ML17305B558
| ML17305B558 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/17/1991 |
| From: | Catherine Thompson Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9106030139 | |
| Download: ML17305B558 (57) | |
Text
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Docket No. 50-530 UNITED STATES NUCLEAR REGULATORY COIVIMISSION WASHINGTON, D.C. 20555 May 17, 1991 LICENSEE
'ACILITY
'UBJECT:
ARIZONA PUBLIC SERVICE COMPANY PALO VERDE NUCLEAR GENERATING STATION, UNIT 3
SUMMARY
OF MEETING WITH ARIZONA PUBLIC SERVICE COMPANY AND COMBUSTION ENGINEERING ON MAY 10, 1991, REGARDING UNIT 3 CYCLE 3 RELOAD ANALYSIS On May 10, 1991, representatives of Arizona Public Service Company and Combustion Engineering met with members of the NRC staff to discuss one of the analyzed events from the Palo Verde Nuclear Generating Station (PVNGS) Unit 3, Cycle 3 reload submittal:
the inadvertent opening of an atmospheric dump valve with loss of offsite power.
A list of those who attended the meeting is attached as Enclosure 1.
A copy of the slides used by the licensee to brief the NRC staff is attached as Enclosure 2.
The inadvertent opening of a steam generator atmospheric dump valve with loss of offsite power was reanalyzed for PVNGS Unit 3 Cycle 3 because of a more adverse power distribution which would result in more fuel rods at high power levels compared to the previous cycle.
The initial analysis for this event resulted in a calculated offsite dose of 228 rem thyroid.
Using a statistical convolution method, the licensee calculated the resultant offsite dose for this event to be 30 rem thyroid and less than 2 rem whole bog, thereby meeting the staff's criterion of a "small fraction" (10% or less) of 10 CFR 100 guidelines.
- However, NRC requested this meeting with the licensee to discuss:
(1) the appropriateness of using the convolution method for this analysis, and (2) additional options for preventing or mitigating consequences of this transient.
To address this analysis, APS presented eight options:
two analytic solutions, five penalty and setpoint
- changes, and one potential plant modification (see ).
For each of'hese
- options, APS reviewed the implementation, impact on the plant, and offered conclusions on feasibility and acceptability.
APS concluded that changing penalty and setpoints may increase the potential for unnecessary reactor trips and place restrictions on plant operations, and were therefore, unacceptable.
Plant modifications were also not found to be feasible.
Furthermore, APS noted that the assumptions and methods used in the analysis for Unit 3 Cycle 3 are conservative.
Based on these analyses and conclusions, APS concluded that the convolution method is a technically valid and appropriate method of determining fuel failure which would not require additional penalties, setpoint changes or design changes.
The NRC staff noted that (1) the options for changing setpoints and penalties or making plant modifications appear to be unacceptable, (2) the assumptions and methods used in the Unit 3 Cycle 3 analysis are conservative, and 3) the 9iOb030i39 9i05i7 I'DR 0%585@0~
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Arizona Public Service Company dose consequences of this analysis using the convolution method meet the acceptable limit of 10% of 10 CFR 100.
Therefore, the staff found that this analysis for PVNGS Unit 3 Cycle 3 is acceptable and plans to issue an SER by May 20, 1991, to support a Unit 3 restart on May 23, 1991.
APS agreed that all future PVNGS reload analyses performed for this event must use the assumptions and methods used for the Unit 3 Cycle 3 analysis.
Any changes to these assump-tions and methods should be submitted to the NRC staff, including a discussion of why the statistical convolution methodology remains acceptable for this event.
cc w/enclosures:
See next page Original signed by Catherine Thompson, Project Manager Project Directorate V
Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation DISTRIBUTION IIRC I LPDR FN CThompson EJordan ACRS (10)
KEccleston RJones LKopp JPartlow JCa 1 dwe1 1 OGC BBoger MVirgilio PD5 r/f PD5 p/f OFC
- LA/PD5/DRPW ~
- PM/PD5/D RPW:D/PD5/DRP W NAME
- DFoster DATE: (/t6/91
- CT o son:pm: JDyer 91:
////91 ocumen arne:
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Arizona Public Service Company Palo Ver de CC:
Arthur C. Gehr, Esq.
Snell 5 Wilmer 3100 Valley Center Phoenix, Arizona 85073 James A. Beoletto, Esq.
Southern California Edison Company P. 0.
Box 800
- Rosemead, California 91770 Senior Resident Inspector U.S. Nuclear Regulatory Commission HC-03 Box 293-NR Buckeye, Arizona 85326 Reg iona l Adminis trator, Reg ion V
U. S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Mr. Char les B. Brinkman, Manager Washington Nuclear Operations ABB Combustion Engineering Nuclear Power 12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Charles Tedford, Director Arizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arizona 85040 Chairman Maricopa County Board of Supervisors 111 South Third Avenue Phoenix, Arizona 85003 Jack R.
- Newman, Esq.
Newman 5 Holtzinger, P.C.
1615 L Street, N.W., Suite 1000 Washington, D.C.
20036 Ignacio R. Troncoso Senior Vice President El Paso Electric Company Post Office Box 982 El Pasco, Texas 79960 Roy P. Lessey, Jr., Esq.
Bradley W. Jones, Esq.
Akin, Gump, Strauss, Hauer and Feld E 1 Paso Electric Company 1333 New Hampshire Ave., Suite 400 Washington, D.C.
20036 Mr. William F.
Conway Executive Vice President, Nuclear Arizona Public Service Company P. 0.
Box 53999 Phoenix, Arizona 85072-3999
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ENCLOSURE 1
NRC-APS MEETING ON PALO VERDE UNIT 3 CYCLE 3 RELOAD HAY 10, 1991 List of Attendees Name Catherine Thompson James Dyer Robert Jones Ken Eccleston Thomas Radack Steve O'Hearn Hario Robels Earl Schulz Charles Brinkman Steve Toelle Nils Breckenridge A. K. Kraini k Stephen Troisi Hike Friedlander Michael E. Powell Ronald J. Stevens Or anization NRC/PDV NRC/PDV NRC/SRXB NRC/PRPB ABB-CE ABB-CE ABB-CE ABB-CE ABB-CE ABB-CE ABB-CE APS APS APS APS APS
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MEETING BETWEEN ARIZONA PUBLIC SERVICE COMPANY AND U. S.
NUCLEAR REGULATORY COMMISSION TO DISCUSS THE PALO VERDE UNIT 3, CYCLE 3 RELOAD SUBMITTAL MAY 10, 1991
J
AGE DA 1.
INTRODUCTION 2.
PURPOSE 3.
BACKGROUND 4
OVERVIEW OF TRANSIENT 5.
CONSERVATISMS IN METHODOLOGY 6.
OPTIONS 7
OPEN DISCUSSIONS 8.
CONCLUSIONS
v
PURPOSE TO PROVIDE THE INFORMATION NECESSARY FOR NRC TO COMPLETE THEIR REVIEW AND ISSUE THE PALO VERDE UNIT 3, CYCLE 3 RELOAD AMENDMENT BY MAY 20, 1991.
1 e4
BACKGROUND r
INADVERTENT OPENING OF AN ATMOSPHERIC DUMP VALVE WITH LOSS OF OFFSITE POWER 0
8 o FUEL FAILURE USED IN INITIAL CORES AND SUBSEQUENT RELOADS UP TO UNIT 2t CYCLE 3 0
12 o FUEL FAILURE USED IN UNIT 2 t CYCLE 3 0
UNIT 3, CYCLE 3
AMENDMENT AND METHODOLOGY ARE IDENTICAL TO UNIT 2t CYCLE 3 RELOAD SUBMITTAL 0
NRC REQUESTED DOSE CALCULATION FOR UNIT 3t CYCLE 3
0 UNIT 3, CYCLE 3 RESULT:
228 REM AT SITE BOUNDARY 0
. NRC REQUESTED REANALYSIS TO MEET SMALL FRACTION OF 10CFR100 0
CONVOLUTION METHOD USED FOR REANALYSIS; RESULTED IN "SMALL FRACTION" OF 10CFR100 0
NRC REQUESTED APS TO ADDRESS ADDITIONAL OPTIONS
OVERVIEW 'OF TRA SIE T
INADVERTENT OPENING OF AN ATMOSPHERIC DUMP VALVE (TOADY) WITH LOSS OF OFFSITE POWER (LOP)
SEQUENCE OF EVENTS:
0 OPERATOR OPENS ADV (ADV STAYS OPEN FOR 30 MINUTES) h 0
REACTOR POWER INCREASE TO 113'FTER 30 SECONDS STEADY STATE HOT CHANNEL DNB ACHIEVED (CLOSE TO SAFDL REACHED W/0 RX TRIP) 0 TURBINE TRIP IS ASSUMED TO CAUSE LOSS OF OFFSITE POWER (45 SECONDS) 0 REACTOR TRIPS ON LOW DNB 'DUE TO LOW FLOW 0
MINIMUMTRANSIENT DNB OCCURS DUE TO HIGH POWER AND LOW FLOW CONDITION (46.1 SECONDS) 0 AFFECTED STEAM GENERATOR DRIES OUT (1150 SECONDS) 0 OPERATOR MANUALLY CLOSES ADV (1800 SECONDS)
ASSUMPTIONS:
0 ADY RELIEF CAPACITY IS 11o OF FULL POWER TURBXNE FLOW 0
TECH SPEC PRIMARY TO SECONDARY LEAKAGE 0
MOST REACTIVE CEA FAILS TO XNSERT FOLLOWXNG REACTOR TRIP 0
NO COINCXDENT CAUSE FOR TURBINE TRIP XDENTXFIED
I I
CONSERVATISMS IN METHODOLOGY FUEL FAILURE CRITERION IS CONSERVATIVE:
0 RODS ARE IN DNB FOR LESS THAN 5 SECONDS g
DUE TO CLAD OVERHEATING IS EXTREMELY UNLIKELY 0
DNBR LIMIT IS BASED ON DATA FROM WORST CRITICAL HEAT FLUX TEST (1.19 VS. 1.13 LIMIT),
0 CE-1 CHF CORRELATION WITH TONG F-FACTOR PROVIDES EXTREMELY CONSERVATIVE PREDICTIONS OF CHF FOR NON-UNIFORM AXIAL POWER DISTRIBUTIONS
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IOADV +
LOP Conditions at Time of IVlDNBR pressure inlet temperature (psia}
(')
2000 core average mass veiocity (Mlbm/hr-ft')
2.14 Range of CE-1 CHF Correlation pressure inlet temperature (psia}
1785
- 2415
(
F) 382
- 644 local mass veiocity (Mlbm/hr-ft')
0.87
- 3.2t
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TRANSIENT METHODOLOGY IS CONSERVATIVE:
0 TRANSIENT IS INITIATED FROM POWER OPERATING LIMIT 0
FLATTEST RADIAL POWER DISTRIBUTION FOR CYCLE IS USED 0
LOP IS ASSUMED TO OCCUR AT WORST TIME IN IOADV TRANSIENT 0
CETOP PROVIDES CONSERVATIVE DNBR VALUES:
CETOP IS CONSERVATIVE RELATIVE TO TORC ALL DNBR VALUES ARE CALCULATED ASSUMING ASSEMBLY INLET FLOW IS 73 o OF THE CORE AVERAGE
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OPTIONS ANALYTIC SOLUTIONS:
0 EXAMINATION OF THE FREaUENCY OF THE AS ANALYZED EVENT 0
STATISTICAL CONVOLUTION METHOD PENALTY 5,
SETPOINT CHANGES (INCREASED OPERATING RESTRICTIONS AND POSSIBLE UNNECESSARY REACTOR TRIPS):
0 LOWER CPC VOPT SETPOINT 0
INCREASE CPC PENALTY ON POWER 0
RESTRICT CPC ASI OPERATING BAND 0
CHANGES TO COLSS CONSTANTS - INCREASE DNB MARGIN 0
DECREASE ALLOWABLE RCS LEAKAGE POTENTIAL PLANT MODIFICATIONS:
0 ADV BLOCK VALVE CLOSURE
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EXAMINATION OF THE FREQUENCY OF THE "AS A ALYZED" EVE T IMPLEMENTATXON:
0 REQUIRES JUSTIFICATION OF LOW PROBABXLITY FOR THIS EVENT 0
EVALUATE ALL POSSIBLE ADV STUCK OPEN SEQUENCES IMPACT:
0 NO IMPACT ON PLANT OPERATXON CONCLUSION:
0 OVERALL FREQUENCY ON THE ORDER OF 10-7/YEAR/UNIT 0
CURRENT ANALYSIS REMAXNS ACCEPTABLE WITH OFFSITE DOSE OF 228 REM
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STATISTICAL CONVOLUTION METHOD IMPLEMENTATI ON:
0 APPLY CONVOLUTION FUEL FAILURE CALCULATION METHOD FOR THE TOADY W/LOP EVENT 0
USED FOR PALO VERDE CEA EJECTION AND SEIZED ROTOR/SHEARED SHAFT EVENTS 0
METHOD HAS BEEN USED ON OTHER DOCKETS IMPACT:
0 ACCEPTABILITY OF CONVOLUTION WOULD REQUIRE NO ADDITXONAL PENALTIES)
SETPOINT CHANGES OR DESIGN CHANGES CONCLUSION:
0 CONVOLUTION IS A
TECHNICALLY VALID METHOD OF DETERMINING FUEL FAILURE 0
CONVOLUTION RESULTS IN OFFSITE DOSE THAT XS A
"SMAI L FRACTION" PF 1GCFR100 0 SIGNIFICANT CONSERVATISMS REMAIN XN BOTH THE FAII URE CRITERION AND THE TRANSXENT METHODOLOGY
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LOWER CORE PROTECTl N CALCULATOR CPC VARIABLE OVER POWER TRIP VOPT SETPOI T
IMPLEMENTATXON:
0 DETERMINE ACCEPTABLE SETPOINT 0
IMPLEMENT SETPOINT CHANGE IN CPC 0
CPC VOPT CEILING SETPOINT WOULD NEED TO BE REDUCED FROM 110 o TO 105 o (THIS IS A PRELIMINARY VALUE) 0 CPC VOPT CEILING 5 OFFSET ARE ADDRESSABLE, RATE CONSTANTS ARE IN THE RELOAD DATA BLOCK IMPACT:
0 INCREASED POSSIBXLITY OF UNNECESSARY REACTOR TRIPS AND CHALLENGES TO SAFETY SYSTEMS 0
NORMAL PLANT MANEUVERING DURING START-UP RESTRICTED REQUIRES EXTENSIVE ANALYSIS TO QUANTIFY NEW SETPOINT ADDING POTENTIAL IMPACT ON STARTUP CONCLUSION:
0 NEW SETPOINT WOULD ENSURE THE OFFSITE DOSE WOULD NOT EXCEED "SMALL FRACTION" OF 10CFR100 0
'TIGHTENING UP THE VOPT TRIP SETPOINT WILL XNCREASE THE POSSIBILITY OF UNNECESSARY REACTOR TRIPS AND RESTRXCTS PLANT OPERATIONS
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I CREASE CPC PE ALTY 0 POWER IMPLEMENTATION:
0 INCREASE CPC ADDRESSABLE CONSTANT (BERR2)
BY 4 o 0
NO ADDITIONALANALYSIS REQUIRED IMPACT:
0 INCREASED POSSIBILITY OF UNNECESSARY REACTOR TRIPS AND CHALLENGES TO SAFETY SYSTEMS I'
ESTIMATED CPC MINIMUM DNBR FOR UNIT 3, CYCLE 3 TIME IN CORE LIFE W/0 PENALTY W/ PENALTY BOC EOC 1.46 1.55
- 1. 34 1.43 PRE-TRIP ALARM = 1. 39, TRXP SETPOINT
= 1. 24 CONCLUSION:
0 NEW SETPOINT WOULD ENSURE THE PLANT WOULD NOT FAXL FUEL IN EXCESS OF SMALL FRACTION OF 10CFR100 0
CPC MARGIN TO TRIP REDUCED BY ~.12 DNBR UNITS MAY BE IN PRE-TRXP CONDITION AT BOC h
0 POTENTIAL TO INCREASE UNNECESSARY PLANT REACTOR TRIPS AND RESTRICTS PLANT OPERATIONS
7 0
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RE TRICT CPC AXIAL SHAPE I DEX ASI OPERATI G
BA D
IMPLEMENTATION:
0 EXTENSIVE ANALYSXS EFFORT TO DETERMINE ASI DEPENDENCE 0
ANALYZE TO DETERMINE REQUIRED PENALTY VS. ASI 0
GENERATE NEW RELOAD DATA BLOCK DISK 0
INSTALL NEW CONSTANTS IMPACT:
0 INCREASED POSSIBILITY OF UNNECESSARY REACTOR TRIPS AND CHALLENGES TO SAFETY SYSTEMS 0
ADDS POTENTIAL IMPACT TO PLANT STARTUP 0
COULD MINXMIZE MARGIN IMPACT AT NOMINAL OPERATXNG CONDITIONS 0
REDUCED CPC DNBR MARGIN AT LIMITING ASI CONCLUSION:
0 FASTER INSERTXON OF NEGATIVE REACTIVXTY DURING
- SCRAM, LEADING TO REDUCED FUEL FAILURE 0
UNNECESSARY PLANT RESTRXCTIONS AND POTENTIAL FOR REACTOR TRIPS
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7 CHANGES TO CORE OPERATING LIMITS SUPERVISORY SYSTEM (COLSS CONSTANTS-INCREASE DNB MARGIN IMPLEMENTATION:
0 CHANGE ANALYSIS METHOD TO DETERMINE MARGIN REQUIREMENTS WHICH PRODUCE A DESIRED, REDUCTION IN FUEL FAILURE IMPACT:
o NEw COLSS MARGIN REQUIREMENTs woULD REsTRIGT PLANT OPERATION 0 'THER EVENTS (E G
~EXCESS TURBINE DEMAND) WOULD BE LIMITINGIN TERMS OF THE AMOUNT OF CALCULATED FUEL FAILURE CONCLUSION:
o NEM LIMITING EYENT woULD REQUIRE coMPLETF EVALUATION OF OTHER SAFETY ANALYSES o
ADDITIONAL MARGIN REQUIREMENT MAY PREVENT FULL POWER OPERATION O
POTENTIAL TO EXERCISE OPERATORS MORE FREQUENTLY
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DECREASE ALLOWABLE RCS LEAKAGE IMPLEMENTATION:
0 REDUCE T.S.
LIMIT ON PRIMARY TO SECONDARY LEAKAGE FROM 1 GPM 0
REDUCTION IN CALCULATED DOSE IS PROPORTIONAL TO THE REDUCTION IN THE ALLOWED TUBE LEAKAGE 0
ALLOWED LEAKAGE WOULD HAVE TO BE REDUCED TO ABOUT 0.05 - 0.1 GPM PER STEAM GENERATOR TO MEET SMALL FRACTION" OF 10CFR100 IMPACT:
0 WOULD NEED TO DETERMINE CHEMICAL OR ISOTOPIC METHOD TO MEASURE AND QUANTIFY A SMALL LEAK 0
DIFFICULT AND TIME CONSUMING TO LOCATE AND REPAIR SUCH A SMALL LEAK 0
REQUIRES PLANT SHUTDOWN TO REPAIR AND INCREASES MAN-REM EXPOSURES CONCLUSION:
0 UNACCEPTABLE IMPACT ON PLANT OPERATIONS
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ADV BLOCK VALVE CLOSURE IMPLEMENTATION:
0 BLOCK VALVE CLOSURE PRIOR TO STEAM DRYOUT 0
DEMONSTRATE VALVES WILL CLOSE AGAINST FULL FLOW 0
ASSESS LIGHTING REQUIREMENTS AT VALVE LOCATION 0
OPERATIONS STAFFING, PROCEDURES AND TRAINING IMPACT:
0 NEW DOSE CALCULATION REQUIRED BASED ON CREDIBLE BLOCK VALVE CLOSURE TIME 0
EXTENSIVE EFFORT (POSSIBLE DESIGN CHANGES)
TO IMPLEMENT CHANGES THAT WOULD ENSURE ABILITY TO CLOSE VALVES 0
OPERATOR WILL RECEIVE DOSE ATTEMPTING CLOSURE CONCLUSION:
0 MOULD NOT REDUCE THE OFFSITE DOSE SIGNIFICANTLY 0
NOT A FEASIBLE OPTION
CONCLUSIONS 0
CONSERVATXSMS ANALYSIS XN METHODOLOGY AND EVENT PROBABXLXTY IS ON THE ORDER OF 1 X 10 -7/YEAR/UNIT 0
THE CONVOLUTXON METHOD XS CONSERVATXVE TECHNXQUE FOR ANALYZXNGTHE CONSEQUENCES OF THE EVENT 0
OTHER OPTXONS NOT FEASXBLE OR RESULT XN UNNECESSARY PLANT OPERATIONAL RESTRXCTXONS AND INCREASED POTENTXAL FOR CHALLENGING SAFETY SYSTEMS 0
NRC SHOULD ISSUE THE UNIT 3, CYCLE 3
RELOAD AMENDMENT NO LATER THAN MAY 20, 1991
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APR 8
1991
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MEMORANDUM FOR:
J.
Dyer, Project Director (13E-16)
Project Directorate V
Division of Reactor Projects
- III, IV, V FROM:
SUBJECT'obert C. Jones, Chief Reactor Systems Branch Division of Systems Technology SER FOR PALO VERDE UNIT 3 CYCLE 3 RELOAD Plant Name:
Uti 1 ity:
TAC No(s).:
Docket No(s).:
Operating License:
Project Directorate:
Project Manager:
Review Branch:
Review Status:
Palo Verde Nuclear Generating Station Unit 3 Arizona Public Service Company 79785 50-530 NPF-74 Project Directorate V
C. Trammell SRXB/DST Complete By letter dated February 21, 1991, Arizona Public Service
- Company, licensee for the Palo Verde Nuclear Generating Station Unit 3 (PVNGS3) submitted a
reload safety analysis report in support of a request to reload and operate PVNGS3 for a third cycle at 100 percent rated core power of 3800 MWt.
The licensee also submitted proposed changes to the Technical Specifications (TS) to support Cycle 3 operation.
We have completed our review of the submitted information regarding fuels, physics, thermal-hydraulics, accident and transient analyses, proposed TS
- changes, and startup test procedures.
Based on our review, v/e find the proposed reload and associated TS changes acceptable.
Our technical evaluation and our SALP evaluation of the licensee are provided as Enclosures I and 2, respectively.
This completes the SRXB effort on TAC Bo. 79785.
)~tu Reactor Systems Branch Division of Systems Technology
Enclosures:
As stated cc w/enclosures:
A. Thadani PEB/SALP (10A-19)
C. Trammell (13E-16)
SRXB Members DISTRIBUTION~
SRXB R/F RJones LPhi 1 lips LKopp LKopp R/F Palo Verde P/F
Contact:
L. Kopp, SRXB/DST Ext. 20879 OFC
- SRXB:DST
- SRXB:D
- SRXB:DST NAME:LKOPP: gnk~~: LP I IP: RJONES BATE g/ I /91
'91 91 Document Name:
MEMO DYER, PALO VERDE/79785
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UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 ENCLOSURE I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TO PALO VERDE UNIT 3 CYCLE 3 RELOAD ARIZONA PUBLIC SERVICE COMPANY PALO VERDE UNIT 3 DOCKET NO. 50-530
1.0 INTRODUCTION
By letter dated February 21, 1991, Arizona Public Service
- Company, the licensee for,the Palo Verde Nuclear Generating Station Unit 3 (PVNGS3),
submitted a reload safety analysis report in support of a request to reload and operate PVNGS3 for a third cycle at 100 percent rated core power of 3800 MMt.
The licensee also submitted proposed changes to the Technical Specifica-tions (TS) to support Cycle 3 operation.
The Cycle 3 core will consist of 241 fuel assemblies.
Seventy-three Batch B
and 48 Batch C assemblies will be removed from the Cycle 2 core and replaced by 88 unirradiated Batch E assemblies.
One-hundred and four Batch D assemblies and 16 Batch C assemblies from the Cycle 2 core will be retained.
In addition, 33 Batch B assemblies discharged at end of Cycle I wi 11 be reinserted.
Burnup distribution is based on a Cycle 2 length of 436 effective full power days (EFPD).
Cycle 3 control element assembly patterns and in-core instrument locations remain the same as in Cycle 2.
The staff has reviewed Reference I and has prepared the following evaluation of the proposed TS changes, the fuel design, nuclear design, thermal-hydraulic design and accident/transient analyses associated with the Cycle 3 core.
2.0 EVALUATION OF FUEL DESIGN 2.1 Mechanical Desi n
The 88 Batch E assemblies to be added to the Cycle 3 core are identical in design to the Cycle 2 Batch D assemblies except for changes to the poison rod
- assembly, the lower end fitting, and center guide tube.
The poison rod assembly was increased in overall length from 160.918 inches to 161.168 inches to improve burnup capability and reduce end-of-life internal pressure.
The two-piece lower end fitting was replaced by a one-piece casting with a recess for the center guide tube.
The length of the center guide tube was increased from 163.715 inches to 163.965 inches to make it compatible with the redesigned lower end fitting.
The above design changes represent minor improvements which do not affect the fuel mechanical design basis.
The staff, therefore, finds these changes acceptable.
Also, based on previous staff reload evaluations, clad collapse analyses of new C-E manufactured fuel do not need to be performed because the time to clad collapse is in excess of any practical core residence time.
2.2 T~hl 2
The thermal performance of cycle 3 fuel was analyzed using the NRC-approved FATES3A code and composite fuel pins that envelope the pins of Batches B, C, D, and E.
A power history that enveloped the power and burnup levels of the peak pin at each burnup interval, from the beginning of cycle to the end of
- cycle, was used.
The maximum peak pin burnup analyzed bounds that expected at the end of Cycle 3.
Based on this analysis, the internal pressure in the most limiting fuel rod will stay below the nominal reactor coolant system (RCS) pressure of 2250 psi.
Because this satisfies Standard Review Plan (SRP)
Section 4.2 criteria, the thermal design of the Cycle 3 core is acceptable.
3.0 EVALUATION OF NUCLEAR DESIGN
- 3. 1 ~1N A general description of the Cycle 3 core is given in Section 1.0.
The Cycle 3
core uses a low-leakage fuel management scheme where previously burned Batch 8
assemblies are placed on the periphery and most of the fresh Batch E assemblies are located throughout the core interior in a pattern which minimizes power peaking.
The highest Batch E enrichment is 3.96 weight percent U-235; the PVNGS fuel storage facilities are approved for a maximum enrichment of 4.05 weight percent U-235.
Expected Cycle 3 lifetime is 390 EFPD.
A comparison of the Cycle 3 nominal characteristic physics parameters with those used in the safety analyses show that the latter are conservative in all cases.
3.2 Power Distribution Calculated "all-rod-out" relative assembly power densities have been presented for beginning of cycle (BOC), middle of cycle, and end of cycle (EOC).
Relative assembly power densities are also given at BOC and EOC for rodded configurations allowed by the power dependent insertion limit at full power.
These configurations consist of part length
The Cycle 3 nominal axial peaking factors are estimated to range from 1.22 to 1.08, at BOC and EOC, respectively.
Physics and power distribution calculations are based on the NRC-approved ROCS and NC codes employing DIT code generated neutron cross-sections.
The power distribution calculations are, therefore, acceptable.
3.3 Control Re uirements The value of the required shutdown margin varies throughout core life with the most restrictive value occurring at EOC hot zero power (HZP) conditions.
This minimum shutdown margin of 6.5 percent delta k/k is required to control the reactivity transient resulting from the RCS cooldown associated with a steam line break accident at these conditions.
For operating temperatures below 350'F, the reactivity transients resulting from any postulated accident are
minimal and a 4.0 percent delta k/k shutdown margin (revised from a value of 3.5 for Cycle 2) provides adequate protection.
Sufficient boration capability and net available CEA worth, including a minimum worth stuck CEA and approriate calculational uncertainties, exist to meet these shutdown margin requirements.
These results were derived by approved methods and incorporate appropriate assumptions and are, therefore, acceptable.
4.0 EVALUATION OF THERMAL-HYDRAULICDESIGN Steady-state thermal-hydraulic analysis for Cycle 3 is performed using the approved thermal-hydraulic code TORC and the CE-I critical heat flux (CHF) correlation.
The design thermal margin analysis is performed with the fast running variation of the TORC code, CETOP-D.
The CETOP-D model has been verified to predict minimum departure from nucleate boiling ratio (DNBR) conservatively relative to TORC.
The uncertainties associated with the system parameters are combined statistically using the NRC-approved modified statistical combination of uncertainties methodology.
Using this methodology, the engineering hot channel factors for heat flux, heat input, fuel rod pitch, and cladding diameter are combined statistically with other uncertainty factors to arrive at overall uncertainty penalty factors to be applied to the DNBR calculations performed by the core protection calculators (CPCs) and the Core Operating V
Limit Supervisory System (COLSS).
When used with the Cycle 3
DNBR limit of 1.24, these overall uncertainty penalty factors provide assurance with a 95/95 confidence/probability that the hottest fuel rod will not experience DNB.
The 1.24 value incorporates all applicab'le penalties, such as for rod bow, the 0.01 DNBR for HID-I grids, and the penalties specified in the statistical combination of uncertainties.
The rod bow value used in the analysis is 1.75 percent DNBR, for burnups up to 30,000 MWD/MTU. for burnups higher than 30,000 MWD/MTU sufficient margin exists to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblies and rods.
There-fore, the rod bow penalty is adequate for all anticipated burnups.
Because the thermal-hydraulic design analyses were performed using approved codes and took into account all applicable penalties, the staff finds these analyses acceptable.
5.0 EVALUATION OF NON-LOCA SAFETY ANALYSIS The design basis events (DBEs) considered in the safety analyses are catego-rized into two groups:
anticipated operational occurrences (AOOs) and postu-lated accidents
( limiting faults).
All events were reviewed by the licensee to assess the need for reanalysis as a result of the new core configuration for Cycle 3.
The DBEs were evaluated with respect to the following four criteria:
fuel performance (DNBR and centerline melt),
RCS pressure, loss of shutdown
- margin, and offsite dose.
The limiting fault events corresponding to each criterion were reanalyzed.
Plant response to the DBEs was simulated using the same methods and computer programs which were used and approved for the Cycle 2 analyses.
These include the CESEC III, STRIKIN-II, CETOP-D,
- TORC, and HERMITE computer programs.
For some of the reanalyzed DBEs, certain initial core parameters were assumed to be more limiting than the calculated Cycle 3 values in order to bound future cycles.
All of the events reanalyzed have results which are within NRC acceptance criteria and, therefore, are acceptable.
Two of the reanalyzed
- events, however, were not bounded by the Cycle 2 analyses.
These are the inadvertent opening of a steam generator safety valve or atmospheric dump valve (ADV) with loss of offsite power and the single reactor coolant pump shaft seizure/sheared shaft event with loss of offsite power and a single active failure of the ADV to close.
This single failure for the latter event maximizes the radiological consequences.
For the former event, the amount of predicted failed fuel increased from 8 percent to 12 percent as a result of more adverse nuclear power distributions.
However, the results of a radio-logical dose calculation were not presented.
The staff has requested a dose calculation be performed by the licensee which is currently underway and will be reviewed by the Radiation Protection Branch of NRR.
For the latter event, an increase in predicted fuel failure from 3.79 percent to 4.5 percent occurs.
The resulting radiological consequences are within 10 CFR 100 guidelines and therefore, meets the appropriate dose criteria and is acceptable.
6.0 EVALUATION OF ECCS ANALYSIS An ECCS analysis was performed for the limiting break size LOCA (a double-ended guillotine break with a 1.0 discharge coefficient) for Cycle 3 to demon-strate compliance with the requirements of 10 CFR 50.46.
The methodology is the same as for the cycle 2 analysis.
The analysis justifies a 13.5 kw/ft peak linear heat generation rate.
Because there have been no significant changes in hardware characteristics for Cycle 3, only fuel rod clad temperature and oxidation calculations were performed.
The code STRIKIN-II was used for this purpose and the fuel performance data were generated using the FATES-3A fuel evaluation code.
It was demonstrated that burnup with the highest initial fuel stored energy was limiting.
The ECCS analysis methods employed have been previously approved and are acceptable.
The results of the limiting break LOCA analysis for Cycle 3 are bounded by the results obtained in the Cycle 2 analysis, i.e.,
a peak clad temperature of 2091'F, a maximum local clad oxidation of 9.0 percent, and a core wide clad oxidation of less than 0.80 percent.
These values are within the 10 CFR 50.46 limits of 2200'F, 17.0 percent, and 1.0 percent, respectively, and are, therefore, acceptable.
Similarly, a review of Cycle 3 fuel and core data has confirmed that the small break LOCA analysis results are bounded by the Cycle 2 analysis.
7.0 TECHNICAL SPECIFICATION CHANGES TS Fi ure 3.1-1A The proposed change increases the required shutdown margin from 3.5 to 4.0 percent delta k/k for the RCS cold leg temperature range zero to 350'F when any full-length CEA is fully withdrawn.
The increased shutdown margin will ensure that the TS are consistent with the safety analyses performed for the Cycle 3 core and that the consequences of DBEs and anticipated operational occurrences are bounded by these analyses.
The proposed change is therefore acceptable.
TS Tables 3.1-2 3.1-3 and 3.1-5 These tables provide frequencies for monitoring RCS boron concentration in the event that one or both startup channel high neutron flux alarms are inoperable.
The proposed changes are more restrictive in that certain monitoring frequen-cies are increased to ensure that the TS are consistent with the safety analyses performed for the Cycle 3 core and that, in the event of an inadvert-ent boron dilution, sufficient time will be available to terminate the event prior to loss of shutdown margin.
The proposed changes are, therefore, acceptable.
TS Fi ures 3.1-3 and 3. 1-4 Figures 3.1-3 and 3. 1-4 provide regulating group CEA insertion limits when the COLSS is in service and out of service, respectively.
The proposed change to Figure 3.1-3 will prohibit insertion of regulating group 3
CEAs above 20 percent of rated thermal power.
This is permitted under the existing TS.
The proposed change to Figure 3. 1-4 will permit slightly increased insertion of regulating group 3
CEAs between 15 percent and 20 percent of rated thermal power.
The proposed revisions are necessary to ensure consistency of the TS with the safety analyses performed for the Cycle 3 core.
These analyses demonstrate that reactor operation in accordance with the revised insertion limits will ensure that the Specified Acceptable Fuel Design Limits (SAFDLs) will not be exceeded during the most limiting anticipated operational occurrence.
The proposed changes are, therefore, acceptable.
TS 3.2.7a TS 3.2.7a ensures that the actual value of the core average Axial Shape Index (ASI) remains within the range of values used in the safety analyses when the COLSS is operable.
The proposed change revises the limits of core average ASI from between -.28 to +.28 to between -.27 to +.27 to make the TS consistent
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with the safety analyses performed for the Cycle 3 core.
The proposed change is, therefore, acceptable.
'TS Fi ures 3.2-2 and 3.2-2A Figure 3.2-2 provides DNBR margin limits when at least one Control Element Assembly Calculator (CEAC) is operable and the COLSS is out of service.
Figure 3.2-2A provides the additional DNBR margin necessary when COLSS and both CEACs are out of service.
Reactor operation within these limits ensures that the SAFDLs will not be violated during an anticipated operational occur-rence.
The proposed changes are necessary to ensure consistency of the TS with the safety analyses performed for the Cycle 3 core and are, therefore, acceptable.
8.0 STARTUP TESTING The licensee has presented a brief description of the low power physics tests and the power ascension testing to be performed during Cycle 3 startup.
The described tests will verify that core performance is consistent with the engineering design and safety analyses.
If the acceptance criterion of any of the startup physics tests are not met, an evaluation will be performed by the licensee.
Resolution wi 11 be required prior to subsequent power escalation.
If an unreviewed safety question is involved, the NRC will be notified.
The staff has reviewed the proposed startup test program for Cycle 3 and finds that it conforms to accepted practices and adequately supplements normal surveillance tests which are required by the plant Technical Specifications.
9.0 EYALUATION FINDINGS The staff has reviewed the fuels, physics, and thermal-hydraulics information presented in the PVHGS3 Cycle 3 reload report.
Also reviewed were the Technical Specification revisions, the startup test procedures, and the
safety reanalyses.
Based on the evaluations given in the preceding sections, the staff finds the proposed reload acceptable.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and issuance of this amendment wi 11 not be inimical to the common defense and security or to the health and safety of the public.
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ENCLOSURE 2
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE FACILITY NAME Palo Verde Nuclear Generating Station Unit 3
SUMMARY
OF REVIEW This SER reviews a reload safety analysis report and associated proposed Technical Specifications (TS) changes submitted in support of a request to reload and operate PVNGS3 for a third cycle.
Based on our review, we find the proposed reload and TS revisions acceptable.
NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE - SAFETY ASSESSMENT/QUALITY VERIFICATION The licensee's submittals were representative of typical reload packages and provided adequate discussion of the Cycle 3 safety analyses and associated TS changes.
No additional information or clarifications were required from the licensee to enable our review.
AUTHOR:
L.
Ko DATE:
2/27/91