ML18139C308

Reload Safety Evaluation for Surry 1 Cycle 7 Redesigned Core (Part 2)

ML18139C308
Person / Time
Site:	Surry
Issue date:	04/30/1983
From:	Cross R, Smith N, Suwal G Virginia Power (Virginia Electric & Power Co)
To:
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ML18139C307	List: ML18139C306 ML18139C308 ML20073H063
References
289, NUDOCS 8304180480
Download: ML18139C308 (18)

Text

NFE TcCHl\\ilCAL ~EP~RT

~~. 2f9 R~LG~D 5~FETY ~VALUATI0~ F~R suqhv 1 CYCL~

7 qEDESISN~D CORE

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e TABLE OF CONTENTS TITLE

1.0 INTRODUCTION

AND SUM~ARY

1.1 INTRODUCTION

1,2 GENERAL DESCRIPTION

1.3 CONCLUSION

S 2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN 2.2 NUCLEAR DESIGN 2.3 THERMAL AND HYDRAULIC DESIGN 3.0 POWER CAPAaILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY 3.2 ACCIDENT EVALUATION 3.2.1 KIN~TICS PARAMETERS 3.2.2 CONTROL ROD ~ORTHS.

3.2.3 CORE PEAKING FACTORS 4.0 TECHNICAL SPECIFICATIONS CHANGES 5.J REFERENCES e

PAGE 1

PAGE 3

3 3

5 6

6 6

B 9

9 9

10 11 11 13 14

e LIST OF TABLES TA5LE -

TITLE 1

FUE~ ASSEMBLY D~SIGN PARA~ETERS 2

KINETICS CrlARACT~RISTICS 3

SHUTDOWN REQUIREMENTS AND MARGI~S LIST OF FIGURES FIGURE TITLE CORE LOADING PATTERN 1

2 CONTROL ROD INSERTION LIMITS FOR 3-LOOP NORMAL OPERATION -

UNIT 1 PAG::

2 PAGE 15 16 17 PAGE 18 19

e PAGE 3

1.0 INTRODUCTION

AND

SUMMARY

l

• l I N_T ROD UC T I ON THIS REPORT PRESENTS AN EVALUATION FOR SURRY POWER STATION UNIT 1, CYCLE

7. WHICH DEMONSTRATES THAT THE REDESIGNED CYCLc 7 CORE WILL NOT ADVERSELY AFFECT THE SAFETY ACCOMPLISHED UTILIZING THE OF THE PLANT.

THIS METHODOLOGY DESCRIBSD EVALUATION WA.S IN WCAP-9272, HWESTI~GHOUSE RELOAD SAFET*Y EVALUATION METHODOLOGY" (REF. 1).

BASED UPO~

THE ABOVE REFERENCED METHODOLOGY, ONLY THOSE I~CIDENTS ANALYZEJ AND REPORTED IN THE FSAR (REF. 2) WHICH COULD POTENTIALLY BE AFFECTED BY THE FUEL RELOAD HAVE BEEN REVIEWED FOR THE CYCLE 7 DESIGN DESCRIBEO HEREIN.

NO NEW TRANSIENT ANALYSES WERE REQUIRED FOR THE CYCLE 7

DESIGN.

THE JUSTIFICATION FOR THE APPLICABILITY OF PREVIOUS RESULTS IS PROVIDED.

1.2 GENERAL DESCRIPTION THE SURRY 1 REACTOR CORE IS COMPRISED OF 157 FUEL ASSEMBLIES ARRA~GEO IN THE CONFIGURATION SHOWN IN FIGURE

1.

DURI~G THE* CYCLE 6/7 REFUELING, 94 FUEL AS~EMBLIES WERE-REPLACED WITH 64 FRESH REGION 9 ASS EM BL IE S Af'JD WITH

'3 0 ASS EM 8 L IE S I~ RADIATED IN EARL IE R CYCLES

• T_H E PATTERN FOR CYCLE 7

IS SHOWN lN FIGURE 1. A

SUMMARY

OF TH~ CYCLE 7 FUEL INVENTORY IS GIVEN IN TABLE 1.

NOMINAL COPE DESIGN PAR~METERS UTILIZED FOR CYCLE 7 ARE AS ~OLLOWS:

CORE POW~R (MWT)

SYSTEM PRESSURE (PSIA).

VESSEL A-VERAGE TEMPERATURE (-Fl THERMAL DESIGN FLOW (GPM)

AVERAGE LI~EAR POWER DENSITY (KW/FT)

(BASED ON HOT, OENSIFIED CORE AVERAGE STACK HEIGHT OF 143.6 INCHES) 2441 2250 574.4 265.500 6.2 PAGE 4

PAGE 5

1.3 CONCLUSION

S FROM THE EVALUATION.PRESENJEO IN THIS REPORT, IT IS CONCLUDED THAT THE

. CYCLE. 7 DE~IGN DOES NOT RESULT IN THE PREVIOUSLY ACCEPTABLE SAFETY LIMITS BEING EXCEEDED FOR ANY INCIDENT AND CONSEQUENTLY NO U~REVIEWED SAFETY QUESTIONS EXIST AS A RESULT OF THIS RELOAD *. THE CONCLUSIONS ARE BASED 0~ THE FOLLOWING:

1. AN ACTUAL CYCLE 6 SURNUP OF 16491 MWD/MTU

2. CYCLE 7 6URNUP WILL NOT EXCEED "13000 ~WD/MTU (NOMINAL END OF REACTIVITY PLUS APPROXIMATELY 1500 MWD/MTU OF COASTDOWN)

3. THE~E IS ADHERENCE TO PLANT OPERATING LIMITATIONS AS GIVEN IN THE TECHNICAL SPECIFICATIONS AND THE AMENDMENT T~ERETO PROPOSED IN SECTION 4

PAGE 6

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN THE MECHANICAL DESIGN OF THE REGION 9 FUEL ASSEMBLIES IS THE SAME AS THE REGION B ASSEMBLIESs TASLE l COMPARES PERTINE~T DESIG~ PARAMETERS OF THE VARIOUS FUEL REGIONS*

THE REGION 9 FUEL HAS BEEN DESIGNED ACCORDING TO THE FUEL PERFORMANCE MODEL IN REFERENCE 3.

THE FUEL IS DESIGNED AND OPERATED SO THAT CLAD FLATTENING WILL NOT OCCUR, AS PREDICTED BY THE WESTINGHOUSE MODEL (REF. 4).

FOR ALL FUEL REGIONS, I

THE FUEL ROD INTERNAL PRESSURE DESIGN BASIS, WHICH IS ACCEPTABLE AS SHOWN IN REFERENCE 5, IS SATISFIED.

WESTINGHOUSE HAS HAD CONSIDERABLE EXPERIENCE WITH ZIRCALOY CLAD FUEL.

THIS EXPER.IENCE IS EXTENSIVELY DESCRIBED IN WCAP-8183, "OPERATIONAL EXPERIENCE WITH WESTINGHOUSE CORESH (REF. 6). THIS REPORT IS UPDATED ANNUALLY.

2.2 NUCLEAR,DESIGN THE CYCLE 7

CORE LOADING HAS A LOCA FQ LIMIT OF 2el8 UNDER NORMAL OPERATING CONDITIONS. THE MAXIMUM ANALYTICALLY PREDICTED FQ FOR CYCLE 7 IS 2.15; FQ IS LESS THAN THE LIMIT AT ALL CORE ELEVATIONS FOR THIS CYCLE.

THEREFORE, FREQUENT AXIAL POWER DISTRI3UTION MONITORING IS NOT REQUIRED.

e PAGE 7

TABLE 2

PROVIDES A

SUMMARY

OF CHANGES IN THE CYCLE 7 KINETICS CHARiCTERISTICS COMPARED WITH THE CURRENT LJMITS 9ASED ON PREVIOUSLY SUBMIT~ED ACCiDENT ANALYSES.

AS SHOWN IN THE TABLE, ONLY ONE OF THE CYCLE 7 PARA_METERS, THE MOST NEGATIVE DOPPLER TEMPERATURE COEFFICIENT, FALLS OUTSIDE ITS CURRENT LIMIT. THIS PARAMETER IS EVALUATED IN SECTION TABLE 3

PROVIDES THE CONTROL ROD WORTHS AND REQUIREMENTS AT THE MOST LIMITING CONDITIONS DURING THE CYCLE. THE REQUIRED SHUTDOWN MARGIN IS BASED ON PREVIOUSLY SUBMITTED ACCIDENT "ANALYSES (REFQ 2)g THE AVAILABLE

.,I SHUTDOW~ MARGIN EXCEEDS THE MINIMUM REQUIRED.

WHILE ALL OTHER DESIGN CONSTRAINTS HAVE BEEN ASSESSED AND.ARE MET FOR THE CU~RENT ROD INSERTION LIMITS, THE PREDICTED RADIAL POWER PEAKING FOR CYCLE 7 rs IN EXCESS OF THE CURRENT DESIGN LIMIT AT HOT FULL POWtR (ONLY FOR BUR~UPS _LESS THAN 1000 MWD/MTU)

AND AT HOT ZERO POWER (MAJORITY OF CYCLE LIFE) FOR THE CONTROL ROD INSERTION LIMITS OF THE

_CURRENTLY APPROVED TECHNICAL SPECIFICATIONS. AS A RESULT, A CHA~GE TO THE TECHNICAL SPECIFICATIONS rs PROPuSED WHICH RAISES THE ROD INSERTION LIMITS.

AT THE REVISED INSERTION LIMITS, PROVIDEJ IN FIGURE 2, THE RADIAL PEAKING FACTORS ARE WITHIN THE APPROPRIATE CURRENT DESIGN LIMITS.

(HOWEVER.

IT SHOULD BE NOTEb THAT THE CURRENT ROD INSERTION LIMITS WOULD BE ACCEPTABLE AFTER ATTAINING A BURNUP OF 1000 MWD/MTU AND IF THE RADIAL POWER PEAKING FACTOR DESIGN LIMIT WERE eASED ON

PAGE 8

A 0.3 PART POwER ~ULTIPLIER, I.E.,

FDH = 1.55(1+0.3(1-Pl)

P= FRACTION OF RATED POWER INSTEAD OF THE. 0.2 VALUE WHICH IS CURRENTLY APPROVED.)

THE LOADING cor~TAINS A TOTAL OF 608 FRESH BURNA9LE POISON RODS LOCATED IN 52 OF THE REGION 9 FUEL ASSEMBLIES, AND 8 DEPLETED BURNABLE POISON RODS IN ONE OF THE REGION BB ASSEMBLlES. THREE SECONDARY SOURCES WILL BE USED AS SHOwN IN FIGURE 1.

2.3 THERMAL AND HYDRAULIC DESIGN NO SIGNtFiCANT VARIATIONS IN THERMAL ~ARGINS WILL RESULT FROM THE CYCLE 7

RELOAD. THE PRESENT DNB CORE LIMITS (REFERENCE 7) HAVE BEEN FOUND TO BE CO~S~RVATIVE FOR CYCLE 7

PAGE 9

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY THE PLANT POWER CAPABILITY IS EVALUATED CONSIDERING THE CONSEQUENCE~ OF THOSE INCIDENTS EXAMINED IN THE FSAR (~EF. 2) USING THE PREVIOUSLY ACCEPTED DESIGN BASIS.

IT IS CONCLUDED THAT THE CORE RELOAD WILL NOT ADVERSELY AFFECT THE ABILITY TO SAFELY OPERATE AT 100 PERCENT OF RATED POWER DURING CYCLE

7.

FOR THE EVALUATION PERFORMED TO ADDRESS OVERPOWER CONCERNS, THE FUEL CENTERLINE TEMPERATURE LIMIT OF 4700-F CAN BE ACCOMMODATED.

WITH MARGIN IN THE CYCLE 7 CORE USING THE METHODOLOGY DISCUSSED IN REFERENCE 1.

THE.TIME DEPENDENT DENSIFICATION MODEL (REF.

8)

WAS USED FOR THESE FUEL TEMPERATURE EVALUATIONS.

THE LOCA LIMIT AT RATED POWER CAN BE MET BY MAINTAINING FQ AT, OR BELOW, 2.18.

3.-z ACCIDENT EVALUATION THE EFFECTS OF THE RELOAD ON THE DESIGN BASIS AND POSTULATED INCIDENTS ANALYZED IN THE FSAR (REF.

Zl WERE EXAMINED.

IN ALL CASES, IT WAS FOUND THAT THE EFFECTS WERE ACCOMMODATED WITHIN THE CONSERVATISM OF THE ASSUMPTIONS USED IN THE P~EVIOUSLY APPLICABLE SAFETY ANALYSES.

PAGE 10 A

CORE RELGAD CAN TYPICALLY AFFECT ACCIDENT ANALYSIS INPUT PARAMETERS IN THE FOLLOWING AREAS:

CORE KINETIC CHARACTERISTICS, CONTROL ROD WORTHS,-

AND CORE PEAKING ~ACTORS.

CYCLE 7 PARAMETERS IN EACH OF THESE THREE AREAS WERE EXAMINED AS DISCUSSED BELOW TO ASCERTAIN WHETHER NEW ACCIDENT ANALYSES WERE REQUIRED.

3.2.1 KINETICS PARAMETERS A

SUMMARY

OF THE EVALUATION OF CYCLE 7 CORE PHYSICS PARAMETERS WITH CURRENT LIMITS IS GIVEN IN TABLE 2.

THE DELAYED NEUTRON FRACTIONS, MODERATOR TEMPERATURE COEFFICIENTS, AND PROMPT NEUTRON LIFETIME ARE WITHIN THE BOUNDS OF THE C0RRENT LIMITS.

THE MODERATOR TEMPERATURE COEFFICicNT OPERATION WILL BE ZERO OR NEGATIVE DURING NORMAL OPERATION, ALTHOUGH WITH A

SLIGHTLY POSITIVE COEFFICIENT IS ALLOWED BELOW FULL POWER OPERATION.

THE MOST NEGATIVE DOPPLER TEMPERATURE COEFFICIENT IS

-2.3 PCM/-F COMPARED TO THE LIMIT OF -1.6 PCM/-F.

THIS COEFFICIENT IS USED IN CONJUNCTION WITH THE DOPPLER POWER COEFFICIENT FOR FUEL TEMPERATURE CHANGES IN TRANSIENTS WHERE THE CORE WATER TEMPERATURE DROPS.

FOR THE MOST SEVERE REACTIVITY ADDITION ACCIDENT (STARTUP O*F AN INACTIVE LOOP),

THIS AMOUNTS TO LESS THAN A 3% INC~EASE IN TOTAL POSITIVE

~EACTIVITY INSERTION.

THIS ~OULD YIELD A NEGLIGIBLE INCREASE IN PEAK POWER WHICH CAN BE ACCOMMODATED IN ALL OF THE FSAR COOLDOWN EVENTS.

PAGE 11 3.2.2 CONTROL ROD WORTHS CHANGES-IN CONTROL ROD WORTH MAY AFFECT DIFFERENTIAL ROD WORTHS, SHUTDO~N

MARGIN, EJECTED ROD

WORTHS, AND TRIP REACTIVITY.

TABLE 2 SHOWS THAT THE MAXIMUM DIFFERENTIAL ROD WORTH OF TWO RCCA CONTROL BANKS MOVING

.TOGETHER IN THEIR HIGHEST WORTH REGION FOR CYCLE 7 MEETS THE CURRENT LIMIT.

TABLE 3

SHOWS THAT THE CYCLE 7

SHUTDOWN ~ARGIN REQUiqEMENTS ARE SATISFIED.

EJECTED ROD WORTHS FOR CYCLE 7 ARE WITHIN THE BOUNDS OF THE CURRENT LIMITS.

AS A

CONDITION FOR USING THE ROD.SWAP TECHNIQUE FOR MEASURING ROD

WORTHS, THE NRC HAS REQUIRED THAT A

COMPARISON BE MADE BETWEEN WESTINGHOUSE AND VEPCO SHUTDOWN MARGIN CALCULATIONS (REFERENCE 9). THIS COMPARISON IS WILL BE PERFORMED AND REVIEWED BY THE STATION NUCLEAR SAFETY ANO OPERATING COMMITTEE AND BY THE SAFETY EVALUATION AND CONTROL STAFF PRIOR TO U~IT 1 STARTUP.

3.z.3 CORE PEAKING FACTORS THE PEAKING FACTORS FOR THE STEAMLINE BREAK HAVE BEEN EVALUATED ~ND ARE WITHIN THE BOUNDS OF THE PREVIOUS SA~ETY ANALYSIS LIMITS. IN ADDITION, THE PEAKING FACTORS FOLLOWING CONTROL ROD EJECTION ARE WITHIN THE LI~ITS OF PREVIOUS ANALYSIS VALUES FOR ALL CASES*

e PAGE 12 AS STATED IN SECTION 2.2, THE MAXIMUM ANALYTICALLY.PREDICTED LOCAL PEAKING FACTOR FOR CYCLE 7

IS LESS THAN THE FQ LIMIT.

THEREFORE, FREQUENT AXIAL POWER DISTRIBUTION MONITORING WILL NOT BE REQUIRED DURING CYCLE

7. IT IS ANTICIPATED THAT A CHANGE IN THE F-DELTA-H PART POWE~

MULTIPLIER FROM 0.2 TO 0.3 WILL 3E REQUESTED IN ORDER TO RESTORE THE ROD INSERTION LIMITS TO THE CURRENTLY APPROVED VALUES UPON COMPLETION OF A0 PROXIMATELY 1000 MWD/MTU OF CYCLE LIFETIME.

I e

PAGE 13 4.0 TECHNICAL SPECIFICATIONS CHANGES AS 0I£CUSSED IN SECTION. 2.2, A CHANGE TO THE CONTROL ROD INSERTION LIMITS OF

~HE TECHNICAL SPECIFICATIONS IS BEING PROPOSED FOR CYCLE 7 OPERATION.

THE REVISED

LIMITS, PRESENTED IN FIGURE 2, REPRESENT A DECREASE (SHALLOWER INSERTION)

OF 3% OF FULL ROD T~AVEL AT HOT FULL POWER AND 20% AT HOT ZERO POWER WITH RESPECT TO THE CURRENTLY APPROVED INSERTION LIMITS.

e PAGE 14

5.0 REFERENCES

1. F. M. BORDELON, ET AL.,~ WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," wCAP-9272, MARCH 1978.

2. SURRY POWER STATION UNIT 1 ANO 2, FINAL SAFETY ANALYSIS REPORT, DECEMBER 1, 1969.

3. MILLER, J. V., (ED), "IMPROVED ANALYTICAL MODELS USED IN WESTINGHOUSE FUEL ROD DESIGN COMPUTATIONStt, W~AP-8785, OCTOBER J.976.

4. GEORGE, R. A., ET AL.,"REVISED CLAD FLATTENING MODEL", WCAP-8377, JULY 1974.

5. RISHER, O. H., ET AL.,"SAFETY ANALYSIS FOR THE REVISED FUEL ROD INTERNAL PRESSURE DESIGN BASIS," WCAP-8964, JUNE 1977.

6. JONES, R. G., IORII, J. A., ttQPERATIONAL "EXPERIENCE 'WITH WESTINGHOUSE CORES", WCAP-818311 REVISION 11, MAY 1982.

7a SURRY POWER STATION UNITS 1 AND 2, TECHNICAL SPECIFICATIONS, DOCKET NOS. 50-280 ANO 50-281, AS AMENDED.

8

• HELLMAN, J

• M. ( ED. ), "FUEL DENS I FICA TI ON EXP ER l:'-1 ENT AL RESULTS AN;J MODEL FOR REACTOR OPERAT.101'~", WCAP-8219-A, MARCH 1975.

9. LETTER FROM R. L. TEDESCO (NRC) TOW. N. THOMAS (VEPCOl, DATED NOVE~8ER 7, 1980.

e e

BATCH.

ENRICHMENT (W/0 U23 5)

DENSITY

(% 1HEORETICAL)

NUMBER OF ASSEMBLiES BURNUP AT BEGINNING OF CYCLE 7 (MWD/MTU)l BURNUP AT END OF CYCLE 7

("1WD/MTU)2 MTU PER REGION3 TAB~E 1 FUEL ASSEMBLY DESIGN PARAMETERS SURRY UNIT 1 CYCLE 7 4C**

S2/6B 6C**

7A 75 3.33 3e20 2Q90 2.,90 3&39 94.40 94.48 94.30 94.50 94.70 22 2

8 10 6

• 26200 27300 25100 24300 30000 35900 32700 37700 32600 42600 10.02 0.92 3.64 4.56 2.75 FROM SURRY 1 CYCLE 4

~ROM SURRY 1 CYCLE 5 8A.

3.22 94.61 12 20100 31000 5.50 PAGE 15 SB 9.........

3.40 3.59 94.58 94.60 33 64 18000 0

32600 14700 15.08 29.38

..... '"'l'"'"'I""

INCLUDES 4 FRESH ASSEMBLIES FROM SUR~Y 2 BATCH 9. ~ITH NOMINAL ENRICHMENT AND DENSITY OF 3.60% AND 94.5 W/0 RESPECTIVELY.

1. ASSUME END-OF-CYCLE 6 BURNUP OF 16500 MWD/MTU

2. ASSUME END-OF-CYCLE 7 BURNUP OF 13000 MWD/MTU; ALL BATCH AVERAGE BURNUPS ARE <37000 MWD/MTU.

3. INITIAL MTU j

e TABLE 2 KINETICS CHARACTERISTICS SURRY 1 CYCLE 7 PAGE 16 PARAMETER-I CURRENT LIMIT CYCLE 7 VALVES MODERATOR TEMPERATURE COEFFICIENT (PCM/-F)l 3.0 TO -35.0 MOST NEGATIVE DOPPLER-ONLY TEMPERATURE COEFFICIENT (PCM/-F)

-1.6 LEAST NEGATIVE DOPPLER-ONLY POWER COEFFICIENT, HZP TO HFP (PCM/% POWER)

-11.4 TO -6.00 MINIMUM DELAYED NEUTRON FRACTION, BOL TO EGL

(%)

0.55 TO 0.44 I

MAXIMUM PROMPT NEUTRON J

LIF::TIME (MICRO SEC)

I 26 I

MAXIMUM DIFFERENTIAL ROD I

WORTH OF TWO BANKS I

MOVING TOGETHER (PCM/SEC) I 75 I

I

1. 1 PCM= 10-5 DK/K

2. SEE SECTION 3.2.1 WITHIN CURRENT LIMITS

-2.3 2 WITHIN CURRENT LIMITS WITHIN CURRENT LIMITS

<26 (75

i e

e TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS SURRY 1 CYCLE 7 CONTROL ROD WORTHS (% DK/K)

ALL R~CS INSERTED LESS WORST STUCK ROD(l)

(ll.LESS 10% (2)

CONTROL ROD REQUIREMENTS (% DK/K)

REACTIVITY DEFECTS (COMBINED DOPPLER, TAVE, VOID, AND REDISTRIBUTION EFFECTS}

ROD INSERTION ALLOWANCE TOTAL REQUIREMENTS (3)

SHUTDOWN MARGIN ((2)-(3))

REQUIRED SHUTDOWN MARGIN

---

~

BOC 5.83 5.25 1

• 00 2.98 2.21 1.11 PAGE 17 EOC 6.79 6.11 3.35 a.so 3.. 85 2.26 1.11