ML20072P903
| ML20072P903 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/15/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072P901 | List: |
| References | |
| NUDOCS 9011290211 | |
| Download: ML20072P903 (9) | |
Text
..-_ - _ -
fse ucy'%
- 4 UNITED STATES
_ h. ! ' ( i *$
NUCLEAR REGULATORY COMMISSION
- d'> g' f WASHINGTON, D.C. 30M6
%.....y SAFETY EVALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 73 TO FACILITY OPERATING LICENSE NO NPF-29 ENTERGY OPERATIONS. INC.
GRAND GULF NUCLEAR STATION. UNIT 1 DOCKET NO. 50-415
1.0 INTRODUCTION
By, letter dated June B,1990 (Ref.1), Entergy Operations, Inc. (the licensee), requested an amendment to Facility Operating 1.icense No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GG1). Enclosed were five attachments providing the requested Technical Specifications (TS) changes and reports discussing the reload and analyses to support and justify Cycle 5 operation, including two reports by Advanced Nuclear Fuels Corporation j
(ANF) (Refs 2 and 3).
~
By letter dated August 15,1990 (Ref. 4), the licensee submitted revisions to the original request which presented changes to several of the initially s
proposed TS and accompanying analyses. Enclosed were revised versions of the five original attachments, including the ANF reports (Refs. 5 and 6).
The revision addressed changes to ANF analyses as a result of a separate NRC review of the new CASM0/MICR0 BURN based methodology and the extended
. time required for NRC review of the new "TIP" uncertainty in that methodology.
This caused ANF to revert (with NRC permission) to the currently accepted (moreconservative)TIPvaluesinordertocomplete,ontime,theGG1 The primary result of the TIP chan Cycit 5 (GGICS) analyses.
increaseof0.01fortheMinimumCriticalPowerRatio(MCPR)gewasansafety limit and: corresponding changes to the operating limit MCPR and associated factors _ In addition (1) the slow flow excursion, and (2) the loss of feedwater heating events were reanalyzed. The first incorporated the increased safety limit in the analysis and the second used the revised (andapproved)MICR0BURNmethodology.
The Cycle 5 reload will replace 284 ANI 8x8 fuel assemblies used in Cycle 4 with ANF 9x9-5 fuel assemblies.' The core loading will retain 512 ANF 8x8 fuel assemblies and 4 lead test ANF 9x9-5 assemblies from Cycle 4.
The reload for Cycle 5 is generally a normal reload with no unusual features or characteristics other than the partial shift to a 9x9 loading pattern. ANF 9x9 fuel has been used in other reactors, and Susquehanna 2, for example, has been operating with an all ANF 9x9 fuel loading. The o
revised application did not-significantly alter the action previously 4
noticed or affect the initial no significant hazards consideration determination published in the Federal Register on July 25, 1990 (55FR30297).
f hal1290211 901113 nwm mowm,
=
g PDC l
.'..s L
2-
-l GG105 TS changes are not extensive and are primarily related to Average Planar Linear Heat Generation Rate (APLHGR), Linear Heat Generation R6te (LHGR), and Minimum Critical Power Retio (MCPR) limits and associated factors for Cycle 5 core operation as calculated by ANF. Several of these changes are the results of changes in the ANF methodologies. Some of this l
change has been indicated above as related to the CASM0/MICR0 BURN methodology and the NRC review. There h65 also been a change from the use of maximum APLHGR (MAPLHGR) multiplying factors (MAPFAC) to the use of LHGR multiplying factors (LHGRFAC)toprovideforfueldesignlimits. Since LHGR limits are monitored directly, MAFLHGR limits need consider only the requirements for LOCA analyses. Exposure dependent MCPR limits are introduced, in-addition to these therrael-hydraulic parameter changes, there is also a changetotwoRcdPatternControlSystem(RPCS)TSreducingthesetpoint for turning off the rod action control system (RACS) from 20 to 10 percent, as approved by the NRC staff in Amendment No.17 to GESTAR II (Ref. 21).
The new nethodologies used by ANF for GG1C5 involve the MCPR safety limit (Ref.12),theANFBcriticalpowercorrelation(Ref.13),theCASM0-3G/
MICRCEURN-B neutronic code (Ref.14), and a revised COTRANSA2 (Ref.15).
These methodologies have all been reviewed and approved by the NRC staff (Refs.10,16,17, and 18').
2.0 EVALUATION 2.1 Fuel Design The GGIC5 reload will include 284 new ANF 9x9-5 fuel assemblies. These centain 76 fuel rods and 5 water rods. The fuel rods are enriched to an 1
average of 3.42 without U-235 with eight to ten of the rods containing gadolinia as burnable poison. The fuel design and safety analysis are described in the GG1CS Reload Sumary Report (Ref 7 and in Refs. 6 and 8). The methodologies and the application to ANF 9x9-5 have been reviewed and approved by the NRC staff. The fuel mechanical design is similar to ANF'9x9 fuel approved for use in other BWRs, e.g., Susquehanna-2 which currently has a complete ANF 9x9 core loading. The design analyses, using approved methodologies, were performed to support assedly discharge
~
burnups.of 39 GWa/MTV fur the remaining 8x8 assemblies and 40 GWd/MTU for.
the 9x9-5 assemblies. The fuel chnhels to be-used for ANF 9x9-5 fuel are manufactured by Carpenter Technology Corporation and are of a similarAll design and are equivalent to GE channels used in previous cycles.
channels, including those from previous cycles, to be used in GGICS, and in future cycles, are being used for only a single bundle lifetime.
MCPR affects from channel bowing have been included.in the safety analyses.
ANF ha3 analyzed the response of the ANF 9x9-5 fuel assemblies during seismic-LOCA events (Ref. 6, Appendix A) by comparison of characteristics to a3 proved 8x8 fuel licensed to operate in GG1.
Because of the similarity of tie dynamic and hydraulic characteristics of the fuel assemblies and channel boxes, the 9x9-5 fuel will have essentially the same static and dynamic response.
-3 Based N. our review of the information presented, and the similarities to previouslu.p;. roved cesigns and analyses, we find the mechanical design of the ANT 9x9-5 fuel for GGICS to be acceptable.
2.2 Nuclear Design The ISF nuclear design methodology is presented in References 9 and 14 The latter provides new methodology for nuclear design er,alysis and t6s recently been reviewed and approved by the NRC staff (Ref.10).
The beginning of cycle (BOC) shutdown margin is calculated to be 1.06 percent delta-K, and BOC is the most limiting condition. Thus, the cycle minimum shutdown margin is well in excess of the required 0.38 percent delta-K. The standby Liquid Control System also fully meets shutdown requirements. The GG1 high density spent fuel storage racks have been reviewed separately for the acceptability of storing fuel from the Cycle 5 reload,anditwasconcluded(Ref.11)thatthestoragerackscansafety accommodate the Cycle 5 fuel.
The GGICS nuclear characteristics have been calculated with approved methodologies, the results are reasonable and fall within expected ranges and the review concludes that the design is acceptable.
2.3 Thermal.,:Jraulic Design The retained ANT 8x8 fuel and the new 9x9 fuel are thermal-hydraulically compatible as determined by approved methodologies and the combination has been approved for use in previous BWR reloads, e.g., Susquehanna 1 and 2.
The use of the ANF 9x9 fuel in GG105 is' acceptable from a thermal-hydraulic viewpoint.
The ANF thermal-hydraulic methodology and criteria used for GG1CS design and analysis is for the most part the same as used for previous GG1
-reloads. This is described in References 19 and 20. New aspects of the methodology were-introduced in this reload, however. They are described in the ANF topical reports presented in References 12, 13, 14 and 15.
'They involve the MCPR safety limit, the ANFB critical power correlation,lic the CASMO-3G/MICR0 BURN-B neutronics code, and the revised themal-hydrau code C0iRAnSA2. These methodology changes have all been recently reviewed by the NRC staff and approved. The safety evaluation for these reports is presented in References 10, 16, 17, and 18. The methodologies used, including the approved changes, are acceptable to the NRC staff for GG1C5 analysis.
Changes were also made to the format for presenting MAPLNGR and LHGR limits. Formerly, the MAPLHGR limits had multipliers (MAPFAC) for off-rated conditions to provide LHGR protection et these conditions.
In the revised format,thisfunctionhasbeentransfetredtoLHGRmultiplier(LHGRFACThe basic andLHGRFAC).
p
,4 i
transfer. The MAPLHGR limit for ANF 8x8 fuel has been determined so as to i
cover (11 ANF 8x8 fuel types in Cycle 5, and similarly the MAPLHGR for ANF 9x9 covers all gx9 fuel types.
Also, the LHGR limits and multipliers have l
been determined to be applicable to all fuel types in the cycle. These changes in the LHCR and MAPLHGR format provide similar limit protection as the previously approved format and are acceptable.
The LHGR limits for the 8x8 fuel have been extendtd to include the expected end of Cycle 5 burnup. This extension falls within approved methodology (Ref.3)andisacceptable.
The MCPR and LHGRFAC limits have been extended below 30 percent core flow.
Thiscoversbothtwoloopandsingleloopoperation(TLOandSLO). This has been justified by the analysis of the flow run out events, including consideration of both Loop Manual and Nun-Loop Manual modes (corresponding to singit and dual recirculation loop flow excursions).
The analysis used approved methodology and is acceptable, j
The MCPR safety limit has been determined to be 1.09 for both TLO and SLO using approved methodology. As discussed previously in this report, a conservative value for t$e 'TIP* uncertainty f act:a was used as the result of the staff review of the methodology. ANF calculated MCPR optrating limits at several cycle exposures and provided an exposure dependent MCPR operating limit as well as the usual power and flow dependent MCPR limits (liCPR and MCPR The MCPR operating limits are based on analyses of planttransient[)tobediscussedlater. The development of these limits p
follows approved methodology and is acceptable.
The effect of channel bowing was included in.the MCPR analysis by ANF.
Channel use will involve only single bundle lifetime. ANT methodology for single bundle lifetime MCPR analysis b s been reviewed by the NRC staff, and is acceptable for channel bow analysis of GGICS.
GG1 is currently opersting under the Interim Recomendations of Stability Actions with a TS approved by the NRC staff in a previous review of GG1 thermal-hydraulic stability. The boundaries of the TS designated operating regions were based on the interim recomendation boundaries. ANF has performed calculations of the stability characteristics of GG1 for Cycle 5, which will contain about a third of a core of ANF 9x9 fuel. The analysis indicated that the decay ratio has not changed significantly from Cycle 4 to 5 for equivalent conditions. Messurements by NRC censultants in the Susquehanna 2 reactor, with various amounts of ANF 9x9 fuel from succeeding reloads, including all 9x9 fuel, indicated no significant deterioration of the decay ratio. The NRC staff review thus concludes that continued use of the current stability TS boundaries is acceptable.
P.4 Anticip6ted Operational Occurrences and Accident Analyses To provide the basis for the TS values of the various operating limits (MCPR, LHGR, and MAPLHGR), ANF has analyzed the system Anticipated Operational Occurrence (A00) events which could provide the most limiting l
l
IIIW 3
conditions. This follows the normal pattern of the approved methodology for o erating limit analysis. This included Load Rejection Without Bypass (LRNB, Feedwater Controller Failure (FWCF), loss of Feedwater Heating (LFWH, Flow Excursion (FE), Control Rod Withdrawal Error (CRWE) and the FuelLoadingError(FLE). Previous analyses have shown that other events are non lin,iting. Plant initial conditions for the analyses covered the full range of Maximum Extended Operating Domain (ME00) approved for GG1.
Analyses were done for End of Cycle (E00) E00-2000 mwd /HTU and E0C+30 EFPD (Effective Full Power Osys) to provide burnup dependent MCPR lin.its.
Results from these analyses were used to provide the TS MCPR and LHM limits as a function of power, flow, and exposure.
The change to the 'TIP' uncertainty, discussed previously, taquired reanalysis of the slow flow excursion and LFWH events. The f ormer was rerun with the new safety limit MCPR of 1.09 and a complete new analysis was run for the LFWH. The LFWH was analyzed with the newly approved MCR0 BURN.B/ANFB following the approach previously approved for Cycle 4, using an expanded GG1 data base.
The analysis of A00 events and the develepment of limiting operating values for MCPR and LHGR used approved methods and considered required events and reactor conditions. The analysis and results are acceptable.
ANFalsoanalyzedoperationvndersingleloopoperation(SLO). The analyses included calculation of the por seizure event and determination of MCPR and MAPLHGR limits.
It was det m ined that the MCPR safety limit of 1.09 was applicable to SLO and the seLHGR limit reduction factor, from LOCA analyses, should be 0.B0. The analyses were done with approved methods and are acceptable.
Compliance with overpressurization criteria was demonstrated by analysis ofthemainsteamisolationvalve(MSIV)closureeventassumingMSIV position switch scram f ailure. The analyses used conservative parameters and resulted in pressure under required limits. The analysis used approved methods and is acceptable.
ANFanalyzedtheLossofCoolantAccident(LOCA)andRodDropAccident (RDA) and determined that required limits are met for GGICS. The analyses used approved methods and are acceptable.
2.5 Technical Soecification Chances The following Technical Specification (TS) changes have been proposed for operation of GG1C5.
(1) Definition 1.8 - Administrative change of "XN-3" to 'ANFB' to reference correct current correlation. The change is acceptable.
(2) 15 2.12 -- The MCPR safety limit is increased to 1.09 for both TLO and SLO. This has been determined with approved methoth for the fuel in GGICS and is acceptable.
1
.., 9 '., '
(3) TS 3/4.1.4.2
'te power level above which the control rods may be bypassed in the o Action Control System (RACS) is reduced from 20 l
to 10 percent, 215 setpoint level is related to the control Rod i
Drop Accident MA) analysis. The problem area was reviewed generically by the NRC staff in connection with hendment No. 17 to GESTAR11(Ref.21),andpermissiontolowerthesetprintto10 percent received generic approval.
It is acceptable '..r GGl.
l (4) TS 3/4.2.1 - The changes delete references to fuel type specific i
MAPLHGR curves, and delete references to MAPFAC curves. As previously discussed, the MAPFAC limits have been transferred to LHGR limits and theanalysisofMAPLHGRlimitsforLOCA(only)havecoveredANF8x8 and AFN 9x9 fuel generically, b,ese changes are acceptable. ANF has also determined that the $LO MAPLHGR multiplier is now 0.80. This change is also acceptable.
Figere 3.2.1-1 is changed to show the new ANF 8x8 and 9x9 MAPLHGR values and Figures 3.2.1 16 through 3.2.1 le, 3.2.1 2, and 3.2.1 3 are deleted because, as indicated above, the values are no longer used. This is acceptable.
(5) TS 3/4.2.3 Reference is now made to exposure dependent MCPR limits and minor administrative changes are made. These are acceptable.
Changes are made to Figures 3.2.3 1 and 3.2.3 3 to reflect the changes to MCPR limits for the cycle which have been discussed previously. These changes are acceptable.
(6) TS 3/4.2.4 -- The text is changed to reflect the additios of the LHGRFAC multipliers as has been previously discussed.
Figures 3.2.41, 3.2.4 2, and 3.2.4-3 have been revised or added to reflect these changesandtheextendedANFLHGRcurve(alsopreviouslydiscussed).
The changes are acceptable.
(7) TS 3/4.4.1.1 -- Reference to TS 2.1.2 is deleted since the changes to T$ 2.1.2, discussed above, make reference to this TS unnecess;;ry.
This is acceptable.
(8) TS 3/4.10.2 -- As was discussed above for TS 3/4.1.4.2, the setpoint for the RAC$ is reduced from 20 to 10 percent. The change is also acceptable for this specification.
i There are also changes to the Bases associated with the above TS to reflect the changes to the specifications or minor administrative changes.
The changes suitably reflect the basis for the changes and are acceptable.
These include Bases 2.1.1 and 2.1.2; 3/4.1.4 and 3/4.1.5; 3/4.2.1, 3/4.2.3, and 3/4.2.4; and 3/4.4.1.
2.6 $UMMARY The NRC staff has reviewed the reports submitted for the Cycle 5 operation of Grand Gulf Unit 1 and concludes that appropriate material was submitted and and that the fuel design, nuclear design, thermal-hydraulic design,ification transient and accident analyses are acceptable. The Technical spec changes submitted for this reload suitably reflect the necessary modifications for operation in this cycle.
l l
O s,,
J.
2.7 References 1.
Lrtter from W. T. Cottle, Entergy Operations, Inc. (E01) to NRC,
" Cycle 5 Reload," dated June 8, 1990.
2.
ANF-90 21, Revision 1, ' Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis," dated May 23, 1990.
3.
ANF 90 21, Revision 1. ' Grand Gulf Unit 1 Cycle $ Plant Transient Analysis," dated May 24,1990.
4.
Letter from W. T. Cottle, E01, to NRC, " Cycle 5 Reload," dated August 15, 1990.
5.
ANF-90-22. Revision 2. ' Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis," dated August 8, 1990.
6.
ANF 90-21, Revision 2, ' Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis,,' dated Aug:sst 8,1990.
- 7. to AECH40/0146, ' Grand Gulf Nuclear Station Unit 1 Cycle 5 Reload Summary Report,' dated August 1990.
" Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload B.
Fuel," XN-NF-85 67(P)(A), Revision 1, Exxon Nuclear Company, Richland, Washington, September 1986.
" Generic Mechanical Design for Advanced Nuclear Fuels 9x9-5 BWR ReloadFuel*ANF-88-152(P). Amendment 1. September 1989, Advanced NuclearFuelsCorporation,Richland, Washington.
9.
XNNF-8019(A), Volume 1. Supplements 1&2,'ExxonNuclearMethodology' for Boiling Water Reactors: Neutronics Methods for Design and Analysis, Exxon Nuclear Co., March 1983.
Letter to R. Copeland ANF, from A. Thadani NRC ' Acceptance for 10.
ReferencingofTopicalReportXNNT-80-19(p}, Volume 1, Supplement 3
' Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for the CASMO-3G/MICR06 URN B Calculation Methodology, dated Au9ust 13, 1990.
- 11. Letter to W. T. Cottle, E01, from L. Kintner,'NRC, " Criticality Analysis for Cycle 5 Fuel in Spent Fuel Storage Racks, dated July 16,1990.
Revision 2 ' Advanced Nuclear Fuels Corporation Critical XHNF524(P),logyforBollingWaterReactors.*includingSupplements.
12.
Power Methodo Advanced Nuclear Fuels Corporation, April 1989.
ANF-1125(P)lsCorporation, April 1989. Supplement 1, 'ANFB Criti 13.
Nuclear Fue l
-. ~. - -. -. -. -.. - - - -. - _-..-
o e p *.
8-
\\
Volume 1, Supplement 3, 'ANF Methodology for BWRs:
i XNNF-6019(P)ltsfortheCASM0-3G/M1CROBURN-BCalculationMethodology,*
14.
Benchmark Resu Advanced Nuclear Fuels Corporation, February 1989, as supplemented by ANF letter RAC:083:90 dated July 20, 1990.
Volume 1, Supplements 1, 2, and 3, 'COTRANSA2: A Computer ANF-913,for Boiling Fater Reactor Transient Analysis,' Advanced Nuclear 15.
1 Program Fuels Corporation, June 1989.
Letter to R. Copeland ANF, from A. Thadani, NRC, ' Acceptance for 16.
ReferencingofTopicalReportANF-524(P Revision 2, 'ANF Critical Power MethodologyforBoilingWaterReactors.)*datedAugust8,1990.
- 17. Letter to R. Copeland, ANF, from A. Thadani, NRC, ' Acceptance for Referencing of Topical Report ANF-1125(P) and Supplement 1
'ANFB Critical Power Correlation, dated March 8, 1990.
Letter to R. Copeland, ANF, from A. Thadani, NRC,'COTRANSA2:' Acceptan 18.
Referencing of Licensing Topical Report ANF-913 A Computer Program for Boilin9, Water Reactor Transient Analyses (TAC No. 68356),
i l
dated May 23, 1990, l
- 19. XNNF-80-19(P)(A),Volums4. Revision 1.'ExxonNuclearMethodologyfor l
Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Co., June 1986.
I
- 20. XN-NF-80-19(P)(!,), Voivme 4 kevii,ian 2. ' Exxon Nuclear Methodology for Boiling Water Reactors THERMF.X: Thermal Limits Methodology Sumary Description,' Exxon Nuclear Co., January 1987.
- 21. NEDE-24011-P-A 9, " General Electric Standard Application for Reactor Fuel' (GESTAR !!), September 1988.
l I
i s
ss d
3.0 ENVit0NMENTAL CONSIDERATION Pursuant to 10 CFR 51.21 51.32, and 51.35, an Environmental Assessment onOctober2,1990(gnifIcantImpactwaspublishedintheFederalReeis and Firding of No Si l
Accordingly, based on the Environmental Assessment, the Comission has determined that issuance of this amendment will not have a significant l
effect on the quality of the human environment.
4.0 CONCLUSION
l The Commission a.ede a proposed determination that this amendn.ent involves l
no significent hazards consideration, which was published in the Federal Re 25,1990(55FR30297),andconsultedwiththeStateof
%gister on JulyNo public comments or requests for hearing were received, sissippi.
and the State of Mississippi did not have any corsnents.
~
i The staff has concluded, based on the considerations discussed above,f the l
that:
(1) there is reasonable assurance that the health and safety o
^
public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in ccepliance with the Corrission's regulations, and the issuance of this amendment will not be inimical to the comon defense and the security, or to the health and safety of the public.
l Dated: November 15, 1990 1
Principal Contributor:
H. kichings t