ML20067D615

Summarizes 821216 Meeting W/Nrc,Bnl,Anl,Westinghouse,Eg&G & Barthold & Assoc Re Shutdown Heat Removal.Viewgraphs Encl

ML20067D615
Person / Time
Site:	Clinch River
Issue date:	12/20/1982
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:S:82:149, NUDOCS 8212210045
Download: ML20067D615 (40)

Text

__ - -_ _ __ __ - _-___-__- ___ - _

Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:149 DEC 2 0 E82 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Check:

MEETING

SUMMARY

FOR THE SHUTDOWN HEAT REMOVAL MEETING, DECEMBER 16, 1982 The purpose of this letter is to sumarize the resolution of items discussed between the Nuclear Regulatory Comission and the Clinch River Breeder Reactor Plant project on December 16, 1982. is the agenda used for the meeting, Enclosure 2 is the agree-ments and commitments from the meeting with item resolution status, and is the list of attendees.

Any questions regarding the information provided or further activities can be addressed to A. Meller (FTS 626-6355) or D. Florek (FTS 626-6188) of the Project Office Oak Ridge staff.

Sincerely, ngviblo Jo n R. Longeneg er Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy 3 Enclosures no l

o cc: Service List V

g Standard Distribution Licensing Distribution j

8212210045 821220 PDR ADOCK 05000537 PDR

BRIEFING ON CRBRP SHUTDOWN HEAT REMOVAL CONCERNS FOR.THE NUCLEAR REGULATORY COMMISSION CRBRP PROGRAM OFFICE BETHESDA, MARYLAND DECEMBER 16,1982 AGENDA

~

DIRECT HEAT REMOVAL SERVICE................................... R. E. H l

SPECIFIC CHANGES TO INCORPORATE f;

o l

SINGLE FAILURE CAPABILITY

[

RESULTS OF LATEST SINGLE FAILURE ANALYSIS o

DETAILS OF BASE-CASE F-2 ANALYSIS o

NATURAL CIRCULATION................... R. R. LOWRIE/R. L. MARKLEY/R

~

RESULTS OF LONG TERM NC ANALYSIS UNCERTAINTY FACTOR APPLICATION TO NC (4.4-9(4.,4-7) i o

l o

NATURAL CIRCULATION TESTING (5.3-19)(4.4 /)

DISCUSSION OF ACRS QUESTIONS.................................. R.

o s

DHRS O'/CRFLOW LINE o

PWST CAPABILITY o

CONDENSATE STORAGE TANK DETAILS (5.3-17) o PRIMARY CHECK VALVE DISCUSSION OF ADDITIONAL OPEN ITEMS..........................

o CLARIFICATION OF AFWS/SWRPRS (5.3-13 & 5.5-1) o PACC ANALYSIS (5.3-14) o AFWS REL! ABILITY EVALUATION (5.3-18) o LOSS OF BULK AC POWER (15.1-7) o PIPE E* LEAK ANALYSIS (15.3-2) o DISCUS!;lON OF REMAINING............................. BROOKHAVEN NATIONAL LAB '

760 SEFilES QUESTIONS R. E. HOTTEL

SUMMARY

(su- - - -.

- - - ^ - ^ ^ - - - -

Inclosure 2 Shutdown Heat Removal Meeting December 16, 1982 Agreements and Commitments 1.

Direct Heat Removal Service The specific changes to incorporate single f ailure capability, the results of latest single f ailure analysis, and the details of the base case F-2 analysis were discussed. (see attached)

The peak temperatures for the single failure assessment slightly exceed (by 17'F) the structural peak temperature This is judged acceptable analyzed for the F-2 transient.

based on a scoping analysis since the capability of the primary boundary to maintain structural integrity, based on creep rupture considerations, at temperatures of approximately 12000F is approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

Asilent The updated (performance) design case and the single failure evaluation case will be placed in the PSAR in addition to the description currently in the PSAR.concerning the structural design case.

This item is resolved for the SER.

2.

Natural Circulation Long Term Analyses The results of long term natural circul ation analyst s were presented (see attached).

This item is considered resolved for the SER.

Oncertainity Factor Application to Natural Circulation - The uncertainty factors applied to natural circulation were discussed (see attached).

This item is considered resolved for the SER.

Natural Circulation Testing - The project commits to conduct a whole plant test program of CRBR natural circulation and The DHRS perf ormance during pl ant startup testing.

is to confirm the objective of the proposed test program thermal-hydraulle computer codes which have been used to predict both the plant natural circulation behavior as well as the plant thermal-hydraulic behavior under DHRS flow conditions.

The natural circulation and DHRS test program will provide Information tot 1.

Ensure adequate prediction of nat'l cir. capability (transttlon as well as steady state flow and n

...).

temperatures) and DHR$ Decay Heat Removal capability.

2.

Ensure adequate prediction of the effect on temp. and flow caused by variations in the heat sinks available (e.g. venting, PACC's, 2-HTS loops).

This item is resolved for the SER.

3.

Dicussion of ACRS Questions The DHRS overflow line, PWST capability, condensate storage tank details, and primary check valve were discussed (see attached).

The Information in the PSAR and meeting discussion was adequate, this item is considered resolved.

4.

Discussion of Additional Open items Cl arif ication of AFWS/SWRPRS - The discussion of the operation of the Isolation valves under AFWS and SWRPRS Initiation resolved this item (see attached).

PACC Analysis - The meeting discussion and recent PSAR amendment resolved this item (see attached).

NRC is reviewing the PSAR amendment.

AFWS Reliability Evaluation - The meeting discussion and PSAR information resolved this item (see attached).

Loss of Bulk AC Power - The meeting discussion resolved this item (see attached).

Pipe Break Analysts (see attached)

As11gn: The DEMO code output and assumptions used for het leg pipe break analysis will be provi ded to NRC.

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The Reliability program, discussed in Appendix C. will provide Y f al verification that SGAMRS removes residual heat following a reactor shut-vgRg g

down with a high level of reliability. Hence it is judged that only

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• ' q steam and feedwater trains backed by 5GAHR5 are necessary to safely and 3

n11 ably remove residual heat followin shutdown from three loop. full power operation. To enhance the relia ility of decay heat removal. the o

DMRS provides a fourth redundant heat removal path and heat sink. The 4

5 1spect on overall shutdown heat renoval reliability by inclusion of the QhNh 41 DHR$ is being determined by the Reliability Pro ran described in Appendix d

C.

The DHR$ provides this supplementary capabi ity by satisfying the 5

of 4 }d following objectives:

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Function to remove reactor decay heat following reactor shutdown g

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lI transfer through the IHK's at the time of reactor trip. Oncra.

tion of three primary pump pony motors and maximum reactor decay 8

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9' power is assumed. Under these conditions, the DHR$ is desioned K

f g-to provide sufficient cooling to ensure primary coolant boundary integrity and prevent loss of in-place coolable geometry of the 5

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gW k To meet this objective. DHR$ components will be sized such that.

l g E q) g under these conditions. the average bulk o imary sodium temera-ture will be limited to soproximately 1140 F.

Capability will g

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be provided to permit remote manual initiation of DHR5 frors the

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Control Room. The overflow and makeup circuit and the spent

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fuel cooling system will be able to be cross connected in a y

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when DHR$ is removing full capacity heat oad.

Qg mannar which permits both EYST NaK coolin trains to be used D

9M Accoanodate the thermal transients resulting from normal. upset.

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emergency and faulted plant events in which continued performance

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Accomodate floods (Section 3.4). tornadoes (Section 3.3). missiles r.

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(Section 3.5) and earthquakes (Section 3.7), in which continued E

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performance of its function is not impaired.

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Function in a manner which will not significantly reduce the I

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reliability and availability of the EYST heat removal chain.

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designed to remove concurrently the heat generated by the spent

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fuel and the reactor decay heat.

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R.4 c u o Q Amend. 41 5.6-20 Oct. 1977 I

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TABLE 5.6-10 IHR$VhlVECLASSIFICATION 26 NORMALLY 0PERATING LCCATION VALVE NO.

ACTIVE / INACTIVE OPEN/ CLOSED MODE Makeup Pump

-t-life Inactive Open Isolation Suction

-f.- 131-Inactive Open Isolation

• 26

~ + hit 114,133 Inactive Closed Isolation 1

Overflow VSL M IF3

) active -(f)-

Open Isolation 46l26 Drain / Sample Return Makeup Pump

+ 10 4-active Open Isolation 261 Dicharges t 13 8 tive

Open, Isolation

'9-101-Active Ope Isolation 4r-187 Kctive Ope Flow Control

+ t03,10 7 Ac'tive

. Closed +

Isolation Overflow Ht.

26 Exch.

Sodium Out.

9 ff8 Inactive Open Isolation 9 If/

Inactive Open Check 46l te149 Inactive-(+)

Open Iso.lation Overflow Ht..

Exch.

21

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to Airblast

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Closed, Isolation (hot leg)

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The position indictated is the valve position during normal plant operation.

Valves with + are those which are manually opened, or 26l closed, prior to initiation of D'RS operating mode.

All active valves are remote -manual ppcrable from uw coritrol room.

Valve numbers are those shown on Figure 5.1-7.

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TABLE 5.6-10 (Continued)

Page 2 DHRS VALVE CLAS$1FICAT10N NORMALLY*

OPERATING 1DCAT'lDN VALVE ND_**

ACTIVF11NAETIVF OPEN/ CLOSED MODF 364,,414 446, g

NaK Drain 20, 7,

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'88 4 Cover Gas

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Isolhtion Sodium Vent Inactive Closed Isolation (Typ.)

f Sodlum Sample -M-1408 Inactive Open isolation The position Indicated Is the valve position during normal plan operation.

Valves with + are those which are manually opened or closed prior to initiation of DHRS operating mode. All active valves are remote-manual operable from the control room.

Value numbers are those shown on Figure 5.1-7.

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5.6.2.1.2 Applicable Code Criteria and Cases The components of the DHRS shall be designed, fabricated erected.

constructed, tested and inspected to the standards of Section III,of the

h. ASME Code,1974 edition through the summer 1974 Addenda, Class 1 o g!j

=.. :,,T A ;.eu.

Applicable code cases will be used to supplement the design analysis required by the ASME Code.

5.6.2.1.3 Surveillance Requirements The need for surveillance of the DHRS piping and components will 1

26. be determined as the system design progresses and as the need to tronitor austenitic stainless steel is determined by ongoing programs.

If a requite-ment is identified, a surveillance program will be designed in accordance with the philosophy of 10 CFR 50, Appendix H.

5.6.2.1.4 Material Considerations Hiah Temperature Design Criteria High temperature components in the DHRS will be analyzed in -

26l accordance with the requirements specified in ASME Boiler and Pressure Vessel Code,Section III, as supplemented by the applicable code cases and

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RDT standards.

Material Specifications Stainless steel materials which satisfy the requirements of th=

ASME Code will be specified for use in the DHRS system, as noted in Table

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26 5.6-13.

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.6.2.1.5 Leak Detection Requirements The DHRS will be monitored for sodium and NaK 1eaks and leak 26 indication will be provided in the Control Room by the leak detection system described in Section 7.5.5.

5.6.2.1.6 Instrumentation Requirements DHRS is remote manually activated and controlled from the Control Room.

Instrumentation required to monitored the condition of the DHRS consistsofthermocouplesontheEYSTsodiumcutletlines(3 loops)and level indicators in the EVST and the Reactor Vessel (RV). These instruments confirm that the sodium levels in the RV remain above the loop outlet nozzles 50 46

'and that temperatures remain below design limits. Other DHR5 diagnostic instrumentation is not essential for DHRS operation as the pumps and air blast heat exchanger are being operated at maximum design rates. When the reactor decay heat load has dro ed sufficiently, the cooling capacity of the system may manually be reduced lowering flow-rates or fan speed, or by shutting down one one of the EV T cooling trains.

26 Amend. 50 June 1979 5.6-21

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A REDUNDANT VALVE AT THE OHX INLET PROVIDES THE CAPABILITY TO WITHSTAND A SINGLE ACTIVE FAILURE WITHIN THE DHRS AIR BLAST HEAT m

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THE INPUTS FOR THE UPDATED DHRS SINGLE FAILURE APdALYSIS INCLUDE.

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• TWO PHTS LOOPS OPERATING NO IHTS HEAT CAPACITY NO SGS HEAT CAPACITY ONE Na MAKEUP PUMP AT 600 GPM ONE NaK PUMP AT 600 GPM ONE AIR BLAST HEAT EXCHANGER CONSERVATIVE DECAY HEAT USED NO EVST HEAT LOAD HEAT LOSSES THROUGH INSULATION INCLUDED T

V

UPDATED DHRS SINGLE FAILURE ANALYSIS AVCRAGE HOT LEG TEMPERATURE ( F) 1140 t

1120

' ;. l 1100 I

1080 1060 1040 1020 1000 O

4 8

12 16 20 24 28 32 44 48 TIME - HOURS i

2-1 Long Term Natural Circulation Event (Two Hour Station Blackout) i The staff requested natural circulation analyses extendtng beyond the 600 second transient time cases previously submitted by the The objective of the extended analyses was to applicant.

demonstrate continuing stable natural sodlum circulation and acceptable transttlon in heat sink configuration from steam venting to PACC's only operation.

The f ollowing series of attachments provides the long term results of natural circulation.

The curves demonstrate that long term natural circuletion provides adequate core cooling and continually removes decay heat from the core.

The analytical assumptions and explanation of these curves follows:

The CRBRP-ARD-0308 analysis is based on maxtmum decay heats and conservative heat sink capability.

Theref ore it Is appropriate for use in evaluating peak core temperature as well as the transition f rom f orced circul ation to natural circulation.

The progress of any long term natural circulation decay heat removal event is controlled by the heat balance at the water-sodium Interface (steam generator system).

The heat inputs and the are the monotonically decreasing reactor decay power sensible heat from the sodium systems.

The heat sinks include the natural draf t PACC's, the turbine driven auxillary feed pump and the SGAHRS vents.

The long term natural circul ation analysis was perf ormed to address the transition from steam venting as the primary heat sink to the ACC's as the only heat sink that is required f or plant cooling.

This analysis assumed 1) nominal decay heat to assure that the transition to the PACC's as the sole heat sink would occur within the analysis run time and 2) a conservative HTS pump coastdown of 100 sec which is less than both water and sodl um testi ng has demonstr ated.

The PACC minimum capability is 4/4 Mu/ loop.

As shown in the following curves, inherent behavior of the present control system assures safety for well over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of station blackout.

Design optimization by the OL stage will enhance these conclustons.

The time scale affords ample l

opportunity for operator control to improve response and Instrumentation will support any required decisions.

Tnroe redundant HTS loops further assure safety even if failure occurs in any one loop.

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J COOLANT HCF = POWER UNCERTAINTIES + FLOW UNCERTAINTIES AT LOW FLOW / POWER

-i 1.

POWER UNCERTAINTIES

$ FULL POWER UNCERTAINTIES AT 3a PLUS 3a DECAY HEAT UNCERTAINTY II.

FLOW UNCERTAINTIES

.t INLET FLOW MALDISTRIBUTION

-- Fult FLOW UNCERTAINTIES PluS 30% UNCERTAINTY ON aP APPLIED 1

9 FLOW DISTRIBUTION CALCULATIONAL UNCERTAINTY l

-- Full FLOW UNCERTAINTIES USED PLUS FULL POWER / FLOW PEAK TO AVERAGE AT CONSERVATIVELY USED AT LOW FLOWS (INTRA-ASSEMBLY FLOW AND HEAT REDISTRIBUTION) 0 SuBCHANNEL FLOW AREA

-- NO CHANGE AT LOW FLOW 8, COOLANT PROPERTIES

-- NO CHANGE AT LOW FLOW I

DATA AND ANALYSES SUBSTANTIATE SIGFIFICANT CONSERVATISM IN NOT TAK}. CREDIT FOR INTRA-AND INTER-ASSEMBr_Y FLOW AND HEAT REDISTRIBUTION, I.E., SELF-COMPENSATING BUOYANCY PLuS HEAT TRANSFER EFFECTS PJ b

'1

SUMMARY

UNCERTAINTY FACTORS APPLICABLE TO NATURAL CIRCULATION PRFDICTIONS. O CRBRP CORE NATURAL CIRCULATION ANALYSES VERY CONSERVATIVE USING 100% 1 ! HCFS PLUS: '[ -- EDI TAKING CREDIT FOR INTER-AND INTRA-ASSEMBLY FLOW AND HEAT ..f. REDISTRIBUTION -- USING LARGE UNCERTAINTY FOR CORE AP 8 MANY CONSERVATIVE, WORST CASE ASSUMPTIONS USED IN ANALYSES - STILL LARGE MARGIN-TO-BOILING S CONSERVATISM OF APPROACH FOR CALCULATING MAXIMUM CORE IEMPERATURES EXEMPLIFIED BY COMPARISONS TO PROTOTYPIC FFTF NATURAL CIRCULATION DATA O HETEROGENEOUS CORE CONFIGURATION AND LONG TERM OPERATION EFFECTS ENVELOPED BY ANALYSES IN CRBRP-ARD-0303 FSAR WILL INCLUDE REFINED MODELING ON INTER-AND INTRA-ASSEMBLY FL 0 AND HEAT REDISTRIBUTION USING A VERIFIED SYSTEM OF THREE COMPUTE (DEM0, COBRA-WC AND FORE-2M) N ACCEPTANCE TEST PHASE EXPERIMENTS WILL BE PERFORMED FOR FINAL CON I 8 TION OF CRBRP NATURAL CIRCULATION CAPABILITY d

1 l l MAJOR ASSUMPTIONS USED IN NATURAL CIRCULATION HOT R0D ANALYSIS l l o O CONSERVATIVE PLANT THDV +20*F INITIAL CONDITIONS (E.G., 750*F REACTOR IhLET) l 0 WORST CASE DOPPLER COEFFICIENT INCLUDING UNCERTAINTIES ( 30%) O MINIMUM CONTROL ROD SHUTDOWN WORTH (ONE STUCK ROD) il ' -7 S CONSERVATIVE FLOW COASTDOWN OF PRIMARY PUMPS 9 3a HOT CHANNEL / SPOT FACTORS (ID TEMPERATURES) i 0 MAXIMUM CORE PRESSURE DROP 0 HIGHEST POWER AND TEMPERATURE HOT RODS AT WORST IIME IN LIFE 8 WORST END OF UNCERTAINTY RANGE USED FOR FUEL PROPERTIES AND FUEL / CLAD 6AP CONDUCTANCE FOR BOTH POWER AND TEMPERATURE CALCULATIONS 0 MAXIMUM DECAY HEAT LOADS INCLUDING 3a UNCERTAINTIES AND IIME IN LIFE EFFECTS 0 NO CREDIT IAKEN FOR INTER-AND INTRA-ASSEMBLY FLOW AND HEAT REDISTRIBUTION 0 NEGATIVE REACTIVITY FEEDBACKS NEGLECTED (E.G., CORE RADIAL EXPANSION, BOWING, AXIAL EXPANSION OF FUEL AND CLADDING) 6 ALL ABOVE ASSUMED TO OCCUR SIMULTANEOUSLY N 1

DHRS OVERFLOW LINE OVERFLOW LINE CAPACITY INCLUDES ALL DHRS OPERATING CONDITIONS. FULL SCALE MODEL TEST CONFlRMS CAPABILITY ~ OF OVERFLOW LINE 'i' ~ DHRS OPERATING PARAMETERS AND OVERFLOW LINE DESIGN PREVENT FREEZING DURING DHRS OPERATION. O I t s

. PROTECTED WATER STORAGE TANK CAPACITY

• PWST INVENTORY ACCOMMODATES ALL DESIGN BASE EVENTS
• WITH PACC OPERATING ON AIR SIDE NATURAL CIRCULATION (BEYOND DESIGN BASIS?, PWST INVENTORY ACCOMMODATES A THIRTY DAY i

MISSION

• WITH DECAY HEAT REMOVAL PERFORMED ONLY BY VENTING CBEYOND DESIGN BASED, PWST INVENTORY ACCOMMODATES A MINIMUM OF NINE HOURS OF HEAT REMOVAL f

3 0

i CONDENSATE STORAGE TANK .c AVAILABLE WATER SUPPLY TO PWST AND TO AFW PUM.P SUCTION (250,000 GALLON CAPACITYD GRAVITY DRAIN TO BOTH PWST AND AFW PUMPS ~ SAFETY-RELATED ISOLATION VALVES NON-SEISMIC CATEGORY TANK i NOT CONSIDERED IN SAFETY ANALYSIS I 9 Y o

PRIMARY CHECK VALVE SCALE-UP OF FFTF CHECK VALVE WITH DASHPOT ADDED MINIMAL PRESSURE DROP UNDER NATURAL !1 CIRCULATION FLOW (~0.03 PSIGD

(l 13 1/2 FREEHANGING ANGLE ON SEAT DASHPOT HAS BEEN PERFORMANCE TESTED AND THERMAL-CYCLE TESTED FAILURE MODES AND EFFECTS ANALYSIS PERFORMED FAILURE MODE EVALUATION AND PROBABILISTIC EVALUATIONS PERFORMED Y

max

CLARIFICATION OF AFWS/SWRPRS USE ~ RESPONSE PROVIDED IN NRC QUESTION RESPONSES CS 421.28 AND CS 760.115 I SWRPRS ACTUATION INVOLVES ONLY AFW ISOLATION VALVES IN THE AFFECTED LOOP DECAY HEAT REMOVAL CONTINUES VIA REMAINING TWO HEAT TRANSPORT LOOPS .Js. _L_ A 12 82 3152 13

PACC SYSTEM ANALYSIS PSAR SECTION 5.6.1.2.3 AND .i ~ 5.~6.1.3.2 UPDATED TO PROVIDE DETAILS FOR SYSTEM DESIGN

i AND ANALYSIS

'j PROJECT IS COMMITTED TO VERIFICATION TESTING + N 82 3152 14

~ AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION / ADDRESSED IN APPENDIX H, SECTION ll.E.1.1

.- l fAMENDMENT 66?

j.;, l EVENT TREE AND FAULT TREE ANALYSIS DONE i UNDER DHR BY KEY SYSTEM REVIEW FMEA AND CCFA BEING DONE UNDER QUALITATIVE RELIABILITY PROGRAM QUANTITATIVE ANALYSIS BEING DONE USING FAILURE STATE MODEL PSAR SECTION 5.6.1.3.12 DETAILS COMPLIANCE l WITH SRP SECTION 10.4.9 AND BTP ASB 10-1 \\ + Y l ^ ~ ^ ni mm e

x LOSS OF BULK AC POWER 1 (PLANT-BLACKOUT) FOR TWO HOURS .q y CURRENTLY ASSUMED AS A CONDITION FOR .' s ~ NATURAL CIRCULATION. EVALUATIONS I t NATURAL l CIRCULATION DESCRIBED PREVfOUSLY 1 l - AND IN CRBRP-ARD-0308 PRESENTS LONG TERM i h AND SHORT TERM ANALYTICAL RESULTS 1s, A :_ i4 052 17

4-5 Task A-44 $1st1Qn_Staskent A loss of all ac power is not a design-basis event f or the CRDRP facility. Nonetheless, requirements have been imposed to ensure that CRBRP will have substantial resistance to a loss of all ac power and that, even If a loss of all ac power should occur, there is reasonable assurance the core will be cooled and the ) health and safety of the pubile assured. These design, requirements are discussed below. A loss of offsite ac power involves a loss of both the preferred and backup sources of of f site power. If offsite ac power is lost, three diesel generators and their associated distributton systems will deliver emergency power to safety-related equipment. If both offsite and onsite ac power are lost, CRBRP is designed to remove reactor generated decay heat on natural circulation with the heat sink provided by the steam generator auxillary neat removal system. This capability ensures that adequate cooling can be maintained ror at least 2 hours, which allows time for restoration of ac power from either offsite or onsite sources. This capability has been described in PSAR Chapter 5.6 and CRBRP-ARD-0308. The decay heat generated in the spent fuel in the Ex-Vessel Storage Tank (EYST) is also capable of being removed by natural circulation. This is provided by the third EVST cooling loop which is designed to remove all decay heat produced in the EVST during natural circulation. This capability Is described in PSAR Chapter 9.1. The Ex-Vessel Transfer Machine Is designed to assure that cladding temperature is maintained within limits by a natural convection cooling system. This assures cooling of a fuel assembly in transit between the reactor and EVST. This capability Is discussed in PSAR Chapter 9.1. A two hour station blackout while handling a bare f uel assembly during normal Fuel Hundling Cell (FHC) operations could result in release of fission products to the environment. The potential radiation doses at the site boundary resulting from such a release are below established limits. The results of such r.eleases are discussed in PSAR Chapter 9.1. Based on the above cpnsiderations, the Project concludes thai there is reasonable assurance that CRBRP can proceed bef ore the ultimate resolution i J of this generic issue without endangering the health and safety l of the public.

M FAILURE LOCATION FOR GUILLOTINE l RUPTURE CASES EVALUATED ~ l l 797*8~ s l NOL. 794*9-r

788, p73311-r 784'7"h1C MSL 782'4"

= 4 i w 778*5-y 779*3" b _.s l j CASEI k d m 4 1D l o,,,,....,

~ ~ ASSUMPTIONS DOUBLE ENDED GUILLOTINE RUPTURES l l TRIP ON FLUX -\\/ PRESSURE ![ PLANT EXPECTED OPERATING CONDITIONS NOMINAL PEAK CORE TEMPERATURES .I-I MINIMUM SATURATION TEMPERATURE ,e ta P! '"* '*

~ DEMO DOUBLE ENDED PIPE RUPTURE RESULTS LBASED ON PLANT BEST ESTIMATE CONDITIONS) INNER i FUEL BLANKET ASSEMBLY ASSEMBLY l-l

• CASEI 1210 F

1265 F CASE ll 1185 F 1185 F CASE Ill 1110 F 1145 F l CONCLUSION: CASE I IS THE WORST CASE. fW

I ~ ~ PIPE RU.RTURE, CASE I POWER AND FLOW COASTDOWN VARIATION IN POWER AND FLOW FOR CASEI PO'NER AND FLOW RATIO 1.20' s 1.00 -j I o.8o l POWER Il o.60 - !l I I I 1 I o.40 - 1 FLOW i o.20 - 1r\\ y o_on 0.00 1.00 2.00 3.00 4.00 s.oo s.oo 7.00 8.00 9.00 10.00 TIME (SECONDS)

l DER AT RV INLET NOZZLE TOTAL BREAK FLOW-LB/SEC (DISCHARGE INTO GV) 12000. l 10000. 8000. 6000. 4000. % i f 2000. O. 10. 20. 30. 40. 50. 60. [ - 2000. TIME-SEC. D I 9 4 QQ Qe *O 99

ll II\\\\ l t ? . :.ii ?. f-06 0 5 E LZ 0 4 ON C E T S 0 E 3 E L M m I N T I 0 V K A A 2 R ERB T F A O 0 T 1 R U E O D SSA .O M M 0 0 0 0 0 0 0 U) 0 0 0 0 0 0 0 0 0 0 0 0 DM 0 0 0 0 0 0 I OB 2 0 8 6 4 2 L S( 1 1 f \\ 1 ) l

l' ll ~ 1 - F ~ n e l 06 0 i E 5 LZZO 0 i 4 N

1. i

) C T EL E E Z .S Z 0 - L O 3E N N M T T T I I F E V L L ET R VU 0 i EO 2 L T V a A N R LM R EO 0 S R 1 E S F E D VDE RR OU TS 0 CA AE 0 0 0 0 0 0 0 o, EM6 5 4 3 2 1 0 g R( 1 1 1 1 1 1 1 i ,Il11 l l 11f'

I o ~ DER AT RV INLET NOZZLE SODIUM LEVEL IN GV-FT (MEASURED FROM RV INLET NOZZLE) 30. e 20. 10. O. - 10., Y - 20. O. 10. 20. 30. 40. 50. 60. 4 -Y TIME-SEC.

SIMPLIFIED HYDRAULIC PROFILE-PHTS REACTOR INLET INTO GUARD VESSEL o i ELEVATION 816' e r, M E J 5 ti E E 6 i J ........ 7 9 7 ' 8

• ^- -- -h-L _

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FORE'2M MAXIMUM TEMPERATURES FOR ~ THE WORST CASE DOUBLE. ENDED ~ PIPE RUPTURE AS CALCULATED BY DEMO (CASE 1) (BASED ON PLANT BEST ESTIMATE CONDITIONS) fI: MAXIMUM COOLANT SATURATION MARGIN i TEMPERATURE TEMPERATURE TO BOILING l FUEL ASSEMBLY 1399 F 1586 F 187"F f INNER BLANKET 1380 F 1586"F 206*F ASSEMBLY OUTER BLANKET 1345 F 1586 F 240*F ASSEMBLY l l + l........

1 SHUTDOWN HEAT REMOVAL MEETING December 16, 1982 Attendees erganisation Name J CRBRP-P0/PHC Al Meller Brookhaven National Lab J. G. Guppy W-ARD R. A. Markley ANL Vipin L. Shah BNL Mohsen Khatib-Rahbar ARD F. M. Heck W-ARD Ron Coffield Westinghouse (WLLCO) Stephen Additon NRC-Idaho Richard E. Ireland NRC - Consultant L. N. Rib NRC CRBR-PO M. B. Holz EG&G Idaho T. L. Kinnaman Barthold & Associates, Inc. W. P. Earthold W-OR D. L. DeMott EG&G Idaho Bob Capp BNL K. R. Perkins BNL G. J. Van Tuyle NRC Stephen Sands NRC Tom King NRC Rich Stark CRBRP-PO-DOE Don Florek W-ARD R. Lowrie W-OR R. Hottel CRBRP-DOE-HQ Bill Murphie i l I e I .}}