ML20055F308

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Summary of 900626 Meeting W/Mhtgr Contractors from ORNL & BNL Re Mhtgr Containment Source Term Design.List of Attendees & Handouts Encl
ML20055F308
Person / Time
Issue date: 07/12/1990
From: Williams P
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To: Joshua Wilson
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
PROJECT-672A NUDOCS 9007160257
Download: ML20055F308 (49)
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{{#Wiki_filter:. _ _ - q A' s; MEMORANDUM FOR: Jerry N. Wilson, Section Leader, ARSS/ARGlB/DRA/RES -FROM: Peter M. Williams, ARSS/ARGIB/DRA/RES.

SUBJECT:

REPORT OF MHTGR CONTRACTOR REVIEW MEETING 'l On June 26, 1990 we met with MHTGR contractors from ORNL and BNL for the p(urposes of (1) discussing the MHTGR containment source term and design,2) review for FY91, and (4} and fission product transport, (3) research ingress modeling - planning discussion of current DOE activities. is a list of meeting attendees, Enclosure 2 is information presented by the NRC staff, Enclosure 3 is the ORNL presentation, and Enclosure 4 is the BNL presentation. Highlights of the meeting are presented below. ' 1. J. N. Wilson reviewed recent-DOE activities noting a cost reduction study is underway and that the DOE program may be redirected as a consequence. A revised work plan is to be effective in October of this year and the ' cost study is to be completed by the first quarter of calendar 1991. Information permitting a decision to proceed with a first-of-a-kind plant is to be available at the end of FY92. LORNL will become the MHTGR safety center for DOE. EPRI issued a " final" evaluation of the MHTGR in February 1990; however, EPRI will also develop by July 16, 1990 an additional-review known to be undertaken by different individuals. The implications of the DOE potential redirection are difficult at this time to assess, but will be considered as much as practical in the development of the RES program for FY91. 2. P. M. Williams discussed the five major criticisms'of the developing staff report, "MHTGR Containment Function and Design Bases." These are: (1) the need to cle? ify the role of the hydrolysis source term, (2) the proper role of mechanistic safety analyses, (3) the need to understand-the talicrc mechanisms-for subspecification fuel failure, (4) the need to clarify the retention f actors given in Table 2 of the report, and (5) the-relevance of emergency planning decisions to the containment study. 3. S. J. Ball (ORNL) reported that the reactor systems code HORECA does not. yet include the balance of plant, but that this can easily be done and system interactions and operator actions can be explored. Ball will investigate Japanese programs and data as validation sources. -The MORECA code report is about 60 percent complete. 90071 57 900712 PDR DJ #sesses E 1 i @<L PDC \\

s 4 p 4. R. P. Wichner (ORNL) presented his discussion and conclusions of-plateout and liftoff phenonmena as will be discribed in a forthcoming report. The behavior of iodine, cesium, strontium, and silver.' has been reviewed with the emphasis on iodine. For dry conditions, it is estimated that iodine transport by dust will be small and that most iodine will remain chemisorbed on the steam generator because of low shear forces. Wet conditions are most certainly to be controlling. DOE's COMEDIE loop experiment may give only limited information since planned experiments have not been based on theory and modeling but are more empirical in nature. 5. P. G. Kroeger (BNL) reported on his preliminary studies with RATSAM-B and ATMOS on blowdown through the relief valve and on core temperature transients from severe water ingress using THATCH and SYMELAN. With rapid blowdown, 80 percent of the helium is released to the environment and 18 and 2 percent remain in the /eactor building and primary system respectively. After blowdown, the reactor building is essentially all helium at temperatures between 100 and 170'C. The concrete walls do not heat up appreciably. With a longer term blowdown, the cavity gas cools in about an hour and there is a net inflow to both the reactor building and primary system for a few hours followed by a net outflow for up to 100 hours. For an extreme water ingress transient, the subsequent power pulso does not appear to cause fuel damage although this conclusion is subject to confirmation of water reactivity worth and fuel thermal properties. Also, as discussed later by telephone, the case for less rapid water. entry needs to be explored to assure that the integrated power pulse does not lead to excessive fuel temperatures of longer duration and thus significant fuel damage. 6. The similarities and differences of the computer codes at ORNL and BNL to calculate transient behavior were identified and discussed. ORNL and BNL will work together and develop a proposal to establish an efficient and consistent code " package" for MHTGR licensing analysis. 7. In the remainder of FY90, ORNL will perform the in-house review of the plateout and lif toff study, complete the report on MORECA modeling and use, and integrate balance of plant modeling into MORECA. ORNL will also explore a subcontract with Dr. John Askew, formally program Director of AGR Development in the U.K., for assistance in reactor physics including the steam ingress reactivity concern. B. In the remainder of Fv90, BHL will continue the study of steam ingress transients and revieu the potential for RATSAM-B to model wet depressurization. 9. Both ORNL and BNL will dcydop proposals in conjunction with RES for FY91 that consider modest increases in availble funding and the ambiguity in the future DOE program. ORNL will continue its study of fission product transport with emphasis on wet depressurization, but defer emphasis on

l ,c s ENCLOSURE 1 MHTGR Contractors Meeting June 26, 1990 f Nane Phone Peter M. Williams NRC RES, ARGIB 492-3736 Peter G. Kroeger-BNL (516)282-2610 FTS 666-2610 Tony Hsia NRC, NRR, PSD Jim Watt NRC, NRR, PSD 492 1858 Maurizio Colagrosi NRC, RES 492-3822 Charles Ader NRC, RES, ARG1B 492-3765 Robert F. aidner ORNL (615)S74-6863 Syd Ball ORNL .(615)574-0415 FTS 624-0415 Jerry Wilson NRC, RES 492-3729 ) o I i

--m_- 3 c o oJ 1 n 3 JUL 121990 fission product holdup in the reactor building until the DOE containment i issue clarifies. Similarly, fuel failure model development including QA/QC concerns and manufacturing weaknesses are also important topics dependent on resolution of the containment issue. In code development and utilization ORNL will select from topics listed in Enclosure 3. BNL 1 will propose topics from Enclosure 4. again considering the present ] ambiguities in the DOE program. I I Peter M. Williams j Advanced Reactors and Generic Issues Branch l Division of Regulatory Applications Office of Nuclear Regulatory Research l

Enclosures:

1. List of Attendees 2. Information by NRC Staff 3. ORNL Presentation 4. BNL Presentation ] Distribution: [MEMD FOR JNW FROM PMW) subj-circ-chron *EThrom

  • WTravers, NRR -
  • ESBeckjord
  • RJohnson
  • CMcCracken, NRR i
  • TPSpeis
  • MColagrossi
  • SWeiss, NRR CJHeltemes M.El-Zeftawy, NRR SBa11, ORNL BitMorris AHsia, NRR PKroeger, BNL FACostanzi JWatt. NRR GVan Tuyle, BNL i

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  • TMurley, NRR
  1. NUDOCS~EProjectrN61V6724

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  • Cililler, NRR

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  • w/o enclosures 1

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  • SEE PREVIOUS CONCURRENCE Cffe: 'ARGIB:DRA ARGlB:DRA
<Name:

PWilliams:rg *JWilson Date: 7//J/90 7/12/90 nog 7,3 OfflCIAL RECORD COPY I h a

.,;,,,,- ~. m I 5 e1 i e 1 1 1 1 l i i ] i i ENCLOSURE 2 l RES Staff Presentations J 1 I w I e

Draft Revised MHTGR Schedule i -Complete Cost Reduction S -Decision on Whether to with GAS NPR 1 -i_icensing Review Basis for MHTGR l -Complete Baseline Design j t -Information Available for Decision on Whether to Proceed with Frst-of-a-Kind l First-of-a-Kind Application First-of-a-K~nid Combined l License [ [ r r r, r, r r j l 4 l l elt 4 -l:, 4-l: !l l-l.xl. l: ;[ ] 1990 1991 1992 1993 1994 1995 1996 Cahmdar Year i h l

4-MHTGR SAFETY ANALYSIS ORNL RESEARCN PROGRAM TOOLS FOR INDEPENDENT STUDY MORECA - HEAT TRANSFER AND FLUID FLOW, EARLIER VERSION USED FOR FORT ST. VRAIN. RECENTLY EXTENDED TO INCLUDE BALANCE OF PLANT INTERACTIONS; RITA - REACTIVITY TRANSIENT ANALYSIS - SNORT TERM TRANSIENT RESPONSES. SOME SUCCESSFUL COMPARISONS WITN FORT ST. VRAIN AND GERMAN DATA. BNL RESEARCN PROGRAM TOOLS FOR INDEPENDENT STUDY THATCH - GENERAL PURPOSE REACTOR CODE. HEAT TRANSFER, FLUID FLOW, TRANSIENT INPUTS. PASCOL - QUASI-STEADY STATE NEAT TRANSFER CODE FOR PASSIVE COOLING. ATMOS - BLOWDOWN ANALYSIS, QUASI-STATIC. RATSAM - FISSION PRODUCT LIFT-OFF, DEVELOPED BY GENERAL ATOMICS. Backup < 4 O

.~- a \\ MHTGR RESEARCH PROGRAM ACTIVITIES FOR FY 90 CODE DEVELOPMENT AND ANALYSIS ACHIEVE FULL PLANT SIMULATION WITH ORECA TO j MODEL BALANCE OF PLANT AND OPERATOR INTERACTIONS ASSESS USEFULNESS OF VALIDATION DATA FROM FORT ST. VRAIN, U.K., FRG, AND JAPAN. PERFORM VALIDATION-CALCULATIONS AS APPROPRIATE. ASSESS' MAGNITUDE AND EFFECTS OF STEAM INGRESS BY PARAMETRIC STUDIES AND IMPROVED AND' REALISTIC MODELING i DEVELOP AND USE THE RATSAM-B CODE OF GENERAL ATOMICS TO ANALYZE RAPID DEPRESSURATION ~ TRANSIENTS USE ATM05-2 CODE FOR FISSION PRODUCT TRANSPORT FOR THE LONGER TIMES ASSOCIATED WITH CORE HEATUP FISSION PRODUCT TRANSPORT REVIEW AND COMMENT IN A FORMAL REPORT ON AVAILABLE INFORMATION ON PLATE 0UT, LIFTOFF, EFFECTS OF DUST, DEPOSITION IN PATHWAYS AND l OTHER TOPICS PERTINENT TO TRANSPORT PHENOMENA. ASSESS DEVELOPING DATA AND INFORMATION FROM ~ COMEDIE AND RELATED EXPERIMENTS GENERAL ASSISTANCE REVIEW AND DISCUSS APPROACHES TO SOURCE TERM DEFINITIONS AND CONTAINMENT DESIGN BASES. ASSIST STAFF IN INDEPENDENT CONTAINMENT STUDY REVIEW AND CCMMENT ON SELECTED TECHNICAL REPORTS FROM DOE, DOE CONTRACTORS, NATIONAL LABORATORIES, FOREIGN LABORATORIES, OTHERS l [ e-l

e ', MHTGR RESEARCH PROGRAM PLANNED ACTIVITIES FOR FY 91 ~ CODE DEVELOPMENT AND ANALYSIS CONTINUE PARAMETRIC STUDIES OF STEAM INGRESS INCLUDING MULTIDIMENSIONAL TURBULENT FLOWS IN PLENUMS, MULTIDIMENSIONAL REACTOR PHYSICS VALIDATE CODES AS OPPORTUNITIES BECOME AVAILABLE STUDY EFFECTS OF STEAM IN REACTOR CAVITY ON RCCS i PERFORMANCE FISSION PRODUCT RELEASE AND TRANSPORT DEVELOP INDEPENDENT FUEL FAILURE MODEL CONTINUE STUDY OF TRANSPORT MECHANISMS WITHIN ~ PRIMARY SYSTEM, REACTOR BUILDING, AND ADD-ON CONTAINMENT / CONFINEMENT FEATURES s l AS FUNDS PERMIT L . REACTOR PHYSICS . ASSEMBLY AND REVIEW OF MATERIALS DATA 1 . STRUCTURAL AND FUEL GRAPHITES . METALS u l. I l 10-

l i 3 i MAOOR COMMENTS MHTGR CONTAINMENT STUDY i. 1. USE OF HYDROLYSIS SOURCE ' TERM Cimrity origin Remmerch Statum Basim for Treatment 1 2. BASIS FOR MECHANISTIC ANALYSIS What im DOE *m approach? Should we require mechanistic approach regardless of DOE 7 ) j 3. MECHANISM FOR SUBSPECIFICATION FUEL FAILURE What in the cmune? What im its probability of occurrence? What in the probable quantity? What in the R&D need? t 4. BASIS FOR RETENTION FACTORS Should liftoff /washoff = 0.5? Im delayed factor of O.005 representative? i I 5. EMERGENCY PLANNING CONCLUSION t 4 n

1 .c+ 4 l ~ l \\ t.*.tOSURE 3 ORNL Presentation 9 i \\ 4 l i I l t ? I \\ t 1 1 l i

7 FISSION PRODUCT PLATEOUT AND LIFTOFF IN THE MHTGR PRIMARY SYNFEM: A REVIEW 5 l R. P. Wichner l i i F USNRC Pro,entation: June 26,1990 l t OUTLINE & CONCLUSIONS -e---. ' = + w a .n...

~~ OUTLINE ^ i RELEASE FROM FAILED FUEL i e AM AND SERVICE FAILURES

  • hor nUP DECAY CHEMICAL CHARACTERISTICS l

1

  • "PLATEOUT' MECHANISMS l
  • BEHAVIOR OF I, Cs, Sr, Ag
  • PS DUST
  • REACFOR DATA
  • TYPES OF DUST j
  • QUANTr1T t
  • PLATEOUT MODELS
  • LIFTOFF MODELS l

i i .i 5 i .~

i CONCLU5 ION 1 CURRENT PLATEOUT AND LIFTOFF MODELS LACK SUFFICIENT MECHANISTIC BASIS FOR CONFIDENT APPLICATION. THE CHEMICAL AND PHYSICAL BASES OF THESE MODELS ARE NOT .i AVAILABLE OR HAVE NOT BEEN APPLIED. REQUIREMENTS e APPLY KNOWN CHEMISTRY OF J, Cs, Sr, and Ag TO MODELS

  • APPLY KNOWN REACTOR DUST DATA
  • UPGRADE DUST LIFTOFF MODEL I

i ~ =

i CONCLUSION 3 t i l RADIOACTIVITY RELEASE FROM PS DUETO DRY DEPRESSURIZATION l l IS LIKELY TO BE EXTREMELY LOW: FOR IODINE, < 0.01% OF TOTAL EX-i FUEL INVENTORY OF ~45 Ci i l REASONS

  • LOW DEGREE OF CHEMICAL DESORFFION
  • LOW SORPTIVITY OF DUST FOR IODINE

[

  • SEQUESTERING OF CESIUM IN OXIDE FILMS j
  • STRONTIUM EXISTS AS REFRACTORY SOLIDS SrC or SrO 2
  • SILVER EXISTS AS' SOLID
  • LOW SHEARS IN SG (THE PRINCIPAL DUST REPOSITORY) j T

i l

-y

k IODINE LIFTOFF

SUMMARY

(SPECULATIVE) ~ j LIFTOFF j PS OIEMICAL DUST ACITVITY DESORFFION LIFTOFP TOTAL (G) (G) (G) (G) CIRCULNFING 1.2x10d 1.2x10* PLATED SORBED ON STEEL 43.2 4.8x10' 4.8x10' SORHED ON GRAPHITE .9 4.6x10' 4.6x10' SORBED ON DUST 1.3 8.7x10' 3.8x10' 125x10d TOTAL 45.4 2.2x193 "FROM SG TUBING ONLY i

~-

.. - ~.

y ~ ~ CONCLUSION 5 O/M ESTIMATES SHOULD BE PERFORMED FOR Cs, Sr, and Ag (as done for I). RECOMMENDED APPROACH

  • EVALUATE COMP 151TIION BETWEEN PERMANENT SURFACES AND DUST FOR REFRACTORY FP FQRMS - Cs,MO,, SrO, SrC, and Ag".
  • EVALUATE LIFTOFF FROM CORE SURFACES. (The SG seems protected by the circulator.)'

--M.. -._a

CONCLUSION 7 THERE IS SIGNIFICANT UNCERTAINTY IN THE AMOUNT OF FP'S IN PS. REASONS

  • CONFIDENCE IN FUEL FAILURE FRACTION, AS - MANUFACTURED AND IN-SERVICE FAILURES
  • FOR IODINES, PROOF OF LONG HOLDUP IN FAILED FUEL AND CONTAMINATION, WHICH RESULTS IN DECAY OF MAJOR AMOUNT
  • FOR SILVER, UNCERTAINTY IN BIRTHRATE OF Ag110m DUE TO LEAKAGE OF Ag109 FROM FUEL
  • FOR STRONTIUM, HOLDUP IN UCO OR RELEASE AS Kr PRECURSOR
  • FOR $1LVER AND CESIUM, UNCERTAIN LEAKAGE FROM INTACF FUEL IN HT ZONES

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l June 25,1990 SJBall i DRAFT - Suggested NRC/RES Research Topics for MHTGR Summary of Topical Areas: 1. Improvements of accident sequence code models: - Vessel to RCCS heat transfer (effects of particulates and water vapor in the interspace, detection of degradation in effective emissivity, plateout of fps during accidents,[ emissivity, heat source], evaluation of in situ integral tests), development of simplified models and confirmatory experiments. t - Water ingress (RELAP-type) modeling, with parametric studies (more realistic estimates of water vapor ingress rates to the core, considering the effects of the' circulator and the { flapper valve).. - Fuel failure model refinement for low release range, with parametric studies. ~ - Incorporation of available reactor data for code validation (FSV - via NPR, and THTR), and development of draft NRC policy guidelines on vaUdation experiments (integral, separate-effects, QA,...) - Containment / confinement building modeling, validation expts. l l 1 r -+ ,x-

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5. -

Other issues: - Develop acceptance criteria for prototype test. - Maintain, beef up foreign liaisons. - Physics data evaluation (John Askew proposal). 9 e 9 e 3

RECENT REPORTS "Modeling and Performance of the MHTGR Reactor Cavity Cooling System" -(J. C. Conklin, NUREG/CR 5514, ORNI/TM 11451) f "MHTGR Short Term Thermal Response to Flow and Reactivity Transients" (J. C. Cleveland, Nuclear Technology paper and ORNL TM report) MORECA code report (SNPR): ORNIJTM writeup (60%) Detailed documentation has been put on a DBMS " validation" data from THTR " Integrated System" code review g "' Lee Smith) l l l i

i MORECA INTERFACE ALLOWS OPERATOR / ANALYST ] INTERACTIONS'WITH MHTUR ACCIDENT SEQUENCES l (UP TO 1000 X REAL TIME) I sustem Narr CaeE IEME a StG sets ( 13.5em ) I t 1 g (eer) 1-aus (F): v.sw (F) = 929 45-18 (S/S) 1.26 Eusts (F): 1 g T-AIn-eur (F) z 23e T-498 (F) = 1455 I e (ser) = 2.48 -{ P (Pssa) 379 l 4 4> i 1.597 -0.373 .i FtSW (S/S) { tenE ACCS TESSEL TW-eWI (F) a 177 I I-eWT (F) = 666 SCS i Saselc5Ien: Step Encore Inefore g5E5 SCS: etCS: SE NES35 BIN: [ SPEED-E:] 300. @ Q 23. g 5. t... E f1 905. l/+ [ 413 300 { g e . m. .___._.____....___.__._.-.__.__.__.._6, _ _ _ _ _ _ e Ami N

yzup: 9 ) i i -{ s u I f ( -1 i } f ENCLOSURE 4 f f BNL Presentation I i I I \\

i - i I SAFETY RELIEF VALVE BLOWDOWN (PRELIMINARY) FOR f MHTGR I Peter G. Kroeger Department of Nuclear Energy Brookhaven National Laboratory 3 June 26,1990 b o t

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l ^ Relief Valve Blowdown l Leak Flow and Temperature i 60 1000 LEGEND Row - I Temperature 50-u.i e e 40-j l I \\ LL l g 30 \\ -500 l 1 L 5 \\ / / / g / q 20-N,_ g IL g 1 g a e i p ~ il 10 : e k. I a 0 . 0 l E O 200 400 000 Time (s) i s - - - -r ~ 2 m

y ya y e e x asn o 4 -<a + w a s s - a e --+a4 4~=*.14 -o-.-~~*--+-a-ano--aa-+--+++ sam-ss-a- > n mennms4.-+-*--~*~ee-,e ---.s.-a--a o q i ,4 e to 4 6 j 'i t A T M O S-2 RESULTS e for ? BLOWDOWN PHASE 1 P h 5 I (i

w a ~. 0 0 ) 6 n w w o dws be 0 r 0 u a Bs 4) (s s n e e vr d P e Vs m a m f G i T N, w ey 0 feit y y ,0 B B 2 t t R v D v v a N o oRR yC EC C r r r e e t k G e w p r i e t t k o P U__ f S S L ~ a S i;; \\, 0" ,1 ~ 0 0 0 0 0 0 0 0 0 0 0 0 8 6 4 2 1 ^ & v OC -LlO g g .e E a it3 (i ?i!lt, ! i la ! i ~ ~ ,r i, .i ' :

+ e-v Safety Relief Vdve Blowdown Heat From Gas to RB Walls 1.0 .LEGB4D 4 ' St'G Covty- _ I.k Side Cavty I'- Lower RB - \\ ___Upp,[Rp,_, i \\ n S n 3 2 0.5 - j /\\ \\ o l, N. y s g c . \\- \\ 6 c ~ t N N l l l s l -l w 'O h ____ !i f = 0.0 -- i 8 0' .200' 400 600 -Time. (s) = e 5

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Covity Gos Pressures -

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. m x a 1:s .n ll, ) ' J ) i. RESULTS FOR BLOWDOWN PHASE t After T4 lowdown, RB. Essentially all Helium 100-170 C Concrete Walls do not Heat Up Appreciably (Smaller Metal Structures'Might) ~ lt) Approximately 80% of Coolant Released to m Environment; 18% in RB; 2% in Primary Loop. b, c,... o J' )

= =, Safety Relief Valve Blowdown RV Gas Temperatures 400 300-LEGEt0 Herum from Reactor i Air fr Reactr Cavty O St Generator 200 - j O '[ !? 10 0 - g ~ 3 4 ( = ^0 O -50 10 0 s Time (hr) I e =-

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a. t v Safety Relief Valve Blowdown ~ ~ 4 RB Wdi Temperatures 45. i LEGEND St G Covty' Side Cavty Lower RB may mamme o O E O i 40-i s b ~ t pl 1 E l = 35 I-0 5 10 [_ Time. (hr) ~~ .rN AsNL'A PmM'" "t u. 2 x W h S .7'L -~i "- Ne u'e -' ^

+ ..: y RESULTS FOR LONG' TERM POST BLOWDOWN TRANSIEFT . Cavity Gas Cools Rapidly (= 1 hr) Net Inflow to RB for Approximately 3 hr Net Inflow to RV for Approximately 4 hr, k Net Outflow to Environment from 5 to 100 hr Dominated by Gas Expansion in Reactor Cavity. Approximately 4% of Post Blowdown Primary Loop Inventory Released to RB Between 5 and 70 hr-i'

c 'e - 4 y 49s t i BOUNDING CORE TEMPERATURE TRANSIENTS FOR-SEVERE WATER INGRESS SCENARIOS P. G. Kroeger June 26,1990 j l

'~: .. : s. . 's COO LANT HO LE 0.50 DI A (6) COOLANT HDLE /).625 Ol A (102). 1 1 "' d ' FUEL HOLE -(: 4 d 0.50 DIA (210). , '... DOWEL PIN l e i. 14,172 F } D' c..

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1 w p.,, ci ; + c 4 \\ Li? m BASE CASE .i f, ~ O f j I. 1 Core from 100% He to 100%. H O in Ramp Time of 10s j 2 n .No Scram 1 3+ u -em y g n No HTS Shutdown t +

Best Estimate Water. Worth and Fuel Thermal Properties y

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z,z g -+, --w._ 2 L.,' ' ' I . ' * ~ y ~. 10 s; Water Ingress Ramp, Best Estimate -Hot Fraction of Fuel 4 l l 0.7 LEGEND i > 1200 *C l l 0.6 - > 1400 *C l > 1600 *C > 1806 *C 0.5- ~ n 2 2 c 0.4 - ~ E 'O .~ c s o 1 E' 3 Lt-O.3 - ~ i ii I 02-5 R k O.1 - (N-f .N,, )lO.,. .,T.,,....,... 6 N a 0.0 ' E 0. 20 40 60-80 10 0' i Time (s) f L

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l __, i' PARAMETRIC VARIATIONS a 1 i f 5 t L Ramp Time g ,-li i m Water Worth 4 q Fuel Thermal Conductivity- ,t I . l l. , r, f f i! [ 1 Fuel Thermal Capacitance g ll'., I .l ~ j ? l' k i E l !I -

l l u u 7 02 p m ..51 a R r ). s 'e s sw eo /1 rP ,0 g e. 1 nn m o I i r s i s T ei t F a W. ,5 s 1 1 .O' ) o 0 0 0 0 - 0 8 6 4 2 oo_ N O ". ie ib ! }, E i - E i, .E iI 9 S'

.x " f,= a_ }} ^~ i 20 s Water hgress Romp a Core Temperatures. 2000 -l LEGEND ~' -3 t Fuel max Fuel ovg Graphi_te max i ~ ___Gmphi_te_ov3,,, 1500 - O 'l' m g o og t, g c ~ -- - __.- 1000.- / 5 ) I 1, . f,. ' ae*# ~ 500~ .n n 's' s' E O 120 40.. 60__ q 80-10 0-Time (s) y = l

= =. I. g* .5 w ~ EFFECT OF WATER REACTIVITY WORTII ON BOUNDING-SEVERE WATER INGRESS TRANSIENTS ~ Maximum Exposure Maximum Fraction Time for any Fuel-Case Peak Fuel of Fuel Exceeding To Temperatures Peak Power. Above Temperature 1600 C 1800 C 1600 C. 1800'C .( C) (%) (%) -(s) -(s). P/Po Base Case

1668 0.9 7

11.3 Water Worth-p/p, = 0.8 1507 p/po = 1.2 1826' 5.5 0.3 14 3 13.2 p/po = 1.4 1979 12.1 3.0 23 6 14.9 e ~ ~ -w w.. I

. c. 2 t 4 -- i h . OVERALL CONCLUSIONS I s.:.a. Even Extreme-Bounding. Water Ingress Transient does~ not! Appear to Cause Fuel Failures I Water Worth and Fuel Thermal Properties must be i L . Established and Controlled with Great Confidence W h k c l t ..4 3

i

f x

+ THATCH Basic Capabilities for Core Heatup Transients x 1 Arbitrary User Sslected Nodalization + (2-Dimensional R-Z Geometry) Transient Conduction Radiative / Conductive Interface Resistances

2-d Radiation in Plena (User Option)'
Water or Air-Cooled RCCS Capability.

Natural Convection Heat Transfer in Plena Reactor Coolant Flow and Convective. Heat Transfer + Gas Expansion in RB and RV during Long Term Transients 1

. 3= p ' g'. -i .. :. w ,.6 j". 1 p-RATSAM-B c Capabilities y 1 o i II .i q q Rapid Blowdown of Primary Loop i Shear Forces.for Lift-Off 1 i '1 [ MHTGR Control Logic'

j 1

s Steam Generator Model u i ~ f u ? 3 %d W is 3

},.. ' SUGGESTED FY 91 EFFORTS LBROOKHAVEN NATIONAL LABORATORY Fission Product Transport Apply.and Extend Current l Water Ingress-Analyses Multi-Dimensional Plena Turbulent Flow and Heat Transfer ~ Multi-Dimensional-Kinetics to THATCH /SYMELAN

qme ^ e 1 g, %. 4 y t ? \\' ' MULTI-DIMENSIONAL PLENA TURBULENT FLOW. . AND HEAT TRANSFER- \\ Upper Plenum Heat Transfer in Pressurized Core

Heatup Scenarios H

Mixing of Helium and Water Vapor during Water 1 Ingress Scenarios-LHot Streaking and Thermal Cycling Loads in Lower. Plenum 'u .l ~l i I . e.,:,. .}}

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