill, Sandia) of the detonaticn cell width (1) as a function of hydrogen concentration (reproduced from (13]).

( D HYDROGEN DETONATIONS: INTRINSIC DETONABILITY IN ALWR CONTAINMENTS DETONATIONS ARE HIGHLY IMPROBABLE WHEN THE DETONATION CELL WIDTH l EXCEEDS THE GEOMETRIC SCALE: A2 TrD TUBE i A 2-D/6.5 OPEN REGION l EXAMPLE: IF TH'E CELL WIDTH IS 1 METER, A TUBE LESS THAN 0.3 M DIAMETER CANNOT SUSTAIN A DETONA-1

TION, AND AN OPEN SPACE LESS THAN 6.5 M CANNOT SUSTAIN A DETONATION.

PROPAGATION OF DETONATIONS OCCURS UNDER MORE RESTRICTIVE CONDITIONS. E ( ).

~. ~ I, [ HYDROGEN DETONATIONS: D EFFECT OF STEAM STEAM IS EXPECTED IN ALWR CONTAIN-MENTS DURING SEVERE ACCIDENTS SIGNIFICANT PRIMARY SYSTEM INVENTORY LOSS MUST PRECED5 HYDROGEN GENER-ATION. THE CELL WIDTH A INCREASES BY A FACTOR OF 5 WITH 10% H 0, BY A 2 FACTOR OF 25 WITH 20% H 0, AND BY A 2 FACTOR OF 125 WITH 30% H 0. 2 STEAM CONCENTRATIONS GREATER THAN 20% ARE EXPECTED (WITH OR WITHOUT SPRAYS) YIELDING A CELL SIZE NEAR 25 METERS REQUIRING AN OPEN SPACE OF 162 M DIAMETER. THEREFORE, FOR REALISTIC ALWR CASES, INTRINSIC DETONABILITY IS NOT = [ ACHIEVABLE. J

7 [ HYDR 0 GEN DETONATIONS: PERSPECTIVE 1 IN NTS TEST P-20, H2 = 12.9%, H O = 1 2 27.8%, A DEFLAGRATION WITH NO SIG-NIFICANT FLAME ACCELERATION WAS OBSERVED. IN NTS TEST P-21, H2 = 13.2%, H O = 2 27.4%, WITH FANS AND SPRAYS T'O INDUCE TURBULENCE, THE FLAME SPEED WAS ONLY TWICE THAT OF TEST P-20. THESE ARE EQUIVALENT TO 18% HYDROGEN IN DRY AIR. 1 VGES TESTS AT SNL WITH AND WITHOUT FANS HAVE BEEN CONDUCTED FOR DRY H - ::TcNATI Ns (TESTS B-69 TO B-74). 2 2 f <tpp-g ( )

~ J HYDR 0 GEN DETONATIONS: ~ h ( NATIONAL RESEARCH COUNCIL NATIONAL ACADEMY OF SCIENCES (NAS/NREC) POSITION THE NAS/NREC CONCLUSIONS SUPPORT THE ARSAP POSITION. "IT IS UNLIKELY THAT INITIATION OF DIRECT DETONATION NEAR THIS [ LEAN 3 LIMIT WOULD OCCUR". "IN WATER-HYDROGEN-AIR MIXTURES OF THE TYPE THAT MUST BE PRESENT IN MOST LARGE-VOLUME CONTAINMENTS FOLLOWING A HYDROGEN-PRODUCING

ACCIDENT, THE VALUE

[ LOWER LIMIT OF DETONATION 3 IS GREATER THAN 13 PERCENT". "IT IS ESTIMATED THAT A 20 PERCENT DILUTION BY EITHER CARBON DIOXIDE OR WATER VAPOR RAISES THE LEAN DETONA-TION LIMIT IN A REACTOR CONTAINMENT TO AROUND 13 PERCENT HYDROGEN". =f E J

4 HYDROGEN DETONATIONS: l DIRECT INITIATION l i THE ENERGY REQUIRED TO INITIATE A DETONATION IS ABOUT NINE ORDERS OF l l MAGNITUDE GREATER THAN THAT TO l l INITIATE A DEFLAGRATION SO IT-IS HIGHLY PROBABLE THAT A DEFLAGRATION WILL BE INITIATED PRIOR TO A DETONA-q TION. BECAUSE LARGE ENERGY SOURCES (TRIGGERS) DO NOT EXIST IN A CON-

TAINMENT, DIRECT INITIATION OF i

DETONATIONS IS HIGHLY IMPROBABLE. l i E J l I ~

) ~ [ D HYDROGEN DETONATIONS: INITIATION BY DDT DEFLAGRATION-TO-DETONATION TRANSI-

TION, DDT, HAS BEEN OBSERVED TO DATE AT A MINIMUM OF 15% HYDROGEN USING OBSTACLES TO ENHANCE FLAME ACCELERA-TION FOR AN UNVENTED CASE IN A 1.8 M X 2.4 M CHANNEL, 30 M LONG.

EXPERIMENTAL EVIDENCE INDICATES THAT THE CHANNEL SIZE CAN BE SCALED UP FOR LOWER HYDROGEN CONCENTRATIONS TO JUDGE THE POTENTIAL FOR DDT - BY USING THE RATIO OF DETONATION CELL WIDTHS. SCALING THESE RESULTS BY THE DETONA-TION CELL WIDTH, A 9 M X 12 M UNVENTED CHANNEL WOULD BE NECESSARY FOR DDT IN A DRY CONTAINMENT WITH 13% HYDROGEN -- AN UNREALISTIC CASE. \\s_ ) m

[ HYDR 0 GEN DETONATIONS: INITIATION BY DDT I 1 l SCALING THE UNVENTED DDT-RESULTS FOR A CASE OF 13% HYDROGEN IN DRY AIR PLUS 20% STEAM RESULTS IN A 22,5 M x 100 M TUBE GEOMETRY (OR LARGER) NECESSARY FOR DDT. SCALING THE VENTED DDT RESULTS FOR A CASE OF 13.% HYDROGEN IN DRY AIR RESULTS IN A 36 M x 48 M TUBE GEOMETRY OR LARGER FOR DDT. SINCE REAL CONTAINMENT REGIO'NS ARE

VENTED, AND REALISTIC CASES WILL INCLUDE STEAM, DDT IS EXTREMELY IMPROBABLE FOR AN ALWR.

E w J

\\ l Oob @ . l 50 - g 40-O DEFLAGRATION - NO OBSTACLES .zg $ DETONATION - NO OBSTACLES 20-O DEFLAGRATION - OBSTACLES E DETONATION - OBSTAGLES f g 20 - ZA 5 GGO O 10 - y a. h ^ A ^ O o 0 5 to w5 v,0 y .u 1 2 a 3D HYDROGEN MOLE FRACTION, % Figure 2-14 FI&lE apparatus DDT results (reproduced from (16]). 9 O

HYDROGEN DETONATIONS: STRATIFICATION I l POOR MIXING OF HYDROGEN (STRATIFICATION) CAN RESULT IN HYDROGEN RICH REGIONS ( C > 13% DRY BASIS) AND THEREFORE HYDROGEN POOR ) REGIONS (C < 13% DRY BASIS). I THE LOCAL POTENTIAL FOR DETONATIONS l IN INCREASED IN HYDROGEN RICH l REGIONS.

HOWEVER, INTRINSIC DETONABILITY IN A STRATIFIED HYDROGEN-RICH REGION IS MORE DIFFICULT BECAUSE THE "OPEN REGION" CRITERION APPLIES, AND DDT IS MORE DIFFICULT BECAUSE THE REGION IS ESSENTIALLY FULLY VENTED.

FURTHER, DETONATIONS IN HYDROGEN RICH REGIONS CANNOT PROPAGATE TO HYDROGEN POOR REGIONS.

E J

( 1

SUMMARY

AND CONCLUSIONS THE EPRI REQUIREMENT FOR HYDROGEN CONTROL IN ALWR'S.IS SUPPORTED FOR BOTH AN IN-VESSEL HYDROGEN SOURCE AND CONTAINMENT RESPONSE. THE HYDROGEN SOURCE TERM IS EVALUATED BY CONSISTENTLY APPLYING MODELS FOR BOTH SIMULATION OF IMPILE EXPERIMENTS AND PREDICTION OF REAC-TDR RESPONSE. CONTAINMENT RESPONSE IS EVALUATED BY DEMONSTRATING THAT DETONATIONS ARE EXTREMELY UNLIKELY GIVEN REALISTIC COMPOSITION, TEMPERATURE, AND GEOMETRY. . CONTAINMENT RESPONSE IS ALSO EVALUATED BY DEMONSTRATING THAT THE MAXIMUM POST-BURN PRESSURE IS LESS THAN THE CONTAINMENT CAPABILITY.

[ D IDCOR-NRC. COMBUSTION ISSUES ARE RESOLVED BY MAAP-DOE SIx POINTS OF DIFFERENCE BETWEEN MAAP AND HECTR ARE' MENTIONED IN NUREG/CR-4993 AND RESOLVED BY MAAP-DOE. 1. IGNITION CRITERIA MAAP-DOE USES A GENERAL H -CO-2 AIR-H 0-C0 -N2 FLAMMABILITY LIMIT 2 2 DIAGRAM. 2. PROPAGATION CRITERIA MAAP-DOE ALLOWS UPWARD PROPAGA-TION OF INCOMPLETE BURNS AND GENERAL PROPAGATION OF COMPLETE BURNS WHEN THE FLAMMABILITY LIMITS ARE SATISFIED. 3. BURN TIME MAAP-DOE CONSERVATIVELY PREDICTS BURN TIME BASED ON FLAME SPEED AND GEOMETRY.

[ N IDCOR-NRC COMBUSTION ISSUES ARE RESOLVED BY MAAP-DOE (CONTINUED) 4. COMBUSTION COMPLETENESS MAAP-DOE ACCURATELY PREDICTS COMBUSTION COMPLETENESS. 5. IN-CAVITY OXIDATION .MAAP-DOE ALLOWS IN-CAVITY OXIDA-TION WHEN THE TEMPERATURE-h DEPENDENT FLAMMABILITY LIMITS ARE SATISFIED AND ALLOWS FOR AUTOIG- { ~ NITION. 6. NATURAL CIRCULATION l MAAP-DOE IS BEING UPGRADED TO INCLUDE STATE-OF-THE-ART NATURAL CIRCULATION CORRELATIONS. d E J

.( i 6 l l l l i I 4 PRESSURE: 1 ATM ] ~ 1 i 3 1 j .8 5 .6 .4 a mzW .2 2 9 H g .1 g .08 m .o 6 m E .04 E3 .02 I - 0 1 0 10 20 30 40 50 60 70 HYDROGEN IN AIR, PERCENT BY VOLUME Figure 2-13 Minimum ignition energy for hydrogen deflagrations (reptoduced from [15).

...y ._....v ( 3 -0 1000 / W i ?. l i i m 500 i O g, I g S, 2.3/4f ~ t a. ~ ~ z C s ~ 100 f \\

1. /

/ f t Z 20 / 2 e 10 :- t i s e t s Om 2 4, NO GO b \\\\ ;. 8 acaoE SANDIA GO 5 r .\\ JA, 0.5 10 20 30 40 50 80 % H in H - AIR 2 2 I, ,,,,,t I 0.3 0.5 0.7 0.9 1 1.21.41.61J 2 2.5 3 4 EQUIVALENCE RATIO - d EDWARDS HYDRODYNAMIC THICKNESS MODEL 1 2 ZIL'DOYlCH-LEE SLAST MODEL 3 LEE HYDRODYNAMIC THICKNESS MODEL 4 DETONAY10N KERNEL MJDEL 5 SURFACE ENEMGY MODEL 8 VASILIEY CELL ENERGY MODEL 7 CHEMICM. ENERGY MODEL 3 WORW.10NS &Wtr.m Figure 2-12 Critical detonation initiation charge as a on of composition '. reproduced from

I 1 ALWR TECHNICAL BASIS FOR ADDRESSING DIRECT CONTAINMENT HEATING ROBERT E. HENRY FAUSKE & ASSGCIATES, INC. 16WO70 WEST 83RD STREET l BURR RIDGE, ILLINOIS 60521 i PRESENTED AT: NRC MEETING TO DISCUSS ALWR SEVERE ACCIDENT APPROACH MARCH 15, 1989 PALo ALTO, CALIFORNIA

~ ~ ALWR APPROACH CONTAINMENT GEOMETRY MINIMIZES POTENTIAL FOR DEBRIS DISPERSAL. .-GEOMETRY IS NOT REQUIRED TO HAVE COMPLETE RETENTION. PRIMARY SYSTEM DEPRESSURIZATION AS WOULD BE DIRECTED BY OPERATOR PROCE-DURES GREATLY REDUCES THE POTENTIAL-DEBRIS DISPERSAL.

P e. o 9 e 6 e L_ l h \\ I e l \\ T N_, e 4 n. L ) ___m_____._

1 Py / A CORE DEBRIS A y n f U c A \\ N/H/HHHHHH/HHH/HHHHHH/HH/HH/HH) l, = L 1 L P mane.oe u..a l l l l 1 l 1

m. 1 l l l l I pCORE DEBRIS en g P, U$ l P 3 HHHHHHHH/HHUfMA o l l L L1__________.

9 .v \\ l j I l l l l TOP PORT to tsJ m)- l - LEVEL 8 f te r 4e ma - SUPPORTSTRUCTURE -- LEVEL 8 f ce sJe ab - .I LEVEL 4 i p l te fJes ma f p LEVEL 2-f j to s.se =) - LEVEL 2 f , CHUTE te a.as ma - lEii / - LEVEL 1 ) / , CAVITY te s.as m6 j k i MELT Y- / OENERATOR i test Figure 4-1 Schematic of the DCH-1 apparacus in the SL*RTSET direct heating test f acility.

N lI ( lIi11 1a_ )Ino e d l da b ee a tR 4 2 l i 0 56 i sr 8 3 3 7 3 2 40 a oe 3 9 H 0 5 8 9 0 v 30 pw 1 1 CD 0 0 2 10 6 3A 33 eo 1 1 0 0 DL to se N sh at M no e d l da b ee a tH 3 2 l i sr 4 3 2 8 0 56 i 4 3 2 oe 0 9 H 0 5 8 0 40 a 0 v 18 pw 1 1 C D 0 0 2 10 6 4A 31 eo 1 1 0 0 DL t T S o se E N sh at T M HC D R n O o F e d l da S b eo N a t R O 2 2 l i S 0 56 i sr 2 0 0 3 3 3 40 a oe 3 5 I H 0 5 3 8 R C 5v 73 pw 0 1 A D 0 0 2 10 6 6A 23 eo 1 1 0 0 P DL M t O o se C N nh iat L F( E D OM )J ]J dd dd ee ee rs rs 2 ur ur 1 2 se se ap 9ap8 E 0 56 40 00 .6

1. M i es1 L

H 0 5 8 5 1 es1 B C 45 01 8Mi0 A D 0 0 2 10 2 25 21 6 D5 0 D0 T g g ns ns i s i s sa sa UM uM-( ( ) s m r p e ( t e er n / mu a s s as e ae e r/s rs M t r ) est ae ) da) u a ) t nn Pr ms el g) t) eP a rM P eoe P ps suk g aK ) mim l e( a rc(k r( E u( M at e al t M ei ( e ( s ( rcr ce id ptd pd sd aiu i s ) S ed sred me d ee d Pds rs) ) ) a t e i at e et e rt e t e e ) ' P ser DPcr Tc r Pc r e v 'm ea m mim M il u i u i u i u t re rds seds kd s kd s sPM m ( ( ( m( ( ( bea snea ae a ae a e od T en v s P C v ere aire er e er e GaA A L LA P DPM MFPM PP M PP M 1 2 3 4 5 IiIl lll

( h TECHNICAL SUPPORT FOR THE EPRI DEBRIS C00 LABILITY REQUIREMENT FOR ADVANCED LIGHT WATER REACTORS MARTIN G. PLYS

MANAGER, METHODS DEVELOPMENT FAUSKE & ASSOCIATES, INC.

16WO70 WEST 83RD STREET BURR RIDGE, ILLINOIS 60521 // e a

~. t (" OUTLINE 1. DESIGN REQUIREMENT QUANTIFICATION DERIVATION DEBRIS POWER l REFERENCE HEAT FLUX DESIGN MARGIN l 2. PHYSICAL PROCESSES ~ DEBRIS CONFIGURATION l DEBRIS-COOLANT HEAT TRANSFER l 3. EXPERIMENTAL SUPPORT l 4. ANALYTICAL SUPPORT 5.

SUMMARY

sG"d ,)

i. 1 [ DESIGN REQUIREMENT QUANTIFICATION D l DERIVATION 1 L THE HEAT FLUX FROM DEBRIS TO COOLANT MUST REMOVE DECAY POWER: l Q A=Q j DEBRIS D l UECAY POWER IS A FRACTION OF REACTOR l THERMAL POWER: 'Q =FQ D DR 1 f THE HEAT FLUX IS REDUCED FROM A REFERENCE VALUE BY A DESIGN FACTOR: i Q =FG DEBRIS S REF 4 1 E t

r DESIGN REQUIREMENT QUANTIFICATION DERIVATION i THE DESIGN REQUIREMENT CAM THEREFORE BE SPECIFIED IN TERMS OF FLOOR AREA PER RATED MEGAWATT REACTOR POWER: F 2 I D y = FQ VW R S REF l l i i ' ~ ~ - - ~.. - _ _ _ _ _ _ _ _ _ _

s [ DESIGN REQUIREMENT QUANTIFICATION 3 DERIVATION THE REQUIREMENT IS QUANTIFIED BY CHOOSING THE FOLLOWING VALUES: DECAY POWER IS 1% OF RATED POWER: F, = 0.01 A DESIGN FACTOR OF 2 IS CHOSEN F = 0.5 s THE REFERENCE HEAT FLUX IS SLIGHTLY LESS THAN THE SATURATED FLAT PLATE CRITICAL HEAT FLUX (CHF) Q = 1.9 W/M2 REF E J l ___-m______m2__._m_-___m_at_____-_____

~ 1 d e DESIGN REQUIREMENT QUANTIFICATION DERIVATION THE DESIGN REQUIREMENT IS THEREFORE 2 h=0.02g R r e E J w

I DESIGN REQUIREMENT QUANTIFICATION DEBRIS POWER l Two FUNDAMENTAL ASSUMPTIONS ARE EM-BODIED IN THE CHOICE OF F I 1. DECAY POWER IS THE ONLY ENERGY SOURCE 2. A FRACTION OF FULL POWER IMPLIES TIME SINCE SCRAM AND BURNUP i l i 3 i sdI J

c r DE51bN KtuUIREMENT QUANTIFICATION ~ 'S g DEBRIS POWER l WHY CONSIDER DECAY POWER ALONE? l EXOTHERMIC REACTIONS BETWEEN CONCRETE OFFGAS AND ZR, CR, AND FE CAN OCCUR. THE ZR REACTIONS IS A STRONG SOURCE OF i

ENERGY, CR IS LESS IMPORTANT, FE IS NEGLIGIBLE.

l THE CHEMICAL ENERGY SOURCE IS LIMITED ~ BY THE AMOUNTS OF 1R AND CR. LONG-TERM COOLABILITY THEREFORE IM-PLIES DECAY HEAT REMOVAL ONLY. sGr ( )

l /' DESIGN REQUIREMENT QUANTIFICATION 'S DEBRIS POWER I WNY 1% OF FULL REACTOR POWER? l CONSIDER THE ANS STANDARD DECAY CURVE CONSIDER VESSEL FAILURE TIME AT 3 HOURS OR MORE AFTER SCRAM: COR-RESPONDING TO THE DOMINANT SEQUENCE (BLACKOUT) l DECAY P'OWER IS 1% OF FULL POWER AT 2 HOURS l l DECAY POWER IS ONLY 40% HIGHER, OR 1.4% FULL POWER, AT 1 HOUR. \\ DUE TO IN-VESSEL FISSION PRODUCT l

RELEASE, ONLY 75% OF DECAY POWER WILL BE PRESENT IN DEBRIS.

1 E j i l l I

/ DESIGN REQUIREMENT QUANTIFICATION 'S REFERENCE HEAT FLUX l 1.0 MW/M2 REPRESENTS A BEST-ESTIMATE LIMITING STEADY HEAT FLUX FILM BOILING CANNOT BE MAINTAINED ON CORE DEBRIS 1 SATURATED FLAT PLATE CRITICAL HEAT FLUX AT 1 ATMOSPHERE PRESSURE = 1.3 MW/M2 - SO NO COOLANT LIMITATION TEXTBOOK STEADY STATE BOILING CURVES DO NOT APPLY TO THE QUENCHING OF CORE DEBRIS l l THE DEBRIS WILL CRACK AND FRAGMENT - IT WILL NOT BE AN IMPERMEABLE SLAB RESISTANCE TO HEAT TRANSFER IS IN THE DEBRIS l a E J

~ / DESIGN REQUIREMENT QUANTIFICATION i SAFETY FACTOR A SAFETY FACTOR OF TWO IS SIMPLY ASSUMED AND IT COVERS INHERENT UNCERTAINTY IN MEASURED HEAT FLUX INHERENT (BUT TRANSIENT) UN'ERTAINTY C IN DECAY POWER 1 ADDITIONAL CONTRIBUTIONS TO THE SAFETY MARGIN ARE: DYNAMIC NATURE OF DEBRIS-WATER INTERACTIONS AND ASSOCIATED HIGH HEAT TRANSFER INCREASE OF CHF AND DRYOUT HEAT FLUX WITH PRESSURE - ABSENCE OF VOLATILE FISSION PRODUCTS

~ [ PHYSICAL PROCESSES: ) DEBRIS CONFIGURATION TWO TYPES OF CONFIGURATIONS CAN EXIST: 1. DISCONTINUOUS DEBRIS BED. CRACKED OR FRAGMENTED SOLID DEBRIS. p 2. CONTINUOUS DEBRIS SLAB OR POOL. -_____._.._-__m__

~. ( PHYSICAL PROCESSES: DEBRIS CONFIGURATION DISCONTINt' US DEBRIS BED: - SOLID DEBRIS CAN FORM WHEN WATER QUENCHES DEBRIS ENTERING THE CAVITY COOLABILITY DEPENDS ON PARTICLE SIZE AND POROSITY E w J

f / PHYSICAL PROCESSES: h 1 DEBRIS CONFIGURATION CONTINUOUS DEBRIS BED: - SOLID,

LIQUID, OR MIXED DEBRIS CAN FORM AT VESSEL FAILURE WHEN INSUFFICIENT WATER IS PRESENT TO QUENCH DEBRIS ENTERING THE CAVITY IF SHALLOW AND SOLID, COOLABILITY DEPENDS UPON THICKNESS AND VOLUMETRIC POWER.

IF DEEP OR PARTIALLY MOLTEN, COOLABILITY DEPENDS UPON LOCAL

FREEZING, CRACKING, AND WATER IN-GRESSION.

~ I s_ _/

~ I / DEBRIS CONFIGURATION EVOLUTION i DISCONTINUOUS DEBRIS CAN DRY OUT, REMELT AND BECOME A CONTINUOUS POOL IE THE WATER SUPPLY IS

DEPLETED, THE POROSITY IS LOW (FRAGMENT SIZE TOO SMALL),

OR BED AREA IS LOW. CONTINUOUS DEBRIS CAN SOLIDIFY,

SHRINK, CRACK, AND BECOME A DISCON-I TINUOUS DEBRIS BED IE WATER IS PRESENT AND THE CAVITY AREA IS

LARGE, AND THIS PROCESS IS AIDED BY AGITATION FROM CONCRETE OFF GAS.

THE ALWP.

ESIGN PROVIDES FOR A CONTIt.".US NATER SUPPLY.

l I

CD DD CW VR ee re re aa ab yb vt sa sc cr or ie ki ui tr et is ts y lo ~r ? ?i n n g ? D E D B i R c N M s I r o u c S c o a h n C c d k r t O y [' W i n o a n N i E t u F = ~ t e o I g s u C r s URAT I(M EUD IR I I O N N D L C o r i o y t n c o t t u l i r n aI r e uo c w u k a s i t n e g r h d I I lllfi l

i. l DEIRIS CONFIGURATION EVOLUTION 3 EXAMPLES l 1. ExTENS;VE CAVITY FRAGMENTATION CoNTINUOw; lATER SUPPLY 2. INCOMPLETE FRAGMENTATION CONTINUOUS WATER SUPPLY 3. No FRAGMENTATION CONTINUOUS WATER SUPPLY >E )

(' DEBRIS EVOLUTION D EXAMPLE 1 RV FAILURE EXTENSIVE FRAGMENTATION IN CAVITY i f STABLE

e [ DEBRIS EVOLUTION 3 EXAMPLE 2 RV FAILURE I N C O M P L.'E F R A G M E N T A T I O N DEBRIS HEATUP ,r

0NCRETE ATTACK DEBRIS MELTING MELT SPREADING CRACKING d

5 STABLE y

e 0 4 [ DEBRIS EVOLUTION 3 EXAMPLE 3 RV FAILURE ~ MELT SPREADING CONCRETE ATTACK 1 CHEMICAL SOURCE EXHAUSTION CRACKING ) STABLE // +

PHYSICAL PROCESSES: DEBRIS-COO.LANT HEAT TRANSFER 1. DEBRIS BED C00 LABILITY - DEPENDS ON PARTICLE SIZE (POROSITY) ENHANCED BY UNEVENLY DISTRIBUTED DEBRIS 2. FILM BOILING CANNOT BE MAINTAINED OVER A ROUGH OXIDE SURFACE 3. NUCLEATE BOILING - OCCURS AFTER FILM BOILING COL-LAPSE LIMITED TO CHF 4. TRANSIENT BOILING - OCCURS DURING CONTACT OF VERY HOT DEBRIS AND WATER - VIOLENT - HEAT FLUXES EXCEED STEADY VALUES (

EXPERIMENTAL SUPPORT ) DEBRIS FRAGMENTATION DEBRIS BED HEAT TRANSFER CRUST STABILITY l WATER INGRESSION 1 FILM BOILING INSTABILITY FAI THERMITE TESTS TMI-2 OBSERVATIONS Y N

I EXPERIMENTAL SUPPORT \\ DEBRIS FRAGMENTATION BENZ (ISPRA, 1979) AND SPENCER (ANL, 1987) EXPERIMENTAL DATA FOR FRAGMEN-l TATION OF METALLIC AND OXIDE MELTS IN WATER. SMALL WATER VOLUMES ALLOW DEBRIS " CHANNELING" THROUGH WATER WITHOUT I JET BREAKUP. WITH LARGE WATER VOLUMES, BENZ'S SMALLEST MEAN PARTICLE SIZE WAS 1.7 MM. 1 WITH SMALL WATER VOLUMES, SPENCER OBSERVED

LARGE, POROUS SLABS OF 43-57% THEORETICAL DENSITY.

d l (' ') EXPERIMENTAL SUPPORT DEBRIS BED HEAT TRANSFER l A LARGE BODY OF EXPERIMENTAL DATA CONFIRMS THAT THE DRYOUT HEAT FLUX FOR 2 MM PARTICLE BEDS IS ABOUT 1 MW/M2 THESE EXPERIMENTS CONSIDER ONE-DIMENSIONAL COUNTERCURRENT FLOW IN AN EVENLY DISTRIBUTED, LEVEL DEBRIS y BED MULTIDIMENSIONAL FLOW WOULD OCCUR IN EVENLY DISTRIBUTED DEBRIS AND WOULD BE EXPECTED IN A REACTOR CASE HEAT TRANSFER RATES UP TO A FACTOR-OF TWO HIGHER ARE POSSIBLE FOR I UNEVEN DEERIS BED k E" ( J

i / EXPERIMENTAL SUPPORT CRUST INSTABILITY EXPERIMENTAL EVIDENCE INDICATES THAT OXIDE CRUSTS ARE UNSTABLE AT SCALES OF ABOUT ONE METER. SPENCER (ANL, 1988) OBSERVED STABLE l CRUSTS IN A 25 CM X 25 CM APPARATUS, l AND UNSTABLE CRUSTS IN A 50 CM X 50 CM APPARATUS WITHOUT WATER. MOST DATA IS FOR METALLIC CRUSTS AT 20 CM SCALE - THESE CRUSTS ARE STABLE (SWISS) AN ALWR CAVITY WOULD BE 870 CM IN

DIAMETER, OR 20 TIMES LARGER.

1 WATER AND CONCRETE OFFGAS AGITATION WOULD PROMOTE CRUST INSTABILITY. ud v

e a ( \\ EXPERIMENTAL SUPPORT WATER INGRESSION CRUST INSTABILITY - CRACKING - ) ALLOWS WATER INGRESSION INTO DEBRIS. WATER INGRESSION TO A DEPTH OF 12 METERS WAS OBSERVED IN A COOLED LAVA (OXIDE) FIELD AT GRIMSVOTN BY BJORNSSON '(1982). THIS WATER INGRESSION OCCURRED WITHOUT SUBSTANTIAL GAS AGITATION. GAS AGITATION FROM CONCRETE DECOM-POSITION WOULD ENHANCE CRACKING IN A REACTOR CASE. k V >$I \\s- ,)

I EXPERIMENTAL SUPPORT \\ FILM BOILING INSTABILITY EXPERIMENTAL DATA CLEARLY DEMONSTRATE THAT FILM BOILING IS DESTABILIZED ON OXIDE SURFACES DUE TO LOW THERMAL CONDUCTIVITY AAD ROUGHNESS. PERIODIC CONTACT ALWAYS OCCURS IN FILM BOILING - LEADING TO COLD PATCHE~ 'N OXIDE SURFACES AND DES-TABIL._ATION. SURFACE AREA WAS ENHANCED BY A FACTOR OF 3 DUE TO ROUGHNESS IN SCENCER'S TESTS. DESTABILIZATION LEADS TO NUCLEATE BOILING. l

\\ I EXPERIMENTAL SUPPORT FAI THERMITE TESTS CONTACT OF WATER WITH HIGH TEMPERA-TURE MELTS RESULTS IN VIOLENT INTERACTIONS AND HIGH HEAT TRANSFER l RATES. ) FAI THERMITE TESTS DEMONSTRATED HEAT TRANSFER RATES AN ORDER OF MAGNITUDE ABOVE THE STEA11Y STATE VALUES. MOLTEN: ~ 107 w/M2 FROZEN: ~ 106 w/g2 ,/ L

0 t: _ - - E__ E~ 1 5 1 l ~ t l i l l l l l 0 l l l l l l 5 i l lI 4 l l l l l l l l l l 1 n l l bi o l i '0 l t I l 4 c l 1 l u N d 1 1 1 1 n 1 A O o 1 1 1 0 c 1 T I 1 1 I T 1 5 t A 1 l n U 3 e l D 1 l L i 1 l 1 I s M O l n 1 1 l a 0 r 1 O S l 1 0 ) t 1 R N 1 1 3 c o 1 1 F l t e O 1 l 1 l d s e 1 I E T 1 1 1 r 0 ( 1 1 T C l a 1 5 p 1 l A U 1 l m 1l 2 E l o M D 1 l c 1 l T N M 1 I 1 1 x 1 l I u 1 S O 0 l 1 E a lll T 1 f 0 1 E G 1 l t l 2 a 1 l 1 E A l e. 1 l hn 1 l o 1 l 8i 1 1 0 t 1 l t u 1 5 l sl 1 l 1I l 1 eo 1 l Ts 1 l 1 l 1 l 7 1 l 1 1 0 1 a A 1 1 1 0 3 l l l l 1 1 e l 1 r l 1 l u l l g l l i 8 ll l F 0 l s A 1 l 5 1 l T a A l I l SI m A l I l E ll a A 1 1 _iy=___1 Tl tf l 1 t~

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0 i 1 00 0 2 0 0 1 1 1 1 T ^uEsg3 C c

0 0 t = - = ~ _ E-i 5 l l t i l l l l l l 0 l i l f 5 l I l l 4 l 1 I 1 1 l I l 0 l i l l l i 0 l h4 l l l i l l e l l m l i l i l 0 l t l l I 5 f l l o l l 3 l l l l n l l o I l l l i l t l 0 c 1 l I n l e! 0 i 1 ) u g f 3 c l 1 l 1 l 1 a e l 1 l 1 s s a l 1 0 l 1 l I ( 5 x l l 1 u 2 E l 1 l l 1 f l 1 Mt l 1 l 1 a l 1 I 0 e l 1 T l I h l = 0 1 l 9 2 l 1 s 1 i 1 t s 1 s i 1 e s 1 T 0 i 1 g I 5 i 9 l 1 i 1 1 1 g 1 i 1 3 g 1 i 1 m e g 1 r i 0 1 u g I i 0 g l 1 i i 1 1 F i 1 i 1 i 1 i 1 i 1 i 9i 1 0 1 i l 5 l l Ti n 1 g 1 Si a 1 i 1 Ei m 1 g 1 Ti g ~__E r3 ~ E_ E-1 g 1 0 0 1 1 2 0 g 0 0 1 g 1 1 y fEs33 C O

\\ / CONSISTENCY WITH TMI-2 OBSERVATIONS DEBRIS COOLING IN THE TMI-2 LOWER HEAD REQUIRED ABOUT 0.47 MW/M2 pon DECAY HEAT. THE TRANSIENT HEAT FLUX IS ESTIMATED TO BE 2.7 TO 5.5 MW/M2 DURING THE DEBRIS QUENCH CONSIDERING STORED ENERGY AND ASSUMING 10-20 MINUTES TIME. THE TOTAL DEBRIS-COOLANT HEAT FLUX l AT TMI-2 WAS THEREFORE 3.2-6.0 MW/M2 AFTER FUEL RELOCATION. THIS IS CC"~ISTENT WITH CHF AT THE TMI-2 S', A PRESSURE. i u i

) b ANALYTICAL SUPPORT ANALYSIS OF SWISS TESTS DEBRIS BED C00 LABILITY CORE-CONCRETE INTERACTION SIMULATION SREADING REGIME ANALYSIS \\ { ( )

h '5 g" g ' ANALYTICAL SUPPORT SWISS TESTS SWISS: SUSTAINED WATER. INTERACTIONS l WITH STAINLESS STEEL . 50 KG STAINLESS STEEL INITIALLY 1925 K CRUCIBLE DIAMETER 21.6 cM LONG-TERM INDUCTION HEATING WATER ADDITION: SWISS-1: 32 MINUTES AFTER STARY SWISS-2: IMMEDIATE ADDITION H E" t J

\\ b ANALYTICAL SUPPORT SWISS TESTS 4 THE EQUIVALENT NET INPUT POWER, AFTER CONSIDERING SIDEWARD LOSSES, l NOT COUNTING CHEMICAL ENERGY, WAS: 1 SWISS-1: 0.9 MW/M2 SWISS-2: 1.2 MW/M2 THE MEASURED UPWARD HEAT FLUX WAS ABOUT 0.8 MW/M2 IN EACH TEST. A STABLE CRUST (METALLIC, SMALL SCALE) PREVENTED WATER INGRESSION. MAAP-DOE RESULTS AGREE WELL WITH EXPERIMENT. Q a

~ [ ANALYTICAL SUPP0P.T CORE-CONCRETE INTERACTION SIMULATION l 1 DECOMP-DOE IS USED TO SIMULATE l QUENCHING OF ALWR CORE DEBRIS IN AN ALWR CAVITY DESIGN 9 DECOMP-D0E HAS BEEN BENCHMARKED AGAINST.MANY EXPERIMENTS:

SWISS, BETA,

TURC, SURC WITH EXCELLENT RESULTS l

t. -1 f ANALYTICAL SUPPORT DEBRIS BED C00 LABILITY UNCERTAINTY IN DEBRIS BED HEAT TRANSFER IS DOMINATED BY THE CAPABILITY TO PREDICT DEBRIS PAR-TICLE SIZE DEBRIS BED DRYOUT HEAT FLUX IN-CREASES WITH PRESSURE - MULTIDIMENSIONAL EFFECTS BoTH INCREASED PRESSURE AND NON-UNIFORM DEBRIS BEDS WOULD BE EXPECTED IN AN ALWR CASE.

l l

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0 1 1 x . 6 O A l 9 i O O O O i i i A i Q i O i e O 9 c i l i n a i t i s i i d i i O 0 n i o C i i . 1 s i E o i r l Q 8S i e O i P i M i O D E C i H O M D E i i T i I 4 P g 8 i T C E O R l i 7 D IS E r R L Z 0 i U A D i O0 gE D i I TX k D i i S A A 2 U R 0D O ~. i i 1 E D 2A 7 i I 1 l l i 1 S D P EI1III i ~ i U A M T i R C O. i I D E i R I E A R TD i 0 i T RE E g6 i H T R i N'D U G P 1 6 i EI O I CMO H OOO A k. n i

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b ANALYTICAL SUPPORT CORE-CONCRETE INTERACTION SIMULATION CE ALWR DESIGN PARAMETERS WERE USED: DECAY POWER ABOUT 38 MW. (1% OF 3800 MW WITH TIME-DEPENDENCE) CAVITY AREA 95 M2 UPWARD HEAT FLUX 1.0 MW/M2 TO WATER 32,550 KG ZIRCONIUM 112,270 KG UO2 35,000 KG STEEL 1

1 I t I ANALYTICAL SUPPORT-l CORE-CONCRETE INTERACTION SIMULATION TWO CASES SHOWN HERE WITH WATER IN THE CAVITY 1. 100% ZIRCONIUM METAL (NO IN-CORE OXIDATION) MAXIMIZING CONCRETE EROSION 2. 50% ZIRCONIUM METAL 50% ZIRCONIUM DIOXIDE RESULTS ERODED QUENCH DEPTH TIME 100% ZR 0.49 M 4200 r, 50% ZR 0.33 M 3300 S \\ ,)

\\ I ANALYTICAL SUPPORT SPREADING REGIME ANALYSIS WILL DEBRIS SPREAD TO COVER THE AVAILABLE FLOOR AREA? GEORGE GREEN (BNL) HAS PERFORMED EXPERIMENTS TO HELP QUANTIFY THE ANSWER. IN AN ALWR CASE, VOILENT DEBRIS / WATER INTERACTIONS WOULD DEFINITELY OCCUR - SO SPREADING WOULD BE EXPECTED EVEN WITH WATER PRESENT GREEN'S METHOD IS USED HERE TO ILLUSTRATE A CONSERVATIVE ANALYSIS w 2 1

n I ANALYTICAL SUPPORT h SFREADING REGIME ANALYSIS SPREADING NUMBER (M/ )1/3 g*ls N = H H*G SP F DIMENSIONLESS THICKNESS T* = (M/ )1/3 7 i M =' DEBRIS MASS P = DEBRIS DENSITY i ENTHALPY H*g = DEBRIS T = DEBRIS THICKNES3 d = WATER HEIGHT Hp = WATER ENTHALPY T* = 0.03 N -1/2 SP A L

r A T A 1 i T D L A O ,E O l i ) l E i C PNWE H L l 1 A d 7 OO1 'P N FWI H T L A E U U OI FWS B S L S M ( LN EN PW I YAA E G RHR OEO E DST CDL E L N R i

: : : 5 E

234 1 EEEEE E NNNNN R GGGGG EEEEE G .8 0 HRRRR s Q. 0 l e$OO& A. 6 1 G e. I i i i i P 2 S i l E M c I G t E 1 R 0 1 l 0 l g,. E g M IGe i E + l R i .c a A i oy 5, O 4 E i A M T I G ' A A E A O R 1 0 g ~ 6 0 a 1 0 e 0 1 0 0 .e 1 s

c \\ I ANALYTICAL SUPPORT SPREADING REGIME ANALYSIS (CONTINUED) " SHALLOW POOL" REGIME T = 0.118 M DEBRIS DEPTH < 0.24 M DEBRIS HEIGHT THEREFORE, THE DEBRIS CAN SPREAD COMPLETELY THROUGHOUT THE CAVITY EVEN WITH MUCH WATER PRESENT. NOTE THAT IF LESS DEBRIS IS ASSUMED, OR LESS WATER IS ASSUMED, THE ALWR REQUIREMENT IS STILL SATISFIED.

a n.

SUMMARY

A HEAT FLUX OF 1.0 MW/M2 CAN BE ANTICIPATED FROM CORE DEBRIS COOLED BY WATER IN CONTAINMENT. THIS HEAT FLUX RESULTS FROM DEBRIS CP.ACKING AND WATER INGRESSION THE REACTOR SCALE AND POTENTIAL FOR CONCRETE OFFGAS PROVIDE PHYSICAL MECHANISMS FOR CRACKING. INITIALLY NON-C00LABLE DEBRIS CAN EVOLVE INTO A COOLABLE STATE THROUGH REMELTING, SPREADING,

FREEZING, AND CRACKING.

A DESIGN FACTOR OF TWO HAS BEEN INCORPORATED INTO THE DESIGN RE-QUIREMENT TO PROVIDE MARGIN.

c; f I ) 'l [

SUMMARY

C0'"rINUED CHEMICAL ENERGY SOURCES ARE TRAN-SIENT AND NEED NOT BE ADDRESSES IN DESIGN. SIMULATION WITH A VALIDATED MODEL CONFIRMS THAT ADEQUACY OF THE RE-QUIREMENT. 9 i i}}