GO2-90-075, Application for Amend to License NPF-21,supporting Cycle 6 Reload & Operation of Lead Fuel Assemblies During Feedwater Temp Reduction Anticipated & Responds to NRC Bulletin 90-002 Re Loss of Thermal Margin Caused by Channel Box Bow
| ML17285B198 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/13/1990 |
| From: | Sorensen G Washington Public Power Supply System |
| To: | NRC/IRM |
| Shared Package | |
| ML17285B199 | List: |
| References | |
| GO2-90-075, GO2-90-75, IEB-90-002, IEB-90-2, NUDOCS 9004230573 | |
| Download: ML17285B198 (98) | |
Text
UTION DEMONKQAYION SYSIKM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR: 9004230573 DOC. DATE: 90/04/13 NOTARIZED: YES FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe AUTH.NAME AUTHOR AFFILIATION SORENSEN,G.C.
Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET 05000397
SUBJECT:
Forwards Tech Spec changes to support SVEA-96 lead FA during FFTR anticipated during Cycle 6 to 900222 amend application.
DISTRIBUTION CODE:
IE38D COPIES RECEIVED:LTR Q ENCL g SIZE:
TITLE: NRC Bulletin 90-002, Loss of Thermal Margin Caused by Channel B x Bow NOTES:
RECIPIENT ID CODE/NAME PD5 LA INTERNAL: AEOD/DOA NRR/DET/EMEB9H3 NRR/DOEA/OGCB11 NRR/DST/
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1 Pg gg,flSik~n Zgg ~Spy HaV NOTE TO ALL"RIDS" RECIPIEÃIS:
PLEASE HELP US TO REDUCE WASTETH CONTACT'BiE DOCUMEMI'ONIROLDESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAMEFROM DISIRIBVHON LISTS FOR DOCUMENIS YOU DON'T NEEDI lq Ro TOTAL NUMBER OF COPIES REQUIRED: LTTR ~
ENCL
WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968
~ 3000 George Washington Way
~ Richland, Washington 99352 April 13, 1990 G02"90-075 Docket No. 50-397 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Mail Station Pl-137 Washington, D.C.
20555
Subject:
NUCLEAR PLANT NO. 2 OPERATING LICENSE NPF-21 MODIFICATION TO THE WNP-2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRC BULLETIN NO. 90"02:
LOSS OF THERMAL MARGIN CAUSED BY CHANNEL BOX BOW
Reference:
- 1. G02-90-032, Letter Attachments:
"Nuclear Plant No.
2, Operating License NPF-21, Request For Amendment to Technical Specifications Reload License Amendment (Cycle 6)"",
dated February 22, 1990
- 2. Technical Report No.
WPPSS-EANF-126, Rev.
1, "WNP-2 Cycle 6
Reload Summary Report", April 1990 3.
NRC Bulletin No.
90-02:
Loss of Thermal Margin Caused by Channel Box Bow, 03/20/90.
4.
DP Chan and DL Larkin, "Finite Element Analysis of Boiling Water Reactor Fuel Channel Bulge and Bow", Nuclear Technology, March 1987.
- 5. Letter to Robert C.
- Jones, NRC from R.
A. Copeland, ANF, "Loss of Thermal Margin Caused by Channel Box Bow", RAC: 030:90, April 9; 1990
- 6. ANF-524(P),
Revision 2,
Supplement 1,
"Critical Power Methodology for Boiling Water Reactors Methodology for Analysis of Assembly Channel Bowing Effects",
- November, 1989.
The purpose of this letter is to provide the NRC with information regarding:
1) changes to the WNP-2 Cycle 6
reload submittal provided in Reference 1;
2)
Technical Specification changes to support the operation of the SYEA-96 Lead Fuel Assemblies during Feedwater Temperature Reduction (FFTR) anticipated in Cycle 6;
and 3) the requirements of NRC Bulletin No. 90-02, Loss of Thermal Margin Caused by Channel Box Bow.
It is requested that the NRC review this information in conjunction with the WNP-2 Cycle 6 reload license submittal provided in Reference 1.
9004230'=
900413 PDFl ADOGI< 05000397 P
Page Two MDIFICATION TO THE MNP"2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRG BULLETIN NO. 90-02 CYCLE 6 REVISED DESIGN In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, the Supply System in Reference 1 requested an amendment to the WNP-2 Technical Specifications.
This amendment was submitted to allow the use of Cycle 6 reload fuel in WNP-2.
Subsequent to the design, analysis and submittal of that request the continued superior performance of WNP-2 and the energy requirements of the region have made it necessary for the Supply System to modify its design of the WNP-2 Cycle 6 reload by the addition of eight ANF-5 reload fuel assemblies and the discharge of eight XN-1 reload assemblies.
The eight additional reload fuel assemblies are identical to the ANF-5 reload assemblies described in Reference 1.
" A modified Cycle 6 reload map, Attachment I, displaying the updated WNP-2 Cycle 6 core and WNP-2 Cycle 6 Reload Summary
- Report, WPPSS-EANF-126, Rev.
1, Attachment II, are provided for information to assist the NRG in completing the necessary reviews requested in Reference 1.
The Supply System has reviewed the use of the modified Cycle 6 reload design in WNP-2 and has concluded that it does not involve an unreviewed safety question.
Analysis of the modified design demonstrates that the proposed Technical Specification changes described in Reference 1 are applicable to the modified design (Reference 2).
The Supply System has also evaluated this r equest per 10CFR50.92 and determined that it does not:
Involve a significant increase in the probability or consequences of an accident previously evaluated.
An analysis using NRG approved methodology has been performed on the modified Cycle 6 reload design to examine the probability or the consequences of an accident or safety related equipment malfunction and the analysis demonstrates no significant change in previously evaluated accidents.
The mechanical, thermal hydraulic, and neutronic characteristics of the reload bundles have been analyzed and in all cases the evaluation of those changes shows that the design complies with established
- criteria, as approved by the NRC.
The results of those analyses are consistent with previous results and have not resulted in a significant reduction in margin of safety (see 3 below).
2.
Create the possibility of a
new or different kind of accident from any accident previously evaluated.
The reload fuel has been analyzed in detail and has been found to be sufficiently similar to the previous reload fuel whose analysis has been reported in the FSAR to preclude the possibility that an accident or malfunction of a different type than that previously analyzed is credible.
These analyses provide assurance that the proposed fuel loading design does not effect previous analyses bases.
Page Three MODIFICATION TO THE WNP"2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRC BULLETIN NO. 90-02 3.
Create a significant reduction in the margin of safety.
The Cycle 6
reload design was the subject of a
thorough analysis with NRC approved methodology, the intent of which was to examine the applicability of the WNP-2 Technical Specifications to the WNP-2 core.
These 'nalyses confirmed some of the existing operating limits and recommended changes in
- others, thereby setting thermal limits for WNP-2 specific to the Cycle 6 core (Reference 1).
Analysis of the modified reload design with NRC approved methodology demonstrates that the pr oposed Technical specification changes described in Reference 1 are equally applicable to the modified design (Reference 2).
With operation constrained by this set of thermal limits there is no reduction in safety margin for, operation of WNP"2.
As discussed
- above, the Supply System considers that this reload design modification does not involve a significant hazard consideration; nor is there a potential for significant change in the types, or. significant increase in the amount of any effluents that may be released offsite; nor does it involve a
significant increase in individual or cumulative occupational r adiation exposure.
Accordingly',
the proposed modification meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22(c)(9) and therefore, per 10CFR51.22(b),
an environmental assessment of the change is not required.
FFTR OPERATING LIMITS FOR SVEA-96 LFAs The recent NRC approval of WNP-2 Technical Specification changes to allow for operation of WNP-2 with Final Feedwater Temperature Reduction (FFTR) requires that the Supply System request an amendment 'o the WNP-2 Technical Specifications to establish an MCPR FFTR operating limit for the ABB SVEA-96 Lead Fuel Assembly (LFA) reload fuel included in the WNP-2 Cycle 6 core.
This Technical Specification change could not be included in the refer enced reload submittal due to uncertainty as to the date of approval and form of the FFTR related Technical Specification changes that would be approved by the NRC.
Included with this letter. as Attachment III is a marked up copy of the relevant Technical Specification identifying the changes proposed to include the SVEA-96 FFTR operating limit for NRC review and approval.
The Supply System has reviewed the proposed Technical Specification change of the MCPR operating limit for SVEA-96 LFA fuel operating in the FFTR condition and has concluded that it does not involve an unreviewed safety question.
The Supply System has also evaluated this request per 10CFR50.92 and determined that it does not:
Page Four MODIFICATION TO THE WNP"2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRG BULLETIN NO. 90-02 1.
Involve a significant 'increase in the probability or. consequences of an accident previously evaluated.
FFTR operation in WNP-2 has been evaluated with NRG appr oved methodology.
This evaluation, which has been approved, assures that the effects of FFTR operation on WNP-2 thermal limits are taken into account.
The methodology for developing SYEA-96 LFA thermal limits is thoroughly discussed, in the referenced submittal (Reference 1).
This methodology was rigorously applied in this instance.
2.
Create the possibility of a
new or, different kind of accident from any previously evaluated.
All FFTR impacts have been the subject of previously approved analyses performed with NRC approved methods.
This proposed Technical Specification change extends that analysis to this specific fuel type.
Operation with SVEA-96 LFA fuel in WNP-2 in FFTR conditions is sufficiently similar to previous reload analyses to preclude the possibility that an accident or malfunction of a different type than that previously analyzed is credible.
3.
Create a significant reduction in the margin of safety.
Analyses of FFTR operation in WNP-2 have been approved based on no significant reduction in the margin of safety.
The use of SVEA-96 LFA fuel has been thoroughly analyzed.
In particular; the methodology for developing thermal limits for SVEA-96 LFA fuel in WNP-2 has been extensively evaluated and rigorously applied in developing the proposed Technical Specification change.
The proposed MCPR change as applied to SYEA-96 LFA fuel", is found to be conser vative (i.e.,
50% larger) when compared to already approved NCPR FFTR limits, especially when the fact that the WNP-2 SYEA-96 LFA fuel is designed to match the normal WNP-2 reload fuel is recognized.
With operation to.the proposed NCPR limit under FFTR conditions, there is no reduction in 'safety margins for operation in WNP-2.
The Supply System considers that this change does not involve a significant hazard consideration, nor, is there a potential for significant change in the types or. significant increase in the amount of any effluents that may be released
- offsite, nor.
does it involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22(c)(9) and therefore; per 10CFR51.22(b);
an environmental assessment of the change is not required.
The above Technical Specification changes have been reviewed and approved by the WNP-2 Plant Operations Committee (POG) and reviewed by the Supply System Corporate Nuclear, Safety Review Board (GNSRB).
Page Five MODIFICATION TO THE WNP-2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRG BULLETIN NO. 90%2
RESPONSE
TO NRC BULLETIN NO. 90-02 The Supply System has received NRG Bulletin No. 90-02 which discusses the loss of thermal margin caused by channel box bow (Reference 3).
The purpose of this bulletin is to request that addressees determine whether any channel boxes are being reused after. their, first bundle lifetime and; if so; ensure that the e'ffects of channel box bow on the critical power ratio calculation are properly taken into account.
It is not the intention of the Supply System to use channel boxes in HNP-2 for, two bundle lifetimes.
The Supply System is aware of the dryout event which occurred at a foreign BWR facility and is therefore sensitive to the potential problems associated with excessive channel box bow.
The Supply System has in the past re-inserted re-qualified channels
- but, based in part on the above information, began a program about one year ago to transition away fr om channel reuse.
At the beginning of Cycle 5;
254 channel boxes had been reinserted into the HNP-2 core.
At the beginning of Cycle 6;
351 channels will have been re-inserted in the core.
Channels which are re-inserted in WNP-2 are first subject to inspection and physical measurement.
Channels are found to be acceptable if their physical dimensions and associated neutron induced distortion are found to be within a
pre-determined acceptance criteria.
'Approximately 80K of the channels measured have been qualified for re-use.
The acceptance criteria are developed based on a
bounding analytical predictive model of channel distortion with irradiation (Reference 4).
Channels re-inser ted in WNP-2 were channels discharged from WNP-2 Cycles 1; 2, 3 and 4.
Three hundred thirty-one (331) are initial core GE channels.
Twenty (20) are CARTECH channels inserted in Cycle 2 to replace GE channels fabricated from mismatched halves and then discharged at the end of Cycle 2.
There are no channels in the core manufactur ed from mismatched halves.
For HNP-2; a fuel assembly lifetime is approximately 5 to 6 fuel cycles.
Therefore, for Cycle 5
none of the re-inserted channels in HNP-2 will exceed a fuel bundle lifetime and the highest exposure channel at the end of Cycle 5 will have an exposure of about 30,500 NWD/NTU.
At the end of Cycle 6 it is projected the highest exposed re-inserted channels will have an exposure of about 37;800 NWD/NTU.
The following table indicates the number. of channels that have been or. will be reinserted for. each cycle.
Channels re-inserted in Cycles 3 and 4 constitute the highest exposure channels during Cycle 6.
Page Six HODIFICATION TO THE XNP-2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRC BULLETIN NO. 90"02 TABLE 1 Reinserted Channel Exposure Information CYCLE RELOAD BATCH SIZE NO.
CHANNELS REINSERTED EXPOSURE AT TINE OF REIN" EST.
EXPOSURE SERTION GWD/NTU AT EOC6 GWD/NTU 3
5 6*
148 152 136 152 56 57 147 50 41 56 1.7 " 9.5 2.6 " 14.6 1.7'- 20.8 14.4 21.6 20.8 - 26.4 23.3 " 37.8 23.9 " 37.8 16.4 - 37.5 22.2 29.4 25.3 30.9
- Fifty-six channels were placed on 56 original core GE assemblies that are being reinserted on the core periphery.
In order to assess the impact of channel box bow on the thermal limits proposed for Cycle 6, Advanced Nuclear. Fuels; Inc.
(ANF) has performed and submitted an analysis to the NRC for review (Reference 5).
This analysis concludes that for the range of channel box exposure expected during Cycle 6 operation, sufficient conservatism exists in the ANF XN-3 critical heat flux correlation (used in the Cycle 6
Transient Analysis; Refer ence
- 1) to offset the delta critical power ratio impacts that will result from channel box bow.
Consequently, the Supply System does not propose to recommend any changes to the Cycle 6 thermal limits proposed in Reference 1 to account for channel box bow; It is the intention of the Supply System to transition under our channel management program to use channels for only a single assembly lifetime and apply the methodology described in Reference 6 for the analysis of futur e cycles.
Our projections conclude that although no channel will be used for, two assembly lifetimes; it is estimated that subsequent to Cycle 6 operation, some channels may enter the 50 - 60 GWD/NTU burn-up range discussed in NRC Bulletin 90-02.
The Supply System will provide the NRC further information on our program for. transitioning to using channel boxes for, only a single assembly lifetime.
An additional response addressing subsequent cycles will be provided by October. 1, 1990 which will complete our response to the bulletin.
Page Seven MODIFICATION TO THE WNP-2 CYCLE 6 RELOAD SUBMITTAL AND RESPONSE TO NRC BULLETIN NO. 90-02 WNP-2 is scheduled to begin the spring outage on April 20; 1990.
The Plant is currently scheduled to resume operation on or about June 2; 1990.
Approval of this Technical Specification ammendment is required prior to Plant restart.
Very Truly Yours, G.
. Sorensen Manager.; Regulatory Programs Attachments:
I.
Revised Cycle 6 Reload Map II.
Cycle 6 Reload Summary Report WPPSS-EANF-126, Rev.
1 III. Technical Specification for, SVEA-96 FFTR Operating Limit WCW:bw cc:
C. Eschels, EFSEC JD Martin, US NRC NS Reynolds, BCP8R RB Samworth; US NRC DL Williams, BPA NRC Site Inspector
STATE OF WASHINGTON)
)
COUNTY OF BENTON
)
Subject:
Modificati on to the WNP-2 yce eoa umi a
and Response to NRC Bulletin No. 90-02 I,
G.
C.
- Sorensen, being duly sworn, subscribe to and say that I am the
- Manager, Regulatory
- Programs, for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that I have full authority to execute this oath; that I
have reviewed the foregoing; and that to the best of my knowledge, information, and belief the statements made in it are true.
DATE:
~~ Arfrd-, 1990 G.
C.
- orensen, Manager Regul tory Programs On this day personally appeared before me G.
C.
- Sorensen, to me known to be the individual who executed the foregoing instrument, and acknowledged that he signed the same as his free act and deed for the uses and pur'poses herein mentioned.
yh GIVEN under my hand and seal this ~l day of
,'990.
STATE OF WASHINGTON Residing at
,- /
A'y commission expires 1
91
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WNP-2 CYCLE 6 RELOAD
SUMMARY
REPORT PREPARED BY:
W.C.
WOLKENHAUER, PRINCIPAL ENGINEER, NUCLEAR FUEL CONCUR WITH:
R..
- VOSBURGH, AGER, SAFETY AND RELIABILITYANALYSIS CONCUR WITH:
R.J.
T L ERT, SUPERVISOR, WNP-2 REACTOR ENGINEERING CONCUR WITH:
R.L.
- KOENIGS, MANAGER, WNP-2 TECHNICAL APPROVED BY:
D.L.
WHITCOMB, MANAGER, NUCLEAR FUEL
NOTICE This report is derived in part through information provided to Washington Public Power Supply System (Supply System) by Advanced Nuclear Fuels Corpora-tion, General Electric Company and ABB Atom, Inc. It is being submitted by the Supply System to the U.S. Nuclear Regulatory Commission in support of the-WNP-2 Application for Technical Specifications Changes Relating to WNP-2 Cycle 6 Operation.
The information contained herein is true and correct to the best of the Supply System's knowledge, information, and belief.
NNP-2 CYCLE 6 RELOAD
SUMMARY
REPORT TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 GENERAL DESCRIPTION OF RELOAD SCOPE.
2.1 Summary of Results.
3.0 HNP-2 CYCLE 5 OPERATING HISTORY 4,0 RELOAD CORE DESCRIPTION.
5.0 FUEL MECHANICAL DESIGN <8x8C FUEL) 6.0 LEAD FUEL ASSEMBLY PROGRAM.
7.0 THERMAL HYDRAULIC DESIGN 7.1 Hydraulic Compatability 7.2 Fuel Cladding Integrity Safety Limit.
7.3 Fuel Centerline Temperature.
7.4 Bypass Flow Characteristics.
7.5 Thermal Hydraulic Stability.
8.0 NUCLEAR DESIGN.
8.1 Fuel Bundle Nuclear Design 8.2 Core Nuclear Design 8.3 Comparison of Major Core Parameters 9.0 ANTICIPATED OPERATIONAL OCCURRENCES.
9.1 Core Hide Transients 9.2 Local Transients 9.3 Reduced Flow Operations 9.4 ASME Overpressurization Analysis 9.5 Increased Flow Operation.
9.6 Single Loop Operation.
9.7 Final Feedwater Temperature Reduction 9.8 Revised Reload Design.
10 12 22 22 22
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22 23 23 24 24 24 25 27 27 28 28 28 29 29 30
HNP-2 CYCLE 6 RELOAD
SUMMARY
REPORT TABLE OF CONTENTS (Continued)
,10.0 POSTULATED ACCIDENTS.
10.1 Loss of Coolant Accident.
10.2 Rod Drop Accident 10.3 Single Loop Operation.
11.0 STARTUP PHYSICS TEST PROGRAM 11.1 Core Load Verification Test.
11.2 Control Rod Functional Test.
11.3 Subcritical Margin Test 11.4 TIP Asymmetry Test.
12.0 REFERENCES
32 32 32 32 33 33 33 33 34 35
WNP-2 CYCLE 6 RELOAD SUHMARY REPORT T
D The fifth reload of the Washington Public Power Supply System Plant No'.
2 (WNP-2) will utilize Advanced Nuclear Fuels Corporation (ANF) 8x8 fuel plus four Lead Fuel Assemblies (LFA) with a 9x9 fuel rod'array manu-factured by General Electric Company (GE) and four LFA assemblies with a 10xl0 fuel rod array manufactured by ABB Atom (ABB). The Bx8 fuel design of this reload batch is identical to the fuel design of the previous reload batch.
The LFA fuel assemblies represent advanced designs described in more detail in Section 6.0 of this report.
This report summarizes the reload analyses performed by ANF, GE and ABB in support of HNP-2 operation for Cycle 6.
In addition, a description of the reload is given along with a comparison of the characteristics of the Cycle 5 and Cycle 6 cores.
A discussion of the proposed physics startup program is also included.
The proposed license amendments (Technical Specification changes) are listed by title in this report for completeness.
This revision of this report is being issued to reflect an addition of eight ANF-5 reload fuel assemblies to the original reload design.
The impact of the addition of the eight ANF-5 reload assemblies is discussed in Section 9.8 and in more detail in Reference 33.
The reload licensing submittal is composed of the HNP-2 Cycle 6 Reload Analysis Report, (Reference 1.0),
the HNP-2 Cycle 6 Plant Transient Analysis Report, (Reference 2.0),
the GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Pro]ect No.
2 Reload 5
Cycle 6, (Reference 3.0),
the Supplemental LFA Licensing Report SVEA-96 LFA's for WNP-2 (Reference 4.0),
the proposed changes to the HNP-2 Technical. Specifications and this report.
Where appropriate, this report summarizes analyses and makes reference to the above reports and other documents for detailed support.
The WNP-2 Cycle 6 Reload Analysis Report is intended to be used in con)unction with ANF Topical Report XN-NF-80-19(P)(A), Volume 4, Revision 1, Application of the ANF Methodology to BHR Reloads (Reference 5.0), which gives a detailed description of the methods and analyses utilized.
2.0 T
M~~P During the fifth refueling outage for WNP-2, the Supply System will replace 144 of'he GE initial core fuel assemblies and 8 XN-1 fuel assemblies with 144 ANF-4/ANF-5 assemblies, 4
The ANF-4 and ANF-5 fuel assemblies are identical.
The GE and ABB LFA's are designed to be compatible with the ANF-4 assemblies.
In addition, there are 32 GE initial core 8x8 fuel assemblies from the Cycle 1
discharge and 24 GE initial core 8x8 fuel assemblies from the Cycle 2
discharge which are being reinserted into WNP-2 in Cycle 6.
These assemblies will displace 56 GE initial core Bx8 fuel assemblies resident in WNP-2 since initial operation.
The ANF-4/ANF-5 assemblies are 8x8 current design reload assemblies which contain a bundle average enrichment of 2.62 weight percent U-235.
The LFA assemblies are described in detail in Section 6.0.
This change in WNP-2 core loading required a reanalysis by ANF and, in the case of the LFA fuel assemblies, required supporting analyses by GE and ABB.
The Loss of Coolant Accident (LOCA) and the Maximum Average Planar Linear Heat Generation (MAPLHGR) analyses relevant to Cycle 6
operation are given in Reference 6.0 as these analyses were performed for all ANF cores as a part of the Cycle 2 (initial reload) analysis.
The LOCA analysis (Reference 6.0) also bounds operation of the 'LFA fuel assemblies as discussed in Section 6.0.
The SVEA-96 and GE LFA's are, by design, bounded by the ANF 8xBC operating limits.
The MAPLHGR and Linear Heat Generation Rate (LHGR) limits for the ANF BxBC assemblies, properly adjusted for different number of fuel rods, can also be applied to the SVEA-96 and GE11 LFA's, since these values are lower than the actual MAPLHGR and MLHGR limits for the SVEA-96 and GE11 assemblies.
In order to provide as much background information as possible and to maintain consistency with past practices
.o the greatest extent possible, SVEA-96 and GEll specific MAPLHGR and MLHGR limits have been included in the technical specifications.
The bases for these values are described in Sections 6.1 and 6.2.
Specific OLMCPR values have been established for the SVEA-96 LFA's using a
conservative approach (leading to high values).
The specific values are discussed in more detail in Section 6.1.
Relevant transient analyses and Minimum Critical Power Ratio (MCPR) analysis for the Cycle 6 loading are reported herein.
Analyses of normal reactor operation consisted of evaluation of the mechanical, thermal hydraulic and nuclear design requirements.
,Operation at extended core flow, single loop operation and final feedwater temperature reduction "are also addressed.
A number of proposed changes to the WNP-2 Technical Specifications have resulted from the design and safety analyses for the Cycle 6 core.
A list of these Technical Specification changes is given in Table 2.1.
INDEX
~QPPj T
TABLE 2.1 1.0 DEFINITIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS; INTRODUCTION 3/4.2.1 3/4.2.3 3/4.2.4 B3/4.2.1 B3/4.2.3 5.3 2.1 Average Planar Linear Heat Generation Rate Hinimum Critical Power Ratio Linear Heat Generation Rate Average Planar Linear Heat Generation Rate
'Minimum Critical Power Ratio Reactor Core r
f'R The 'limiting transient MCPR results for the analyses described in Sections 6,
8 and 9 of this document are summarized in Table 2.2.
The proposed technical specification cha'nge for the HCPR limits includes monitoring the G.E. initial core fuel to the ANF 8x8C fuel HCPR limits.
This results in a less complex HCPR technical specifi-cation.
The analysis performed demonstrate that the G.E. initial core fuel in the proposed Cycle 6 locations is bounded by the ANF 8xBC fuel MCPR limits (References 1
and 2).
WNP-2 will be entering its sixth cycle of operation and is approach-ing an equilibrium cycle.
Analysis results for Cycle 6 and recent previous fuel cycles have shown little change.
TABLE 2.2 MCPR OPERATING LIMITS CYCLE 6 - ANALYSIS RESULTS 0 3750 HWD/MT 3750 EOC 3750 EOC 3750 EOC 3750 EOC 0 -
EOC Normal Scram Times TS Scram Times ANF **
EBG.X38J.
1.24
Normal Scram Times 1.36 ABB LB HKL 1.37
- 1. 48 1.55 1.55
- 1. 61 1.54
1
In this portion of the fuel cycle, operation with the given HCPR operat-ing limits is allowed for both normal and Technical Specification scram times and for both RPT operable and inoperable.
The MCPR operating limits for ANF Sx8C fuel are also applicable to the GE initial core fuel, the GE11 LFA fuel and the ANF 9x9 LFA fuel.
ll h
3.0 WNP-Y P
AT H
T RY WNP-2, a 3323 mwt BWR 5, began Cycle 5 operation on June 26, 1989.*
The end of Cycle 5 is expected to be April 20, 1990.
During Cycle 5, the plant was base loaded at or near 100 percent power unti 1 the all rods out condition was achieved.
At this point, upon NRC approval, a partial feedwater temperature reduction was initiated followed by a thermal coastdown.
Figure 3.1 gives a power history of Cycle 5 through March 1990 for WNP-2.
The Cycle 5 operating highlights and control rod sequence exchange schedule are found in Table 3.1.
Significant thermal operation began on this date.
Significant electrical generation commenced on June 28, 1989.
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TABLE 3.1 T
N Began Electric Power Production End of Cycle Date End of Cycle Core Average Exposure (Design)(mwd/mtm)
Number of Fresh Assemblies Gross Generation (FPD) (Design)
June 26, 1989 April 20, 1990 18,400 136 237 DAT August 6, 1989 October 23, 1989 December 12, 1989 January 30, 1990'FR A2 82 Al Bl B2 Al Bl Al June 28, 1989 through June 30, 1989 August 6, 1989 through August 8, 1989 August 11, 1989 through August 17, 1989 September 21, 1989 through September 29, 1989
alii ERELJIR The WNP-2 core consists of 764 fuel assemblies.
For the Cycle 6 reload, the core will consist of the loading described in Table 4.1.
All reinserted fuel has been qualified for use by means of a documented reinsertion fuel inspection program.
The 144 ANF fresh assemblies consist of 16 reload 8x8C assemblies originally manufactured for loading in Cycle 5 (ANF-4) and 128 reload 8x8C assemblies manufactured for t
loading in Cycle 6 (ANF-5).
The LFA reload assemblies were designed to be compatible with the ANF-4 reload assemblies and are described in detail in Section 6.0.
Table 4.1 lists the assembly type, quantity and initial enrichment for the assemblies which will make up the Cycle 6 core.
Number of
~~i~i~
144 (1) 4 (2) 4 (2) 132 (3) 4 (4) 152 (5) 148 (6) 120 (7) 56 (8)
TABLE 4-1 WNP-Y Bx8C GEll SVEA 96 8xBC 9x9 8x8C Bx8C 8xBC GE PBx8R
~~r) merit 2.62 w/o U-235 2.43 w/o,U-235 2.51 w/o U-235 2.62/2.64 w/o U-235 2.53/2.59 w/o U-235 2.64/2.72 w/o U-235 2.72 w/o U-235 2.72 w/o U-235 1,76 w/o U-235 (2)
(3)
(4)
(5)
,(6)
(7)
(8)
Sixteen (16) of these assemblies were originally fabricated for Cycle 5
(ANF-4) and one hundred twenty-eight of these assemblies were fabricated for Cycle 6 (ANF-5),
They are identical in design.
These designs are discussed in more detail in Section 6.0.
Four (4) of these assemblies were fabricated for reload in Cycle 4 (ANF-3) and have an enrichment of 2.64 weight percent U-235 and one hundred twenty-eight (128) of these assemblies were fabricated for reload in Cycle 5 (ANF-4) and have an enrichment of 2.62 weight percent.
Two (2) of the ANF-4 assemblies contain characterized fuel rods.
Two (2) of these assemblies are 9x9-IX LFA's with a average enrichment of 2.53 weight percent U-235 and two (2) of these assemblies are 9x9-9X LFA's with a average enrichment of 2.59 weight percent U-235.
These assemblies are described in more detail in Section 6.0 of Reference 7.0.
Twenty-four (24) of these assemblies were originally fabricated for reload in Cycle 3 (XN-2) and have an enrichment of 2.72 weight percent U-235 and one hundred twenty-eight (128) of these assemblies were fabricated for reload in Cycle 4 (ANF-3) and have an enrichment of 2.64 weight percent U-235.
Two of the ANF-3 assemblies are lead test assemblies (LTA) of standard 8x8C design.
Thirty-six (36) of these assemblies were originally fabricated for reload in Cycle 2 (XN-1) and one hundred twelve (112) of these assemblies were fabricated for reload in Cycle 3 (XN-2).
They are effectively identical.
Two (2) of these assemblies are lead test assemblies (LTA) of standard ANF Bx8C design.
Thirty-two (32) of these assemblies were discharged from the initial core after Cycle 1
and are being reinserted.
Twenty-four (24) of these assemblies are initial core fuel which were discharged from the core after, Cycle 2 and are being reinserted.
ih
The mechanical design of the 8x8C Cycle 6 ANF reload fuel for WNP-2 is described briefly in Reference 8.0 and more completely in References 9.0, 10,0 and 11.0.
This fuel is identical to the 8x8C Cycle 2
ENC fuel described in Reference 6.0.
The fuel assembly design uses 62 fuel rods and two centrally located water rods, one of which functions as a spacer capture rod.
Seven spacers maintain fuel rod pitch.
The design uses a
quick-removable upper tie plate design to facilitate fuel inspection and bundle reconstitution of irradiated assemblies.
The fuel rods utilize lircaloy-2 cladding, 35 mi ls thick.
The fuel rods are pressurized, and contain either U02 - GdqOq or U02 with a nominal density of 94.5 percent of theoretical density tTD), and an 8.5 mil nominal diametrical pellet to clad gap for the enriched pellets.
Natural uranium is loaded in the top and. bottom six inches of each fuel rod for greater neutron economy.
The enriched pellets have a slightly larger diameter than the natural pellets.
The fuel mechanical design analysis performed on the ANF 8x8C reload fuel evaluated the following items (References 8.0.and 11.0):
~
Cladding steady state strain and stress
~
Transient strain and stress
~
Cladding fatigue damage
~
Creep collapse
~
Corrosion
~
Hydrogen absorption
. ~
Fuel rod internal pressure
~
Differential fuel rod growth
~
Creep bow
~
Grid space design The analyses presented in References 8.0 and 11.0 justify irradiation to a 35,000 MWD/MT peak assembly burnup in WNP-2.
Some ma]or results of these analyses are:
The maximum end-of-life (EOL) steady state cladding strain is well below the 1 percent design limit.
Cladding steady state stresses are calculated below the material strength limits.
~
The transient strain does not exceed 1.0 percent.
~
The cladding fatigue usage factor is within the 0.67 percent design 1 imi t.
~
The cladding diameter reduction due to uniform creepdown, plus creep ovality at maximum densification, is less than the minimum initial gap.
Compliance with this criteria prevents the formation of fuel column gaps and the possibility of creep collapse.
~
The maximum level of the corrosion layer was calculated to be well within the design limit.
~
The maximum concentration of hydrogen was calculated to be well within the design limit.
~
Evaluations of the fuel assembly growth and differential fuel rod growth show that the fuel assembly design provides adequate clearance.
~
The plenum spring complies with design limits,
~
The spacer spring meets all design'requirements.
~
The maximum fuel rod internal rod pressure remains below ANF's criteria limit.
~
The fuel centerline temperature remains below the melting point.
The structural response of the 8x8C ANF-5 reload fuel is the same as the structural response of the 8x8C ANF-4 fuel, the SxSC ANF-3 fuel, the Sx8C XN-2 ANF fuel, the 8xSC XN-1 fuel and the.P8x8R GE fuel which also reside in the NNP-2 Cycle 6 core.
As part of Cycle 6 operation, some of the 8xSC Cycle 5 ANF reload fuel assemblies may be channeled with new 100 mil channels fabricated by ABB as has been the practice in the past.
These channels are equivalent to the initial core channels.
The remainder of the reload fuel bundles will be channeled with channels which have been previously discharged from NNP-2.
Prior to reuse, these channels are measured with the NNP-2 channel measuring machine and qualified for reuse based upon a predetermined criteria.
Therefore, the seismic LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertions will not be inhibited following occurrence of the design basis seismic LOCA event.
An LHGR limit is placed on ANF Sx8C Cycle 6 reload fuel assemblies for monitoring for the reasons given previously in Reference 6.0, Page 10, for ENC Sx8C Cycle 2 fuel.
11
6.0 P
A The lead fuel assembly program for Cycle 6 of HNP-2 consists of four SVEA-96 fuel assemblies and four GEll fuel assemblies.
The SVEA-96 fuel assembly is designed as a reload assembly for BHR/3 through BHR/6 cores, both 0 and C lattice plants (HNP-2 is a
C lattice BWR-5).
.The assembly consists of a 10xl0 fuel rod array with four central rods removed.
The most obvious feature of the SVEA-96 design is the integral central water cross providing significantly more non-boiling water in the center of the assembly.
The GEll fuel assembly consists of a 9x9 fuel rod array with seven central fuel rods removed and replaced with two large central water rods.
Several of the fuel rods are part length fuel rods (PLR) which are selectively located in the lattice to reduce two-phase pressure drop and increase cold shutdown reactivity margin.
The PLR's terminate just past
.the top of the 5th spacer.
Table 6.1 gives a comparison of key parameters for the ANF-5 8xBC assembly and the SVEA-96 assembly.
For the GEll key parameters, see Reference 3.0.
The LFA's will be placed in core locations which have been analyzed to have sufficient margin such that they are not expected to be limiting assemblies in the core on either a nodal or a bundle power basis.
This approach is intended to prevent the possibility of the LFA's from ever being the limiting fuel assembly.
The LFA's were designed to match the ANF 8xBC reload fuel in neutronic and thermal-hydraulic characteristics.
Because of this similarity in design characteristics, the operating limits of the ANF BxBC reload fuel bound the LFA fuel except where specifically noted.
- 12
PAf~M~T TABLE 6.1 FUEL ASSEMBLY DESIGN PARAMETERS Total Number of Rods Fuel Rod Pitch, Inches Fuel Assembly Loading, KgUO Fuel Assembly Loading, KgU Fuel Pellet Material Density g/cc Percent of T.D.
- Diameter, inches enriched natural Dish Volume
('/ of pellet vol~me) enriched UO enriched UO -Gd203 natural Fuel Rod (6" natural enriched material on each end)
Fuel length, inches Cladding Material Clad I.D., inches Clad O.D.,
inches Enriched length, inches Fuel Rod Inventory Inert Water Rod Unvoided Hater Area (Square Inches) 64
.641 199.7 176.0 U02 10.36 94,5
.4055
.4045 1.50 1.00 0.00 150 2r-2*
~ 414
,484 138 1 (1.5 w/o U-235) 5 (2 '
w/o U-235) 9 (2.5 w/o U-235) 6 (2.5 w/o U-235
+ 2.0 w/o Gd203) 21 (2.64 w/o U-235) 20 (3.43 w/o U-235) 2
.368 96
.488(within each bundle) 200.0 176.3 U02 10.5 95.8
.3224
.3224 1.0 1.0
.43 150 Zr 2**
.3291
.3787(.3921 spacer capture rod) 138.2 4 (1.63 w/o U-235) 4 (2.34 w/o U-235) 16 (2.49 w/o U-235) 24 (2.64 w/o U-235) 8 (2.64 w/o + 2.55 w/o Gdy03) 24t2.81 w/0 U-235) 16 (3.0 w/o U-235) 1 Water Cross 2.16
- Gdg03, Rods use Beta heat treated Zr-2 A17 rods use Beta heat treated 1r-2 13
l
Analyses have been performed to establish a licensing basis for the four ABB SVEA-96 LFA assemblies to be included in the NNP-2 Cycle 6
core (Reference 4.0).
The analyses demonstrate the applicability of the HNP-2 Cycle 6 operating limits to these four LFA's unless stated otherwise herein.
The design basis of these LFA's are set forth in Reference 4.0.
Fulfillment of the design basis assures that the SVEA-96 LFA's will not:
~
involve a significant increase in the probability or consequences of an accident previously evaluated; or
~
create the possibility of a new or different kind of accident from any accident previously evaluated; or
~
involve a significant reduction in a margin to safety.
The dynamic response of the four SVEA-96 LFA's is sufficiently similar to the ANF 8x8C reload fuel and the GE initial core fuel to provide assurance that the control blade insertions will not be inhibited following occurrence of the design basis seismic LOCA event.
Insertion of four SVEA-96 LFA's in the Cycle 6 core will have negligible effects upon core wide transients.
The bundles were designed to match the nuclear characteristics of the ANF 8x8C reload fuel bundles' SVEA-96 LFA specific analysis has been performed to develop MCPR operating limits for the LFA's.
SVEA-96 LFA MCPR values have been developed and are included in Table 6.2 and in the proposed technical specification changes.
For reduced flow operation, the MCPR operating limit is the larger of the MCPR operating limit shown in Table 6.2 or the value given by Figure 3.2.3-1 (the reduced flow MCPR curve) in the WNP-2 Technical Specification (Reference 12).
The SVEA-96 LFA's have been designed to be hydraulically compatible with the Cycle 6 core and specifically to match the co-resident ANF-4 8x8C.
Steady state thermal hydraulic analysis has shown that the total bundle flow and bypass flow of the SVEA-96 LFA's will be sufficiently similar to the ANF reload assemblies at rated condi-tions to assure that the presence of the LFA's will not affect the thermal performance or core flow distribution in an adverse manners 14-
~vn
- LRNB, EOC RPT Inoperable, NSS
- LRNB, EOC RPT Inoperable, TSSS FWCF RPT Operable FWCF RPT Inoperable LRNB, 0 MWD/MTU - 3750 MWD/MTU TABLE 6.2 WNP-2 SVEA-96 MQP~~
1.48 1.55 1.55 1.61 1.42 1
~ 52 1.37 MCPR value using the 1.06 Safety Limit.
- Technical Specification Scram Speed (TSSS)
0,
Within the nuclear safety area the following evaluations have been performed:
Influence of inserting SVEA-96 LFA's on the shut down margin has been determined.
It was shown that the change in shut down margin at the most limiting time is insignificant-approximately 30 percent milli K.
The insertion of the SVEA-96 LFA's will have no significant effect on the performance of the standby liquid control system.
~
The licensing basis for the spent fuel storage racks will not be invalidated by insertion of SVEA-96 LFA's.
The control rod withdrawal error has been analyzed (Reference 4.0) and it has been concluded that the insertion of-the SVEA-96 LFA's will not invalidate the results for this event in the Reload Safety Evaluation (Reference 1.0).
Furthermore, it has been concluded that the control rod withdrawal error event will not be limiting for the SVEA-96 LFA's at EOC.
The loss of coolant accident event has been analyzed and it has been concluded that the SVEA-96 LFA's can operate at higher planar power than the co-resident 8x8C reload fuel assemblies and the peak cladding temperatures in the SVEA-96 LFA's will still be lower than the peak cladding temperatures in the 8xBC reload assemblies.
It has been concluded that the MAPLHGR limit specified for the Bx8C reload assembly, properly adjusted to account for different number of rods, can be used for the SVEA-96 assemblies as well.
The recommended MAPLHGR limit for SVEA-96 is shown in Table 6.3 along with the 8xBC value.
Based on an evaluation against the SVEA-96 design bases, it has been determined that the Maximum Linear Heat Generation Rate (MLHGR) in the SVEA-96 LFA's should not exceed 11.6 kW/ft.
The value is higher than the corresponding MLHGR limit for the 8x8C fuel derived after properly accounting for the different number of fuel rods and errors in the power monitoring of the SVEA-96 LFA's.
Therefore, the values for the 8x8C fuel can be applied to the SVEA-96 LFA's as well.
The recommended LHGR for the SVEA-96 LFA's is included in the proposed technical specification changes.
The control rod drop accident has been assessed and it has been concluded that the insertion of the SVEA-96 LFA's will not invali-date the conclusions reached in the Reload Safety Evaluation (Reference 4.0),for this accident.
The maximum fuel enthalpy for the SVEA-96 LFA's will remain below the limits set by acceptance criteria.
16-
h
The channel flow stability characteristics of the SVEA-96 assemblies have been compared to the same characteristics for 8x8 fuel assem-blies.
Based on experimental results from loop'easurements as well as from reactor operation, it is concluded that the coolant flow through SVEA-96 assemblies is more stable than through Bx8 fuel assemblies.
Insertion of the SVEA-96 LFA's will therefore not increase the risk for instabilities.
The nuclear design of the SVEA-96 LFA's is completely symmetric.
There will therefore be no appreciable power distribution change resulting from the misorienting of the SVEA-96 LFA's.
Hisorienting a SVEA-96 LFA wi 11 have less effect than misorienting an ANF-4 assembly since the major difference is that the enrichment distri-bution within the SVEA-96 assembly is symmetric.
Identifiable features on the SVEA-96 LFA's permit verification of proper orienta-tion of the assembly.
These features reduce the probability of misorientation.
17
The control rod reactivity worth of the cells containing the GEll
.LFAs was calculated..
Two cases were run for comparison purposes.
The case for which the LFAs were included demonstrates greater shutdown margin (0.21. delta K) for the cell containing the LFA and for the adjacent cells.
The LFAs will also result in increased margin for the Standby Liquid Control System shutdown condition.
The effect of the LFAs on core-wide transients is established by the fuel thermal time constant and the
- Scram, Doppler, and Void reacti-vities.
The LFAs have a reduced time constant relative to the ANF 8x8C reload fuel assemblies.
This will have a negligible but bene-ficial impact on core response.
Also, because the LFAs represent
- only a very small fraction of the core, the core average
'oppler and Void reactivities are not affected.
Finally, because the LFAs will have higher thermal margins than the ANF 8x8C reload fuel assembly during normal operation, sufficient margins will also be present during transients to compensate for the 4/. larger CPR changes associated with the LFAs shorter time constant.
Therefore, the MCPR operating limit for the ANF fuel may be conservatively applied to the LFAs.
For the Rod Withdrawal Error event, evaluations demonstrate that the LFAs have greater thermal margins than the standard reload assemblies.
The Fuel Loading Error event is precluded by core verification following loading.
LOCA evaluations demonstrate that the GEll LFAs will have about 150 degrees-F lower peak clad temperature than typical 8x8C reload fuel assemblies at comparable planar powers.
Therefore, the MAPLHGR limits for the ANF 8x8C reload fuel assemblies may be conservatively applied to the GEll LFAs being monitored as 8x8 assemblies.
As added conservatism, the ANF 8x8C reload fuel assembly MAPLHGR values were modified to account for the larger number of fuel rods in the GEll LFA fuel assembly relative to the ANF 8x8 fuel assembly.
It is these modified values which are being placed in the WNP-2 technical specifications for GEll fuel.
18
TABLE 6.3 WNP-2 SVEA-96 AP R
HIT Bundle Average Exposure (HWd/MTU) 5,000 10,000 15,000 20,000 25,000 30,000 35,000 ANF 8xBC MAPLHGR (IQW/ft) 13.0 13.0 13.0 13.0 13.0 11.3 9.4 7.9 SVEA-96 MAPLHGR (IQW/ft) 8.90 8.90 8.90 8.90 8.90 7.74 6.44 5 '1 Confirmation that the SVEA-96 LFA's satisfy Safety Limit Minimum Critical Power Ratio (SLMCPR) Oesign Basis requires either that the continued adequacy of the WNP-2 plant Safety Limit MCPR(SLMCPR) with the SVEA-96 LFA's installed is demonstrated or that appropriate modifications to the SLMCPR be identified.
A SLHCPR of 1.05 was computed for a full SVEA-96 core.
Therefore, it is concluded that the current plant SLMCPR of 1.06 can be conservatively applied to the SVEA-96 LFA's (Reference 4.0).
Based upon an evaluation against the SVEA-96 design basis, it has been determined that the Maximum Linear Heat Generation Rate (MLHGR) in the SVEA-96 LFA's should not exceed 11.6 kw/ft.
The recommended LHGR for the SVEA-96 LFA's is included in the proposed technical specification changes'19-
6.2 1
1 m
1 The nuclear design of the GEll LFAs was chosen to represent, as closely as practical, the characteristics of the ANF-4 Bx8C reload assembly.
- However, because of the advanced features of the GE11 LFAs, increased margin exists to design and licensing limits relative to the ANF SxSC assembly (Reference 3.0).
The GE11 LFAs were analyzed using the NRC approved GESTR-MECHANICAL, GEMINI NUCLEAR and ODYN methods.
The response of the GE11 inter-active channel has been reviewed for dynamic response and has been found to be acceptable (Reference 13).
The NRC approved GEXL-Plus critical power correlation for the GESx8NB fuel design was applied to the GEll LFAs.
The R-factor was adjusted to fit available prototype test data for the GE11 fuel assembly design.
The CHASTE model was used to compare MAPLHGR limits.
Because the nuclear, mechanical and thermal-hydraulic characteris-tics of the ANF 8x8C reload fuel assembly are similar to other approved GE Sx8 designs, it too can be readily modeled using the approved GE methods.
Comparisons were made based on an ANF 8xSC reload fuel assembly in a limiting core location and a
GE11 LFA in the same core location operating at equivalent nodal powers.
'In practice,
- however, the LFAs will be loaded in core locations near the periphery such that they will not be the most limiting fuel assemblies in the core at any time during their residence in the core.
Results of the calculations demonstrate that the GE11 LFAs have greater margins to licensing-and design limits than the ANF 8xSC reload fuel assemblies that are used as the models in the reload analyses.
The increased LHGR margin is due to the larger number of fuel rods in the LFA and increased MCPR margins result from the application of the GEXL-Plus critical power correlation for LFA assemblies with the high performance spacer.
The results also showed that replacing four ANF 8x8C reload fuel assemblies with the four GEll LFAs does not significantly impact core-wide character-istics.
Therefore, modeling and tracking the GE11 LFAs as standard ANF 8xSC reload fuel assemblies is conservative and the LFAs can be loaded into the WNP-2 core and operated without any special considerations.
Typical beginning-of-cycle and peak core reactivity conditions were calculated using approved methods for an equilibrium core of ANF 8x8C reload fuel.
Cases were run without and with the LFAs.
The results demonstrate that the LFAs have an insignificant effect on core average performance and on the performance of adjacent assem-blies.
The results also indicate that the LFAs have greater margin to MCPR and LHGR limits than the standard ANF assembly operating under similar conditions.
The control rod reactivity worth of the cells containing the GEll LFAs was calculated.
Two cases were run for comparison purposes.
The case for which the LFAs were included demonstrates greater shutdown margin (0.2'/ delta K) for the cell containing the LFA and for the adjacent cells.
The LFAs will also result in increased margin for the Standby Liquid Control System shutdown condition.
The effect of the LFAs on core-wide transients is established by the fuel thermal time constant and the Scram,
- Doppler, and Void reacti-vities.
The LFAs have a reduced time constant relative to the ANF 8x8C reload fuel assemblies.
This will have a negligible but bene-ficial impact on core response.
Also, because the LFAs represent only a very small fraction of the core, the core average
- Scram, Doppler and Void reactivities are not affected.
Finally, because the LFAs will have higher thermal margins than the ANF 8x8C reload fuel assembly during normal operation, sufficient margins will also be present during transients to compensate for the 4'/ larger CPR changes associated with the LFAs shorter time constant.
Therefore, the MCPR operating limit for the ANF fuel may be conservatively applied to the LFAs.
For the Rod Hithdrawal Error event, evaluations demonstrate that the LFAs have greater thermal margins than the standard reload assemblies.
The Fuel Loading Error event is precluded by core verification following loading.
LOCA evaluations demonstrate that the GE11 LFAs will have about 150 degrees-F lower peak clad temperature than typical 8x8C reload fuel assemblies at comparable planar powers.
Therefore, the MAPLHGR limits for the ANF 8x8C reload fuel assemblies may be conservatively applied to the GEll LFAs being monitored as 8x8 assemblies.
As
,added conservatism, the ANF 8x8C reload fuel assembly MAPLHGR values were modified to account for the larger number of fuel rods in the GEll LFA fuel assembly relative to the ANF 8x8 fuel assembly.
It is these modified values which are being placed in the HNP-2 technical specifications for GE11 fuel.
21
The goal of the thermal hydraulic design analysis is to demonstrate that the ANF reload fuel meets and/or exceeds the primary thermal hydraulic design criteria.
Principal design criteria considered in the thermal hydraulic analysis are found in XN-NF-80-19(A), Volume 4, Revision 1
(Reference 5.0).
Analyses performed to demonstrate that these criteria are met include:
~
Hydraulic compatability
~
Fuel cladding integrity safety limit
~
Fuel centerline temperature
~
Bypass flow characteristics
~
Thermal hydraulic stability The analyses discussed in this section are for the Bx8 fuel.
Specific thermal hydraulic design considerations for the LFA fuel are discussed in Section 6.0.
Thermal hydraulic design considerations applicable to the ANF 9x9 LFA's loaded last cycle are discussed ln Reference 7.0.
7.1 The hydraulic flow.resistances for the ANF reload fuel and the GE initial core Bx8 fuel have been determined in single phase flow tests of full scale assemblies.
XN-NF-80-19(A), Volume 4, Revi-sion 1 (Reference 5.0), reports the resistances measured and evalu-ates the effects on thermal margin of mixed ANF and GE 8x8 cores.
The close geometrical similarity between the fuel designs and their-measured performance characteristics demonstrate that the fuel designs are sufficiently compatible for co-residence in NNP-2.
7 2
F 1
1 in ri f
imi The HCPR fuel cladding integrity safety limit for Cycle 6 is 1.06 which is equal to the HCPR safety limit for all previous fuel cycles.
The methodology used in the HCPR safety limit calculations is found in XN-NF-80-19(A), Volume 4, Revision 1 (Reference 5.0).
The NNP-2 Cycle 6
HCPR safety limit analysis methodology and input parameters are described in ANF-90-01, Cycle 6 Plant Transient Report (Reference 2.0).
7.3 F
rl n
Tm The LHGR curve in Figure 3.4 of Reference 11.0 for ANF Bx8 fuel is everywhere greater than 120 percent of the LHGR limit curve in Reference 11.0.
Therefore, the fuel centerline temperature is protected for 120 percent over power and fuel centerline melt is protected for all fuel exposures within the bounds of the referenced LHGR curve.
II'I
7.4 Fl w
Core bypass flow was computed using the methodology of XN-NF-524(A)
(Reference 14.0).
The bypass flow for the WNP-2 Cycle 6 is 10.4 percent of the total core flow which is similar to the Cycle 1 value of 11.8 percent and to the Cycle 5 value of 10.7 percent.
The change ln bypass flow will have no adverse impact on reactor operation.
The WNP-2 Technical Specifications include surveillance requirements for detecting and suppressing power oscillations.
In addition, the ANF COTRANSA code (Reference 15.0) was used to specifically deter-mine that the worst case value of decay ratio is less than 0.5 in the area of the power flow map bounded by the APRH rod block line at 45 percent rated flow.
The worst case decay ratio is less than 0.9 in the area of allowable low flow operation (detect and suppress region).
The bounding power flow points in the detect and suppress region are the APRH rod block line at 27.6 percent core flow (47 percent power minimum allowable two pump flow) and the APRH rod block line at 23.8 percent core flow (42 percent power-natural circulation) (Reference 1.0).
II
- 8. 0 NQQJ~RD~[95 The neutronic methods for the design and analysis of the WNP-2 Cycle 6
reload are described in References 15.0 and 16.0.
These methods have been reviewed and approved by the U.S.
Nuclear Regulatory Commission for generic application to BWR reloads.
8 1
1 n
1 N
1 D
The Cycle 6 8xS ANF reload assemblies (labeled ANF-5) are identical to the ANF-4 reload bundles (used in the WNP-2 Cycle 5 reload) in nuclear design in all major parameters including fuel enrichment.
Major nuclear design characteristics for the ANF reload fuel assemblies (ANF-5) are:
8.2
~
The 8x8C fuel assembly contains 62 fuel rods and two water rods.
One of the water rods also acts as a spacer capture rod.
~
The ANF-'5 8xSC fuel assembly average enrichment is 2.62 w/o U-235.
The top and bottom six inches of the fuel rods contain natural uranium.
The central 138 inch portion of the fuel rods has an average enrichment of 2.79 w/o U-235.
~
Five enrichment levels are utilized in the ANF-5 SxBC fuel assembly to produce a local power distribution which results in a balanced design for Minimum Critical Power Ratio (MCPR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) 1 imits.,
~
Each ANF-5 8x8C fuel assembly contains six fuel rods with 2.0 w/o GDq03 blended with 2.50 w/o U-235 enriched U02 to reduce initiaT assembly reactivity.
The enrichment distribution of the ANF reload design was selected on the basis of maintaining a balance between the local power peaking
For the central enriched region of the ANF-5 SxS assembly, one rod is enriched to 1.5 w/o U-235, five rods to 2.0 w/o U-235, nine rods to 2.50 w/o U-235, 21 rods to 2.64 w/o U-235, 20 rods to 3.43 w/o U-235, and six rods to 2.50 w/o U-235 plus 2.00 w/o GD203.
The core exposure for the end of Cycle 5 (EOC5),
the core exposure for the beginning of Cycle 6 (BOC6),
and the core exposure for the end of Cycle 6
(EOC6) were calculated with the XTGBWR Code (Refer-ence 15.0).
. In addition, BOC core reactivity characteristics for the cold core were calculated along with the standby liquid control system reactivity.
The results of these analyses are summarized in Table 8,1.
TABLE 8,1 CORE NUCLEAR DESIGN Core Exposures at EOC5 (mwd/mtm) 18,700 Core Exposures at BOC6 (mwd/mtm)
Core Exposures at EOC6 (mwd/mtm)
BOC6 Cold Keff, all rods out BOC6 Cold Keff, strongest rod out 12,800 18,800 1.1141
.9879 Reactivity Defect/R-Value, percent
'/ DK/K 0.21 Standby Liquid Control System (SBLC)
Reactivity, 660 PPM Boron, Keff
. 9641 8.3 m ri n f M r
r Prm~i~
Some of the major core parameters for WNP-2 Cycle 5 and Cycle 6 are listed in Table 8'.
TABLE 8.2 COMPARISON OF MAJOR CORE PARAMETERS
~Pr m~r HCPR Limit* (0 -3750mwd/mtm)
Doppler Defect
( /. hK/K/T)
Cycle Length** (Design; FPD)
Core Average Exposure (BOC; mwd/mtm)
Core Average Exposure (EOC; mwd/mtm) 1.24
-1.0x10 5
227 12,300 18,100 1.24
-1.0xlO 5
227 12,800 18,800
- Based on
- CRWE, 106/.
RBH setpoint
'* All rods out; full power, 105/ flow
The major differences between the Cycle 5 core and the Cycle 6 core are found in the core loading pattern.
The Cycle 5 core consisted of a scatter load of:
132 ANF 8x8C reload assemblies, 4 ANF 9x9 LFA assemblies, 152 once irradiated ANF Sx8C reload assemblies, 148 twice irradiated 8x8C reload assemblies, 128 thrice irradiated SxSC reload assemblies, 200 P8x8R initial core GE assemblies.
The Cycle 6 core will consist of:
144 fresh 8x8C reload assemblies, 4 GE11 fresh LFA assemblies, 4 SVEA-96 fresh LFA assemblies, 132 once irradiated ANF 8x8C reload assemblies, 4 once irradiated ANF 9x9 LFA assemblies, 152 twice irradiated ANF 8x8C reload assemblies, 148 thrice irradiated ANF 8xSC reload assemblies, 120 ANF 8x8C reload assemblies with four cycles of reactor
- exposure, 56 initial core GE P8x8R fuel assemblies.
9.0 T
P T
P AT N
ANF considers eight categories of potential system core wide transient occurrences for ]et pump BWRs (Reference 17.0) and has provided analysis results for the three most limiting transients for WNP-2 Cycle 6 to determine the Cycle 6 thermal margins.
The three transients determined to be most limiting for Cycle 6 are:
~
Load Re]ection No Bypass (LRNB)
~
Feedwater Controller Failure (FWCF)
~
Loss of Feedwater Heating (LOFH)
The discussion in Reference 17.0 demonstrates that the other plant tran-sient events are inherently nonlimiting or clearly bounded by the above events.
ANF's methodology for developing thermal limits is found in Reference 18.0.
Two local events, Control Rod Withdrawal Error (CRWE) and Fuel Loading Error (FLE) were analyzed with the methodology described in Reference 15.0.
The CRWE, analyzed at a
106 percent RBM setpoint, was demonstrated to be bounding for certain parts of the fuel cycle.
The analysis reported here is applicable to the GE initial core fuel, the ANF 8x8C reload fuel and the ANF 9x9 LFA fuel included in the Cycle 6
core (References 1.0 and 2.0).
It is also applicable to the GEll and SVEA-96 LFA fuel as discussed in Section 6.
The results of the core-wide and local transient analyses are provided in the WNP-2 Cycle 6 Reload Analysis Report (Reference 1.0) and in the WNP-2 Cycle 6 Transient Analysis Report (Reference 2.0).
Additional analyses were performed to determine the MCPR operating limit with a 107 percent and 108 percent RBM setpoint for the CRWE event.
9.1 r
Wi Tr n
i n
The plant transient model used to evaluate the pressurization tran-
- sients, the LRNB and FWCF events, consists of the ANF COTRANSA (Reference 17.0) and XCOBRA-T (Reference 19.0) codes.
This axial one dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred.
This phenomenon was accounted for explicitly in determining thermal margin changes in the transient.
All pressurization transients were analyzed on a
bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel model.
The LRNB at 106/ core flow and EOL was found to be the most limiting core wide transient for all analyzed events.
All core wide transients were analyzed using bounding values as input.
The Loss of Feedwater Heating (LOFH) events were evaluated with the ANF core simulator model XTGBWR (Reference 15.0) by representing the reactor in equilibrium before and after the event.
Actual and pro-
]ected operating statepoints were used as initial conditions.
Final conditions were determined by reducing the feedwater temperature by 100'F and increasing core power such that the calculated eigenvalue remained unchanged.
Based on a bounding value analysis for LOFH, a
MCPR operating limit of 1.15 with a HCPR safety limit of 1.06 is supported (i.e.,
a change in CPR of 0.09).
The HCPR safety limit for Cycle 6
continues'o be 1.06; hence the LOFH transient requires a
HCPR operating limit of 1.15 (Reference 2.0).
9.2 Tr in Analyses given in Reference 1.0 show that the FLE transient is bounded by the CRWE transient and is therefore nonlimiting.
Based on the
.CRWE results, the HCPR operating limit is a function of the RBM setpoint.
Analyses were performed to support a
RBM setpoint of 106 percent, 107 percent, and 108 percent.
The change in CPR for the CRWE with a 106 percent RBH setpoint is 0.18 for ANF Bx8 fuel and GE initial core fuel, for a 107 percent RBH setpoint 0.21 for ANF Bx8 and GE initial core fuel, and for a 108 percent RBM setpoint 0.23 for ANF Bx8 fuel and GE initial core fuel.
9.3 The recirculation flow run-up analysis performed for WNP-2 Cycle 5
was reviewed and was found to remain applicable for WNP-2 Cycle 6
operation and for Final Feedwater Temperature Reduction (FFTR).
(Reference 2.0).
9.4 A H v r r ri i
An In order to demonstrate compliance with the ASME Code over pressuri-zation criteria of 110 percent of vessel design pressure (1375 psig),
the Hain Steam Isolation Valve (HSIV) closure event with failure of the MSIV position switch scram was analyzed with ANF's COTRANSA code (Reference 17.0).
The WNP-2 Cycle 6 analysis assumed six safety relief valves out of service.
The maximum pressure observed in the analysis is 1317 psig in the vessel lower plenum.
This is 105 percent of the reactor vessel design pressure which is below the 110 percent design criterion of 1375 psig.
The steam dome safety limit of 1325 PSIG was also examined.
The calculated steam dome pressure corresponding to the 1317 psig peak vessel pressure is 1291 psig, for a vessel differential pressure of 26 psig.
The RPT is assumed to initiate at a pressure setpoint of 1170 psig.
The current Technical Specification Safety limit of 1325 psig is based on dome pressure and conservatively assumes a
50 psi vessel pressure differential (1375-1325),
Since the calculated vessel differential pressure is 26 psi, the steam dome safety limit of 1325 psig assures compliance with the ASHE criterion of 1375 psig peak vessel pressure.
9.5 Inr Fl w
r i
The Cycle 2 transient events analyzed at the design basis power condition with increased core flow (1064) were found to bound the same transients analyzed at the design basis power and rated flow 100'ondition for HNP-2 Cycle 2 (Reference 20.0).
ANF has also performed analyses which demonstrate that the XN-1 Sx8C fuel bundle can operate satisfactorily from a mechanical standpoint at this increased core flow (Reference 2.!.0).
In addition, GE has performed analyses for the reactor internals and for the GE initial core fuel assembly which considered the loads created by operation at this flow level and the impacts of these loads on the HNP-2 core internals and the GE fuel assembly.
Also, flow induced vibration of the core internals as a result of increased core flow was analyzed.
Finally, analyses were performed for feedwater nozzle and feedwater sparger fatigue at increased core flow (Reference 22.0).
The results of all these analyses when considered along with the sim-ilarityy with the fuel types utilized in Cycle 6, corfirm the capability of HNP-2 to operate at 100 percent power and 106 percent core flow during Cycle 6 operation.
A review of the LFA fuel, discussed in References 3.0, 4.0 "and 7.0, confirm their capability for increased flow operation.
A containment analysis was performed to determine the impact of operation at increased core flow on the WNP-2 containment LOCA response.
The results show that the containment LOCA response for increased core flow operation is bounded by the corresponding FSAR results (Reference 23.0).
9.6 In summary, all relevant neutronic, thermal hydraulic, mechanical, and safety analyses have been performed to demonstrate that HNP-2 can operate safely with extended core flow up to 106 percent of rated core flow during Cycle 6.
ANF confirmed analyses for WNP-2 which demonstrate the safety of plant operation with a single recirculation loop out of service at 75/ of rated power for an extended period of time (Reference 1.0).
These analyses were performed for the most limiting transient
- events, the pump seizure accident and the loss-of-coolant accident (LOCA) for the maximum extended power state during HNP-2 single loop operation (SLO).
The results of the SLO analyses are summarized below:
~
The two loop MCPR operating limits (rated conditions) bound the transient requirements for SLO.
The single loop transient analyses need not be performed on a cycle by cycle basis and a
MCPR limit equal to 1.35 is appropriate for single loop conditions.
tI
~
The postulated pump seizure accident, evaluated for SLO condi-
- tions, is calculated to have a less severe radiological release than the LOCA.
The radiological consequences of this postu-lated accident are bounded by the radiological evaluation per-formed by GE for the LOCA and are well within the 10CFR100 limits.
~
The single loop ECCS analysis supports the use of the WNP-2 two loop HAPLHGR limits for ANF Bx8 fuel when the reactor is operating in the SLO mode consistent with the single loop MCPR Operation limit.
Single loop operation of WNP-2 with the two loop ANF Bx8 fuel MAPLHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S.
NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.
The transient and pump seizure accident analyses are described in Reference 24.0 and the LOCA analyses are described in Reference 25.0 With a single recirculation loop in operation, the GE initial core analyses supported continued operation with an increase of 0.01 in the MCPR safety limit (Reference 23.0),
ANF performed a single loop HCPR safety limit calculation and found that less than one tenth of one percent of the rods to be in boiling transition which supports a
MCPR safety limit of 1.07.
Because of the similarity between the ANF and GE fuel types making up the core, and because of the simi-larity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value can be used for operation with ANF fuel and single loop analyses.
For Cycle 6 operation with both recirculation loops in operation, the MCPR safety limit is 1.06, which is the same value as was used for the previous cycles.
For Cycle 6 operation with a single recir-culation loop in service, the HCPR safety limit is 1.07, which is also the same value used for the previous cycles.
The LFA single loop operation limits are bounded by the two-loop operation limits (References 1.0, 3.0 and 4.0).
9.7 in 1
F w
m r
r R
i n
Reference 26.0 presents a final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2.
This analysis has beem approved and is applicable to future WNP-2 fuel cycles.
The FFTR analysis was performed for a 65'F temperature reduction.
This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature.
The FFTR analysis results show that CPR changes for the LRNB and FWCF transients of
+ 0.02 and 0.01 respectively, are applicable to these respective anticipated operational occurrence (AOO) events.
That is, these LRNB and FWCF limit changes are applicable when Cycle 6 reactor operation is being extended with thermal coastdown at FFTR cond>-
tions and are applicable to all fuel designs except the SVEA-96 LFA fuel (References 1,0, 3.0 and 4.0),
A supplemental analysis (Refer-ence 34.0) demonstrates that a 0.03 CPR addition is applicable for the SVEA-96 LFA fuels
9.7 The postulated pump seizure
- accident, evaluated for SLO condi-
- tions, is calculated to have a less severe radiological release than the LOCA.
The radiological consequences of this postu-lated accident are bounded by the radiological evaluation per-formed by GE for the LOCA and are well within the 10CFR100 limits.
f The single loop ECCS analysis supports the use of the HNP-2 two loop MAPLHGR limits for ANF Bx8 fuel when the reactor is operating in the SLO mode consistent with the single loop MCPR Operation limit.
Single loop operation of HNP-2 with the two loop ANF Bx8 fuel MAPLHGR limits assures that the emergency core cooling systems for the HNP-2 plant wi 11 meet the U.S.
NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.
The transient and pump seizure accident analyses are described in Reference 24.0 and the LOCA analyses are described in Reference 25.0.
With a single recirculation loop in operation, the GE initial core analyses supported continued operation with an increase of 0.01 in the MCPR safety limit (Reference 23.0).
ANF performed a single loop MCPR safety limit calculation and found that less than one tenth of one percent of, the rods to be in boiling transition which supports a
MCPR safety limit of 1.07.
Because of the similarity between the ANF and GE fuel types making up the core, and because of the simi-larity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value can be used for operation with ANF fuel and single loop analyses.
For Cycle 6 operation with both recirculation loops in operation, the MCPR safety limit is 1.06, which is the same value as was used for the previous cycles.
For Cycle 6 operation with a single recir-culation loop ln service, the MCPR safety limit is 1.07, which is also the same value used for the previous cycles.
The LFA single loop operation limits are bounded by the two-loop operation limits (References 1,0, 3,0 and 4.0).
Fin 1
F w
Reference 26.0 presents a final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2.
This analysis has beem approved and is applicable to future WNP-2 fuel cycles.
The FFTR analysis was performed for a 65'F temperature reduction.
This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature.
The FFTR analysis results show that CPR changes for the LRNB and FHCF transients of
+ 0.02 and - 0.01 respectively, are applicable to these respective anticipated operational occurrence (AOO) events.
That is, these LRNB and FHCF limit changes are applicable when Cycle 6 reactor operation is being extended with thermal coastdown at FFTR condi-tions and are applicable to all fuel designs except the SVEA-96 LFA fuel (References 1.0, 3.0 and 4').
A supplemental analysis (Refer-ence 34.0) demonstrates that a 0.03 CPR addition is applicable for the SVEA-96 LFA fuel.
~
The postulated pump seizure
- accident, evaluated for SLO condi-
- tions, is calculated to have a less severe radiological release than the LOCA.
The radiological consequences of this postu-lated accident are bounded by the radiological evaluation per-formed by GE for the LOCA and are well within the 10CFR100 limits.
~
The single loop ECCS analysis supports the use of the WNP-2 two loop HAPLHGR limits for ANF 8x8 fuel when the reactor is operating in the SLO mode consistent with the single loop HCPR Operation limit.
Single loop operation of WNP-2 with the two loop ANF Bx8 fuel MAPLHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S.
NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.
The transient and pump seizure accident analyses are described in Reference 24.0 and the LOCA analyses are described in Reference 25.0.
With a single recirculation loop in operation, the GE initial core analyses supported continued operation with an increase of 0.01 in the MCPR safety limit (Reference 23.0).
ANF performed a single loop MCPR safety limit calculation and found that less than one tenth of one percent of the rods to be in boiling transition which supports a
HCPR safety limit of 1.07.
Because of the similarity between the ANF and GE fuel types making up the core, and because of the simi-larity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value can be used for operation with ANF fuel and single loop analyses.
For Cycle 6 operation with both recirculation loops in operation, the HCPR safety limit is 1.06, which is the same value as was used for the previous cycles.
For Cycle 6 operation with a single recir-culation loop in service, the MCPR safety limit is 1.07, which is also the same value used for the previous cycles.
The LFA single loop operation limits are bounded by the two-loop operation limits (References 1.0, 3.0 and 4.0).
9.7 n
~ F w
Reference 26.0 presents a final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2.
This analysis has beem approved and is applicable to future WNP-2 fuel cycles.
The FFTR analysis was performed for a 65'F temperature reduction.
This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature.
The FFTR analysis results show that CPR changes for the LRNB and FWCF transients of
+ 0.02 and - 0.01 respectively, are applicable to these respective anticipated operational occurrence (AOO) events.
That is, these LRNB and FWCF limit changes are applicable when Cycle 6 reactor operation is being extended with thermal coastdown at FFTR condi-tions and are applicable to all fuel designs except the SVEA-96 LFA fuel (References 1.0, 3.0 and 4.0).
A supplemental analysis (Refer-ence 34.0) demonstrates that a 0.03 CPR addition is applicable for the SVEA-96 LFA fuel.
~
The postulated pump seizure accident, evaluated for SLO condi-
- tions, is calculated to have a less severe radiological release
'han the LOCA.
The radiological consequences of this postu-lated accident are bounded by the radiological evaluation per-formed by GE for the LOCA and are well within the 10CFR100 limits.
The single loop ECCS analysis supports the use of the WNP-2 two loop MAPLHGR limits for ANF 8x8 fuel when the reactor is operating in the SLO mode consistent with the single loop MCPR Operation limit.
Single loop operation of WNP-2 with the two loop ANF 8x8 fuel MAPLHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S.
NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including. the double-ended severance of,a reactor coolant pipe.
The transient and pump seizure accident analyses are described in Reference 24.0 and the LOCA analyses are described fn Reference 25.0.
With a single recirculation loop in operation, the GE initial core analyses supported continued operation with an increase of 0.01 in the MCPR safety limit (Reference 23.0).
ANF performed a single loop MCPR safety limit calculation and found that less than one tenth of one percent of the rods to be in boiling transition which supports a
MCPR safety limit of 1.07.
Because of the similarity between the ANF and GE fuel types making up the core, and because of the simi-larity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value can be used for operation with ANF fuel and single loop analyses.
For Cycle 6 operation with both recirculation loops in operation, 'the MCPR safety limit is 1.06, which is the same value as was used for the previous cycles.
For Cycle,6 operation with a single recir-culation loop in service, the MCPR safety limit is 1.07, which is also the same value used for the previous cycles.
The LFA single loop operation limits are bounded by the two-loop operation limits (References 1.0, 3.0 and 4.0).
9.7 F
F w
m Reference 26.0 presents a final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2.
This analysis has beem approved and is applicable to future WNP-2 fuel cycles.
The FFTR analysts was performed for a 65'F temperature reduction.
This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature.
The FFTR analysis results show that CPR changes for the LRNB and FWCF transients of
+ 0.02 and 0.01 respectively, are applicable to these respective anticipated operational occurrence (AOO) events.
That is, these LRNB and FWCF limit changes are applicable when Cycle 6 reactor operation is being extended with thermal coastdown.at FFTR condi-tions and are applicable to all fuel designs except the SVEA-96 LFA fuel (References 1.0, 3.0 and 4.0).
A supplemental analysis (Refer-ence 34.0) demonstrates that a 0.03 CPR addition is applicable for the SVEA-96 LFA fuel.
1
The effects of the revised WNP-2 Cycle 6 core loading including 8
additional ANF assemblies on the system transient results and thermal limits as reported in References 1
and 2 was examined.
The limiting system transient at end of cycle is the load rejection without bypass.
This transient is a pressurization event which is most influenced by the core pressure reactivity coefficient and axial power distribution.
The core pressure coefficient and axial power distribution for the revised core loading with the 8 addi-tional assemblies was computed and compared against the values obt'ained for the reported Cycle 6 analysis, The pressure reactivity coefficient was calculated to be unchanged by the addition of the 8
assemblies, and the axial power distribution used for the initial Cycle 6 analysis was found to bound the results determined for the revised core loading.
On this basis, it is concluded that the reported system transient analyses for WNP-2 Cycle 6 remain appli-cable and conservative for the revised core loading including the 8
additional new assemblies.
No change in the thermal limits reported for Cycle 6 is required due to the revised core loading.
The limiting neutronic safety analyses including shutdown margin, control rod withdrawal, and control rod drop were evaluated.
The revised shutdown margin results are presented in Table 8.1.
For the control rod withdrawal, control rod drop and other safety analysis, the results reported in Reference 1 are applicable to the revised Cycle 6 core with 1S2 reload asemblies (Reference 33).
31
0
For Cycle 2, ANF analyzed the LOCA to determine MAPLHGR limits for ANF, Bx8 fuel.
The results of this analysis are presented in Reference 27.0.
These Cycle 2 results are equally applicable to Cycle 6.
ANF's methodo-logy for the LOCA analysis is given in References 28.0; 29.0, and 30.0.
'In addition, the Rod Drop Accident (RDA) was analyzed to demonstrate compliance with the 280 cal/gm design limit.
ANF's methodology for the RDA analysis can be found in Reference 15.0.
10.1 f
n i
n Reference 31.0 describes ANF's WNP-2 LOCA break spectrum analysis which defined the limiting break for WNP-2.
The analysis of this event for WNP-2 is described in Reference 32.0.
The LOCA analysis described in Reference 32.0 was performed for an entire core of ANF 8x8C fuel and therefore provides MAPLHGR 1'imits for ANF 8x8 fuel.
These results are applicable to operation in WNP-2 Cycle 6 (Refer-ence 1.0).
These MAPLHGR limits are also applicable to the GE11 LFA fuel and the SVEA-96 LFA fuel as discussed in Section 6.0.
MAPLHGR limits for ANF 9x9 LFA fuel are discussed in Reference 7.0.
10.2 ANF 8x8 reload fuel is hydraulically and neutronically compatible with the GE initial core fuel.
Therefore, the existing GE LOCA analysis and MAPLHGR limits are applicable to GE initial core fuel during Cycle 6 operation.
ANF's methodology for analyzing the Rod Drop Accident (RDA) is given in Reference 15.0.
For WNP-2 Cycle 6, the analysis shows a value of 97 cal/gm for the maximum deposited fuel rod enthalpy during the worst case postulated RDA (Reference 1.0).
This is well below the design limit value of 280 cal/gm.
10.3 in 1
To support operation of WNP-2 with a core composed of GE Cycle 1
fuel and ANF 8x8 reload fuel with a single recirculation pump operating, ANF recommends the conservative use of GE MAPLHGR limits for the GE fuel design with a multiplier of 0.84 applied for single loop operation.
The single loop ECCS analysis supports the use of the WNP-2 two loop MAPLHGR limits for ANF 8x8 fuel when the reactor is operating in the SLO mode.
Single loop operation of WNP-2 with the two loop ANF 8x8 fuel MAPHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S.
NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.
Two loop MAPLHGR values also apply to the LFA's as discussed in Section 6.0 and References 3.0, 4.0 and 7.0.
The Supply System has developed a restart physics test program to be carried out prior to or during Cycle 6 startup.
This program includes a
core loading verification test, a control rod functional test, an in sequence shutdown margin test, and a TIP asymmetry test.
The proposed test goals and a brief description of each test are given below.
V rifi i
T
~ To assure that the NNP-2 Cycle 6 Core is loaded according to the design analyzed by ANF.
~T~~Dm~r~g~ This test will be performed with the aid of a television camera.
A series of initial passes will be made with the fuel mast set at a predetermined height to assure that all fuel assemblies are fully seated in the core.
Then, with the aid of the camera and a visual readout on the refuel floor, the assembly serial
- numbers, their orientation and location will be visually checked and recorded on video tape.
Subsequently, a review of the tapes will be made to check the initial verification.
During the process, con-firmation will be made that the SVEA-96 and GE11 LFA's are in their designated core locations.
11.2 g~gQ i
T Qggl To determine and verify control rod mobility and functional i ty.
~T.s~.u.cri~i
- Following the completion of fuel loading, for each cell of four fuel assemblies, the control blade for that cell will be fully withdrawn and inserted.
This will demonstrate the mobility of that blade, the absence of overtravel for that blade and the fact that the lattice is subcritical with that blade withdrawn.
This in turn will verify that there are no gross reactivity discrepancies between the actual core and the analyzed design.
After the core is fully loaded, verify that the control rod drive insertion and withdrawal times are within design specifications and technical specification limits.
This action will also verify that the core" is subcri tical with any single rod fully withdrawn.
gay'. - To assure that the Technical Specification shutdown margin requirement is satisfied.
The data is taken during a normal insequence startup criticality.
Critical control rod positions are obtained and corrected for reactor period and moderator temperature coeffi-cient effects.
The results are compared to predicted control rod positions and from this information, the shutdown margin with the analytically determined strongest control rod withdrawn is confirmed.,
1
11 4 T mm r
g~l To assure proper TIP systems operation and to verify that the TIP system uncertainty is within the limits assumed for HCPR safety limit analysis.
ferably above 75 percent power.
An octant symmetric control rod pattern is utilized.
Data is gathered from all available TIP locations, and the total average uncertainty is determined for all symmetric TIP pairs.
1.0 ANF-90-02, "HNP-2 Cycle 6 Reload Analysis Report",
Advanced Nuclear Fuels Corporation, January 1990 2.0 ANF-90-01, "HNP-2 Cycle 6 Plant Transient Analysis Report",
Advanced Nuclear Fuels Corporation, January 1990 3.0 GEll, "Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No.
2 Reload 5 Cycle 6", General Electric Company December 1989 (Proprietary) 4.0 UK90 126, "Supplemental Lead Fuel Assembly Licensing Report-SVEA 96 LFAs for HNP-2",
ABB Atom, January 1990 (Proprietary) 5.0 XN-NF-80-19(A), Volume=4, Revision 1,
"Exxon Nuclear Methodology for Boiling Hater Reactor:
Applications of the ENC Methodology to BHR Reloads",
Exxon Nuclear
- Company, September 1983 6.0 HPPSS-EANF-101, "WNP-2 Cycle 2 Reload Summary Report",
February 1986 7.0 HPPSS-EANF-124, "WNP-2 Cycle 5 Reload
.Summary Report",
February 1989 8.0 ANF-89-158, "Washington Public Power Supply System, HNP-2 Reload ANF-5, Cycle 6 Fuel Design Report",
December 1989 9.0 XN-NF-81-21(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BHR Reload Fuel",
Exxon Nuclear
- Company, January 1982 10.0 XN-NF-81-21(A), Revision 1, Supplement 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BHR Reload Fuel",
Exxon Nuclear Company, March 1985 11.0 XN-NF-85-67(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BHR Reload Fuel",
Exxon Nuclear
- Company, September 1986 12.0 J. Lindner, Letter, "Flow Dependent MCPR Limit Implementation",
ABB 90-070, ABB Atom, February 16, 1990 13.0 J.T. Worthington, Letter and Attachments, "Seismic/LOCA Capability for Interactive Channels on HNP-2 GEll LFAs",
JTW 90-012, General Electric Company, January 19, 1990 14.0 XN-NF-524(A), Revision 1,
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactor",
Exxon Nuclear
- Company, November 1983 15.0 XN-NF-80-19(A), Volume 1
and Volume 1 Supplements 1
and 2, "Exxon Nuclear Methodology for Boiling Hater Reactor:
Neutronics Methods for Design and Analysis",
Exxon Nuclear
- Company, March 1983 0
16.0 ANF-90-006(P),
"WNP-2 Cycle 6 Fuel Cycle'esign Report", Revision 1,
Advanced Nuclear Fuels Corporation, April 1990 r
17.0 XN-NF-79-71(P), Revision 2 (as supplemented),
"Exxon Nuclear Power Plant Transient Methodology",
Exxon Nuclear.Company, November 1981 18.0'N-NF-80-19(A),
Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors:
THERHEX Thermal Limits Methodology Summary Descriptions",
Exxon Nuclear
- Company, January 1987 19.0 XN-NF-84-105(A), Volume 1, Volume 1 Supplement 1,
Volume 1
Supplement 2,
"X-COBRA-T:
A Computer Code for BWR Transient Thermal Hydraulic Core Analysis", Advanced Nuclear Fuels Corporation, February 1987 20.0 J.B.
Edgar, Letter to Washington Public Power Supply System, Supplemental Ana.lysis Results, ENWP-86-007, Exxon Nuclear
- Company, April 15, 1986 21.0 J.B.
Edgar, Letter to Washington Public Power Supply System, ENWP-86-0033, Exxon Nuclear
- Company, February 13, 1986 22.0 NEDC-31107, "Safety Review of Washington Public Power Supply System Nuclear Project No.
2 at Core Flow Conditions Above Rated Flow Throughout Cycle 1
and Final Feedwater Temperature Reduction",
General Electric Company, February 1986 23.0 "Final Safety Analysis Report, Washington Public Power Supply System Nuclear Project No. 2",
as reviewed through Amendment 35, November 1984 24.0 ANF-87-119, "WNP-2 Single Loop Operation Analysis", Advanced Nuclear Fuels Corporation, September 1987 25.0 ANF-87-118, "WNP-2 LOCA Analysis for Single Loop Operation",
Advanced Nuclear Fuels Corporation, September 1987 26.0 XN-NF-87-92, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction",
Advanced Nuclear Fuels Corporation, September 1987 27.0 XN-NF-86-01, Revision 1,
"WNP-2 Cycle Exxon Nuclear
- Company, February 1986 2 Reload Analysis Report",
28.0 XN-NF-80-19(A), Volumes 2, 2A, 2B and 2C, "Exxon Nuclear Methodology for Boiling Water Reactor:
EXEH ECCS Evaluation Model", Exxon Nuclear
- Company, September 1982 29.0 XN-NF-CC-33(A), Revision 1,
"HUXY:
A Generalized Hultirod Heatup Code With 10CFR50, Appendix K, Heatup Options",
Exxon Nuclear
- Company, November 1975 0
1I
30.0 XN-NF-82-07(A), Revision 1,
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