ML17289A722: Difference between revisions

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At less than rated conditions, the MCPR limit is the maximum of the rated power MCPR limit, the reduced power MCPR limit, and the reduced Washington Nuclear-Unit 2 COLR 92-8 Rev.0 920701 flow MCPR limit.This stipulation assures that the safety limit MCPR will not be violated throughout the WNP-2 operating regime.The LHGR limits for the GEl1 LFA's are the same as the SNP 8x8 reload fuel, except that a ratio ([64-2]/[81-7])
At less than rated conditions, the MCPR limit is the maximum of the rated power MCPR limit, the reduced power MCPR limit, and the reduced Washington Nuclear-Unit 2 COLR 92-8 Rev.0 920701 flow MCPR limit.This stipulation assures that the safety limit MCPR will not be violated throughout the WNP-2 operating regime.The LHGR limits for the GEl1 LFA's are the same as the SNP 8x8 reload fuel, except that a ratio ([64-2]/[81-7])
is applied to account for the different number of fuel pins in the two designs.The LHGR limits for the SVEA-96 LFA's are taken directly from Reference 5.4.Preparation, review and approval of this report were performed in accordance with applicable Supply System procedures.
is applied to account for the different number of fuel pins in the two designs.The LHGR limits for the SVEA-96 LFA's are taken directly from Reference 5.4.Preparation, review and approval of this report were performed in accordance with applicable Supply System procedures.
The specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis are referenced in Section 5.0 2.0 Average Planar Linear Heat Generation Rate (APLHGR)Limits for Use in Technical Specification 3.2.1 The APLHGR's for use in Technical Specification
The specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis are referenced in Section 5.0 2.0 Average Planar Linear Heat Generation Rate (APLHGR)Limits for Use in Technical Specification 3.2.1 The APLHGR's for use in Technical Specification 3.2.1 shall not exceed the limits shown in Figures 1, 2, 4, and 5 when in two-loop operation.and in Figures 1, 3, 4, and 5 when in single loop operation.
 
====3.2.1 shall====
not exceed the limits shown in Figures 1, 2, 4, and 5 when in two-loop operation.and in Figures 1, 3, 4, and 5 when in single loop operation.
The limits for each fuel type as a function of Average Planar Exposure are provided for the Siemens Nuclear Power fuel, including the SNP LFA's, the SVEA-96 LFA fuel, and the GE11 LFA fuel.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 HRIIIR~I=L+aHHHIHHHRh5-:~=IIIIIIIEIIIIIIII
The limits for each fuel type as a function of Average Planar Exposure are provided for the Siemens Nuclear Power fuel, including the SNP LFA's, the SVEA-96 LFA fuel, and the GE11 LFA fuel.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 HRIIIR~I=L+aHHHIHHHRh5-:~=IIIIIIIEIIIIIIII
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.,11 555=11111111illl 111 11111111 920701 3.0 MinimUm Critical Power Ratio (MCPR)Limit for Use in Technical Specification 3.2.3 The MCPR limit for use in Technical Specification
.,11 555=11111111illl 111 11111111 920701 3.0 MinimUm Critical Power Ratio (MCPR)Limit for Use in Technical Specification 3.2.3 The MCPR limit for use in Technical Specification 3.2.3 shall be: a)Greater than or equal to the greater of the limits determined from Tables la and 1b and Figures 6 and 7a through 14b.b)The full power limit is determined at 104%power and 106%core flow.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 Table la WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures (4244 MWd/MT Condition Limit Nss'" Full Power Flow Dependent SLMCPR~1,07+SNP 8xg SNP 9x9 SNP 9x9 SVEA-96 GF.11 LFA LFA 1.23+1.23+1,25 136+Figure 6 Tsss~n Power Dependent Fig.7a Fig.7a Fig.8a Fig, 7a Full Power Flow Depcndcnt 1.24 1.25 1.32 1.37 Figure 6 Nssu~Power Dependent Fig.9a Fig, 9a Fig.10a Fig.9a RFI'ull Power Inoperable Flow Dependent Power Dependent SLO NSS Full Power Flow Depcndcnt 1.26 1.28 1.37 1.40 Figure 6 Fig.13a Fig.13a Fig.14a Fig.13a 186 1.36 1:36 1,$5 None Power Dependent Fig.7a Fig.7a Fig.8a Fig.7a SLO~VSSS Full Power Flow Dependent 186 1.36 136 1.85 None Power Dependent Fig.9a Fig.9a Fig.10a Fig.9a SLOP NSS RFr Full Power Inoperable Flow Dependent 186 136 1.36 1.85 None Power Dependent Fig.13a Fig, 13a Fig.14a Fig, 13a'ashington Nuclear-Unit 2-9-COLR 92-8 Rev.0 92%01 Table lb WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures>4244 MWd/MT SLMCPR~1.07m SLMCPR~1.07 FFfR Condition Limit Nssu)Full Power Flow Dependent Power Dependent TSSS<'>1.24 1.25 192 137 Figurc 6 1.26 1.27 1.34 1.40 Figure 6 Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 SNP sx8 SNP 9x9 SNP 9x9 SVBA-96 SNP 8x8 SNP 9x9 SNP 9x9 SVBA-96 GB11 LFA" LFA GB11 LFA LFA NSSto Full Power Flow Dependent Power Dependent 1.28 1.29 138 1.43 Figure 6 Fig, 9b Fig.9b Fig.10b Fig.9b Not Analyzed RFI'ull Power inoperable Flow Dependent Power Dependent SLO(" NSS 1.31 1.33 1.46 1.48~Figurc 6 Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Full Power Flow Dependent Power Dependent SLOmTSSS 186 1.36 1.36 None 1.85-1.56 1.36 136 1.85 Noae Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 Full Power Flow Dependent Power Depcadcat SLOP NSS RPT Full Power Inoperable Flow Dependent Power Dcpeadent 1.56 1.36 1.36 1.85 None Fig.9b Fig.9b Fig.10b Fig.9b 156 136 136 1.85 None Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Not Analyzed W'ashington Nuclear-Unit 2-10-COLR 92-8 Rev.0 Notes for Table 1 Note 1: These MCPR values are based on the SNP Reload Safety Analysis performed using the control rod insertion times shown below (defined as normal scram speed: NSS).In the event that Surveillance 4.1.3.2 shows these scram insertion times have been exceeded, the plant thermal margin limits associated with NSS default to the values associated with the Technical Specification scram speed (TSSS).The scram insertion times must meet the requirements of Technical Specification 3.1.3.4.Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5 Slowest measured average control rod insertion times t specified notches for all operable control rods for each grou of four control rods arranged in a two-by-two array (seconds)0.380 0.690 1.500 2.750 Note 2: For Single Loop Operation, the SLMCPR increases by 0.01.Note 3: The control rod withdrawal error (CRWB)analysis was performed with the nominal rod block monitor (RBM)setting value of 1.06.Use of the nominal setpoint is in accordance with the methodology described in Reference 5.12, consistent with approved industry practice.CRWH is limiting for the noted full power limits for cycle exposures less than 4244 MWd/MT.The load rejection without bypass (LRNB)event is limiting for the remaining full power events.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 f%g gg g%%%g g%m%wow~~:~aaaaaaaaaaaa I I h<I I i i EE hi<I~El I El 3 3333333333333
 
====3.2.3 shall====
be: a)Greater than or equal to the greater of the limits determined from Tables la and 1b and Figures 6 and 7a through 14b.b)The full power limit is determined at 104%power and 106%core flow.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 Table la WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures (4244 MWd/MT Condition Limit Nss'" Full Power Flow Dependent SLMCPR~1,07+SNP 8xg SNP 9x9 SNP 9x9 SVEA-96 GF.11 LFA LFA 1.23+1.23+1,25 136+Figure 6 Tsss~n Power Dependent Fig.7a Fig.7a Fig.8a Fig, 7a Full Power Flow Depcndcnt 1.24 1.25 1.32 1.37 Figure 6 Nssu~Power Dependent Fig.9a Fig, 9a Fig.10a Fig.9a RFI'ull Power Inoperable Flow Dependent Power Dependent SLO NSS Full Power Flow Depcndcnt 1.26 1.28 1.37 1.40 Figure 6 Fig.13a Fig.13a Fig.14a Fig.13a 186 1.36 1:36 1,$5 None Power Dependent Fig.7a Fig.7a Fig.8a Fig.7a SLO~VSSS Full Power Flow Dependent 186 1.36 136 1.85 None Power Dependent Fig.9a Fig.9a Fig.10a Fig.9a SLOP NSS RFr Full Power Inoperable Flow Dependent 186 136 1.36 1.85 None Power Dependent Fig.13a Fig, 13a Fig.14a Fig, 13a'ashington Nuclear-Unit 2-9-COLR 92-8 Rev.0 92%01 Table lb WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures>4244 MWd/MT SLMCPR~1.07m SLMCPR~1.07 FFfR Condition Limit Nssu)Full Power Flow Dependent Power Dependent TSSS<'>1.24 1.25 192 137 Figurc 6 1.26 1.27 1.34 1.40 Figure 6 Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 SNP sx8 SNP 9x9 SNP 9x9 SVBA-96 SNP 8x8 SNP 9x9 SNP 9x9 SVBA-96 GB11 LFA" LFA GB11 LFA LFA NSSto Full Power Flow Dependent Power Dependent 1.28 1.29 138 1.43 Figure 6 Fig, 9b Fig.9b Fig.10b Fig.9b Not Analyzed RFI'ull Power inoperable Flow Dependent Power Dependent SLO(" NSS 1.31 1.33 1.46 1.48~Figurc 6 Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Full Power Flow Dependent Power Dependent SLOmTSSS 186 1.36 1.36 None 1.85-1.56 1.36 136 1.85 Noae Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 Full Power Flow Dependent Power Depcadcat SLOP NSS RPT Full Power Inoperable Flow Dependent Power Dcpeadent 1.56 1.36 1.36 1.85 None Fig.9b Fig.9b Fig.10b Fig.9b 156 136 136 1.85 None Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Not Analyzed W'ashington Nuclear-Unit 2-10-COLR 92-8 Rev.0 Notes for Table 1 Note 1: These MCPR values are based on the SNP Reload Safety Analysis performed using the control rod insertion times shown below (defined as normal scram speed: NSS).In the event that Surveillance 4.1.3.2 shows these scram insertion times have been exceeded, the plant thermal margin limits associated with NSS default to the values associated with the Technical Specification scram speed (TSSS).The scram insertion times must meet the requirements of Technical Specification 3.1.3.4.Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5 Slowest measured average control rod insertion times t specified notches for all operable control rods for each grou of four control rods arranged in a two-by-two array (seconds)0.380 0.690 1.500 2.750 Note 2: For Single Loop Operation, the SLMCPR increases by 0.01.Note 3: The control rod withdrawal error (CRWB)analysis was performed with the nominal rod block monitor (RBM)setting value of 1.06.Use of the nominal setpoint is in accordance with the methodology described in Reference 5.12, consistent with approved industry practice.CRWH is limiting for the noted full power limits for cycle exposures less than 4244 MWd/MT.The load rejection without bypass (LRNB)event is limiting for the remaining full power events.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 f%g gg g%%%g g%m%wow~~:~aaaaaaaaaaaa I I h<I I i i EE hi<I~El I El 3 3333333333333
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====3.2.4 shall====
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Revision as of 06:58, 6 May 2019

Rev 0 to Washington Nuclear Plant 2 Cycle 8,Core Operating Limits Rept.
ML17289A722
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 06/30/1992
From:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17289A721 List:
References
COLR-92-8, COLR-92-8-R, COLR-92-8-R00, NUDOCS 9207150308
Download: ML17289A722 (41)
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Text

COLR 92-8 Rev.0 Controlled Copy No.9207 0l WNP-2 Cycle 8 Core Operating Limits Report June 1992 Washington Public Power Supply System 9207150308 92070b DOGY 05000397 PDR A PDg P-I 1 r WNP-2 Cycle 8 Core Operating Limits Report 920701 List of Effective Pages~Pa e 1 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35~Reviei 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0~0 0 0 0 0 0 0 0 0 0 0 0 0 I

920701 WNP-2 Cycle 8 Core Operating Limits Report Table of Contents~Pa~e 1.0 Introduction and Summary..................................

1 2.0 Average Planar Linear Heat Generation Rate (APLHGR)Limits for Use in Technical Specification 3.2.1.................................

2 e'3.0 Mnmnum Critical Power Ratio (MCPR)Limit for Use in Technical S 1J~g pecuicahon 3.2.3.......................................

8 4.0 Linear Heat Generation Rate (LHGR)Limit for Use in Technical S 1J~g\pecuicabon3.2.4

....................,.............,....

27 5..0 References

.......................................

~....33 W'ashington Nuclear-Unit 2 COLR 92-8 Rev.0 I

920701 1.0 Introduction and Suxnlnary This report provides the Average Planar Linear Heat Generation Rate (APLHGR)limits, the Mnimum Critical Power Ratio (MCPR)limits, and the Linear Heat Generation Rate (LHGR)limits for WNP-2, Cycle 8 as required by Technical Specification 6.9.3.1.As required by Technical Specifications 6.9.3.2 and 6.9.3.3, these limits were determined using NRC-approved methodology and are established so that all applicable limits of the plant safety analysis are met.The thermal limits for SNP fuel given in this report are documented in the Cycle 8 Plant Transient Analysis Report (Reference 5.1), the Cycle 8 Reload Analysis Report (Reference 5.2), and the Power Dependent MCPR Limits for WÃP-2 Cycle 8 letter report (Reference 5.9).The thermal limits determined through the approved methodology are modified for the GE11 and SVEA-96 LFA's as discussed below.The WNP-2 Cycle 8 reload will include four Siemens Nuclear Power (SNP), four General Electric (GE), and four ABB Atom (ABB)Lead Fuel Assemblies (LFA's).The four SNP LFA's were inserted during the reload for Cycle 5.The four GE and ABB LFA's were inserted at the beginning of Cycle 6 and were designed to be compatible with the reload fuel utilized in Cycle 6.The Supply System will load the LFA's in core locations which, according to analysis, have sufficient margin such that the LFA's are not expected to be the limiting assemblies.

This approach is intended to exclude the LFA's ever being the limiting fuel assemblies on either a nodal or an assembly power basis.The GE11 LFA is described in the GEll Lead Fuel Assembly Report for washington Public Power Supply System Nuclear Project No.2, Reload 5, Cycle 6 (Reference 5.3).This reference describes the design goals of the GE11 LFA's and provides support for monitoring the GE11 LFA's at thermal limits based on the SNP 8x8 reload fuel thermal limits.The SVEA-96 LFA is described in the Supplemental LFA Licensing Report-SVZA-96LFA

's for TVÃP-2 (Reference 5.4).The process for developing thermal limits for the SVEA-96 LFA fuel based upon the SNP 8x8 reload fuel thermal limits is described in this reference and Reference 5.5.The MAPLHGR limits for the GE11 LFA's are the same as the SNP 8x8 reload fuel, except that a ratio ([64-2]/[81-7])

is applied to account for the different number of fuel pins in the two designs.The MAPLHGR limits for the SVEA-96 LFA's are the same as the SNP 8x8 reload fuel, except that a ratio ([64-2]/[100-4])

is applied to account for the different number of fuel pins in the two designs.Furthermore, the MAPLHGR limits for the SVEA-96 LFA's are multiplied 1)by 1.04 to account for a different estimation of the local power in the output from POWERPLEK compared to ABB'Atom methods and 2)by 1.02 to account for a different estimation of exposure in the output from POVFEM'LEX compared ABB Atom methods.This produces a combined multiplier of 1.06.A power dependent MCPR is specified in this report to define operating limits at other than rated power conditions.

For the WNP-2 core, feedwater-controller-failure transients from reduced power are calculated to be more severe than from full power conditions.

A flow dependent MCPR is specified in this report to define operating limits at other than rated flow conditions.

The reduced flow MCPR operating limit provides bounding protection for the limiting recirculation flow increase transient.

At less than rated conditions, the MCPR limit is the maximum of the rated power MCPR limit, the reduced power MCPR limit, and the reduced Washington Nuclear-Unit 2 COLR 92-8 Rev.0 920701 flow MCPR limit.This stipulation assures that the safety limit MCPR will not be violated throughout the WNP-2 operating regime.The LHGR limits for the GEl1 LFA's are the same as the SNP 8x8 reload fuel, except that a ratio ([64-2]/[81-7])

is applied to account for the different number of fuel pins in the two designs.The LHGR limits for the SVEA-96 LFA's are taken directly from Reference 5.4.Preparation, review and approval of this report were performed in accordance with applicable Supply System procedures.

The specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis are referenced in Section 5.0 2.0 Average Planar Linear Heat Generation Rate (APLHGR)Limits for Use in Technical Specification 3.2.1 The APLHGR's for use in Technical Specification 3.2.1 shall not exceed the limits shown in Figures 1, 2, 4, and 5 when in two-loop operation.and in Figures 1, 3, 4, and 5 when in single loop operation.

The limits for each fuel type as a function of Average Planar Exposure are provided for the Siemens Nuclear Power fuel, including the SNP LFA's, the SVEA-96 LFA fuel, and the GE11 LFA fuel.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 HRIIIR~I=L+aHHHIHHHRh5-:~=IIIIIIIEIIIIIIII

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.,11 555=11111111illl 111 11111111 920701 3.0 MinimUm Critical Power Ratio (MCPR)Limit for Use in Technical Specification 3.2.3 The MCPR limit for use in Technical Specification 3.2.3 shall be: a)Greater than or equal to the greater of the limits determined from Tables la and 1b and Figures 6 and 7a through 14b.b)The full power limit is determined at 104%power and 106%core flow.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 Table la WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures (4244 MWd/MT Condition Limit Nss'" Full Power Flow Dependent SLMCPR~1,07+SNP 8xg SNP 9x9 SNP 9x9 SVEA-96 GF.11 LFA LFA 1.23+1.23+1,25 136+Figure 6 Tsss~n Power Dependent Fig.7a Fig.7a Fig.8a Fig, 7a Full Power Flow Depcndcnt 1.24 1.25 1.32 1.37 Figure 6 Nssu~Power Dependent Fig.9a Fig, 9a Fig.10a Fig.9a RFI'ull Power Inoperable Flow Dependent Power Dependent SLO NSS Full Power Flow Depcndcnt 1.26 1.28 1.37 1.40 Figure 6 Fig.13a Fig.13a Fig.14a Fig.13a 186 1.36 1:36 1,$5 None Power Dependent Fig.7a Fig.7a Fig.8a Fig.7a SLO~VSSS Full Power Flow Dependent 186 1.36 136 1.85 None Power Dependent Fig.9a Fig.9a Fig.10a Fig.9a SLOP NSS RFr Full Power Inoperable Flow Dependent 186 136 1.36 1.85 None Power Dependent Fig.13a Fig, 13a Fig.14a Fig, 13a'ashington Nuclear-Unit 2-9-COLR 92-8 Rev.0 92%01 Table lb WNP-2 Cycle 8 MCPR Operating Conditions Cycle Exposures>4244 MWd/MT SLMCPR~1.07m SLMCPR~1.07 FFfR Condition Limit Nssu)Full Power Flow Dependent Power Dependent TSSS<'>1.24 1.25 192 137 Figurc 6 1.26 1.27 1.34 1.40 Figure 6 Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 SNP sx8 SNP 9x9 SNP 9x9 SVBA-96 SNP 8x8 SNP 9x9 SNP 9x9 SVBA-96 GB11 LFA" LFA GB11 LFA LFA NSSto Full Power Flow Dependent Power Dependent 1.28 1.29 138 1.43 Figure 6 Fig, 9b Fig.9b Fig.10b Fig.9b Not Analyzed RFI'ull Power inoperable Flow Dependent Power Dependent SLO(" NSS 1.31 1.33 1.46 1.48~Figurc 6 Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Full Power Flow Dependent Power Dependent SLOmTSSS 186 1.36 1.36 None 1.85-1.56 1.36 136 1.85 Noae Fig.7b Fig.7b Fig.8b Fig.7b Fig.11 Fig.11 Fig.12 Fig.11 Full Power Flow Dependent Power Depcadcat SLOP NSS RPT Full Power Inoperable Flow Dependent Power Dcpeadent 1.56 1.36 1.36 1.85 None Fig.9b Fig.9b Fig.10b Fig.9b 156 136 136 1.85 None Fig.13b Fig.13b Fig.14b Fig.13b Not Analyzed Not Analyzed W'ashington Nuclear-Unit 2-10-COLR 92-8 Rev.0 Notes for Table 1 Note 1: These MCPR values are based on the SNP Reload Safety Analysis performed using the control rod insertion times shown below (defined as normal scram speed: NSS).In the event that Surveillance 4.1.3.2 shows these scram insertion times have been exceeded, the plant thermal margin limits associated with NSS default to the values associated with the Technical Specification scram speed (TSSS).The scram insertion times must meet the requirements of Technical Specification 3.1.3.4.Position Inserted From Fully Withdrawn Notch 45 Notch 39 Notch 25 Notch 5 Slowest measured average control rod insertion times t specified notches for all operable control rods for each grou of four control rods arranged in a two-by-two array (seconds)0.380 0.690 1.500 2.750 Note 2: For Single Loop Operation, the SLMCPR increases by 0.01.Note 3: The control rod withdrawal error (CRWB)analysis was performed with the nominal rod block monitor (RBM)setting value of 1.06.Use of the nominal setpoint is in accordance with the methodology described in Reference 5.12, consistent with approved industry practice.CRWH is limiting for the noted full power limits for cycle exposures less than 4244 MWd/MT.The load rejection without bypass (LRNB)event is limiting for the remaining full power events.Washington Nuclear-Unit 2 COLR 92-8 Rev.0 f%g gg g%%%g g%m%wow~~:~aaaaaaaaaaaa I I h<I I i i EE hi<I~El I El 3 3333333333333

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k kk g'%g kkk kkk II IL ARR III kkk kkkk la kk kkk~kk~kk~k k Qaa Ila a~al kh>kW k 5 k I aa I~'i OS~I k II~I laa laal 55M~~kkk kk k 5 aaaI IL lla aa IaaaII laI'Iaa ll lllahmla Iaal L I I haa<III I I I~~~I I k k k ak ilk SQS Slays/SSS SSS~ak ASS SSS SSS XSS ASS SSAII~ka SSS~ka SSS SSS lll SSS akk SSS SSS SSS S~SS S S kaa Illa a~kk kkk~~kk Lkgk kkk SSS SSh Q g S~ll lla F.SS SSS SOS I~~~I RES~5IINIIHI SSS SS SS SWISS SSS SSS I'o c a SS SS ak S ak~ak Iak~kl SSS~ak SSS SSS a l'l'a Ilk kl~kk a SS SSS~a~S SS SSa q S kkkgk kk~ak SSS S S R S S S~I~a:i ala ASS Si~kl I SSS SSS SS SS~SS SS i(k kkk ha'a SSS ASS SSSS'a ll S ii SS SSSS SSS 920701 4.0 Linear Heat Generation Rate (LHGR)Limit for Use in Technical Specification 3.2.4 The LHGR limit for use in Technical'Specification 3.2.4 shall not exceed the values shown in Figures 15, 16, 17, 18, and 19.Washington Nuclear-Unit 2-27-COLR 92-8 Rev.0 DIES ESgSSEgE Smg EpE 0 8 SS I Iil ill Ill Ill I a ae E~1 5555555a555

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92070l 5.0 References 5.1 EMF-92-039, Revision 1, WNP-2 Cycle 8 Plant Transient Analysis, Siemens Nuclear Power Corporation, June 1992.5.2 EMF-92-040, Revision 1, WNP-2 Cycle 8 Reload Analysis, Siemens Nuclear Power Corporation, June 1992.5.3 GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No.2, Reload 5, Cycle 6, General Electric Company, December 1989 (Proprietary).

5.4 UK 90-126, Supplemental Lead Fuel Assembly Licensing Report-SVE'A-96LFA's for WNP-2, ABB Atom, January 1990 (Proprietary).

5.5 ATOP-91-120, W.R.Harris, ABB, to D.L.Whitcomb, Supply System, Assembly Treatment in WNP-2 Cycle 7 Core Operating Limits Report, May 1, 1991.5.6 SNPWP-92-0059, Udell Fresk, Siemens Nuclear Power Corporation, to R.A.Vopalensky, Supply System, Comments on WNP-2 Cycle 8 Draft COLR Report, May 21, 1992.5.7 5.8 JTW:92-087, J.T.Worthington, General Electric Company, to D.L.Whitcomb, Supply System, WNP-2 Cycle 8 Core Operating Limits Report, Contract No.C-21099, Gall Lead Fuel Assemblies, May 22, 1992.ATOP-92-062, W.R.Harris, ABB Atom, to D.L.Whitcomb, Supply System, SVE'A-96 Lead Fuel Assembly Treatment in'WNP-2 Cycle 8 Core Operating Limits Report, May 15, 1992.5.9 SNPWP-92-0072, Udell Fresk, Siemens Nuclear Power Corporation, to R.A.Vopalensky, Power Dependent MCPR Limits for WNP-2 Cycle 8, June 24, 1992.5.10 ANF-89-014(P)(A), Revision 1 and Supplement 1&2, Generic Mechanical Design for Advanced Nuclear Fuels 9'-1X and 9x9-9X Reload Fuel, Advanced Nuclear Fuels Corporation, Richland, WA, October 1991.5.11 XN-NF-79-71(P), Revision 2, including Supplements 1,2 and 3(A), Enon Nuclear Plant Transient Methodology for Boiling Water Reactors, Exxon Nuclear Company, Inc., Richland, WA, November 1981.5.12 XN-NF-80-19(P)(A), Volume 1 Supplement 1 and 2, Exron Nuclear Methodology for Boiling Water Reactors Neutronic Methods for Design Analysis, March 1983.Washington Nuclear-Unit 2-33-COLR 92-8 Rev.0 5.13 i 92%01 XN-NF-80-19(P)(A), Volume 1 Supplements 3 and 4, Exxon Nuclear Methodology for Boiling Water Reactors Neutronic Methods for Design Analysis, Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.5.14 XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors THERMEK Thermal Limits Methodology Summary Description, Exxon Nuclear Company, Inc., Richland, WA, January 1987.5.15 XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, Inc., Richland, WA, June 1986.5.16 ANF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2, 3, and 4, COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.5.17 ANF-524(P)(A), Revision 2 and Supplements 1 and 2, Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.5.18 ANF-1125(P)(A) and Supplements 1 and 2, ANFB Crirical Power Correlation, Advanced Nuclear Fuels Corporation, Richland, WA, April 1990 5.19 5.20 5.21 Letter, R.C.Jones (NRC)to R.A.Copeland (ANF), NRC Approval of ANFB Additive Constants for 5%9-5K BWR Fuel, November 14, 1990.Letter ENWB-96-0067, J.B.Edgar (ANF)to Supply System, Supplemental Licensing Analysis Results, April 15, 1986.ANF-90-01, WNP-2 Cycle 6 Plant Transient Analysis, Advanced Nuclear Fuels Corporation, Richland, WA, January 1990.5.22 XN-NF-84-105(P)(A), Volume 1 and Supplements 1, 2,&4, XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA, February 1987.5.23 XN-NF-81-21(P)(A), Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Relouf Fuel, Exxon Nuclear Company, Inc., Richland, WA, September 1982, and Supplement 1, March 1985.5.24 XN-NF-85-67(P)(A), Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, Inc., Richland, WA, September 1986.Washington Nuclear-Unit 2-34-COLR 92-8 Rev.0 5.25 TAC M83319, Docket 50-397, Proposed Bx8 Fuel Burnup Extension for Cycle 8-WNP-2 PAC No.M83319), R.R.Assa, NRC, to G.C.Sorensen, July 1, 1992.5.26 XN-NF-81-58(A), Revision 2, RODEX2: Fuel Rod Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA, March 1984.5.27 XN-NF-87-92 and Supplement 1, WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction, Advanced Nuclear Fuels Corporation, Richland, WA, June 1987'and May 1988.5.28 ANF-87-119, WNP-2 Single Loop Operation Analysis, Advanced Nuclear Fuels Corporation, Richland, WA, September 1987.5.29 ANF-87-118, WNP-2 LOCA Analysis For Single Loop Operation, Advanced Nuclear Fuels Corporation, Richland, WA, September 1987.5.30 Letter, R.B.Samworth, USNRC, to G.C.Sorensen, Supply System, Issuance of Amendment No.62 to Facility Operating License No.NPF-21-WPPSS Nuclear Project 2 (TAC No.67538), August 5, 1988.5.31 XN-NF-85-138(P),LOCA Break Spectrum for a BWR 5, Exxon Nuclear Company, Inc., Richland, WA, December 1985.5.32 XN-NF-85-139, WNP-2 LOCA-ECCS Analysis, MAPLHGR Results, Exxon Nuclear Company, Inc., Richland, WA, December 1984.5.33 ANF-CC-33(P)(A), Supplement 2, HLXY: A Generalized Multirod Heatup Code with 10 CFR SOAppendix ZHeatup Option, Advanced Nuclear Fuels Corporation, Richland, WA, January 1991.5.34 XN-NF-81-22(P)(A), Generic Statistical Uncertainty Analysis Methodology, November 1983.5.35 NEDE-24011-P-A-6, General Elecmc Standard Application for Reactor Fuel, April 1983.Washington Nuclear-Unit 2-35-COLR 92-8 Rev.0

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