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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML17284A9121999-10-13013 October 1999 Proposed Tech Specs 3.3.6.1,Table 3.3. 6.1-1, Primary Containment Isolation Instrumentation. ML17284A8521999-07-29029 July 1999 Proposed Tech Specs 3.4.9, RHR Shutdown Cooling Sys - Hot Shutdown. ML17284A8491999-07-29029 July 1999 Proposed Tech Specs,Revising SR 3.5.2.2 Re Condensate Storage Tank Water Level ML17284A8461999-07-29029 July 1999 Proposed Tech Specs,Revising Table 3.3.5.1-1, ECCS Instrumentation Items 1.a,2.a,4.a & E.A. ML17284A8421999-07-29029 July 1999 Proposed Tech Specs Revising SR of TS 3.8.4, DC Sources - Operating & SR 3.8.5.1 of TS 3.8.5, DC Sources - Shutdown. ML17333A0021999-04-20020 April 1999 Proposed Tech Specs Section 3.4.11,replacing Existing Reactor Pressure Temp Limit Curves by 000630 ML17292B6341999-04-0707 April 1999 Proposed Tech Specs Modifying MCPR Safety Limits to Allow Continued Power Operation at Plant Following Restart from R-14 RFO ML17292B4881998-12-17017 December 1998 Proposed Tech Specs SR 3.8.1.8,allowing Capability to Manually Transfer Between Preferred & Alternate Offsite Power Sources During Modes 1 & 2 by 990125 ML20198A7051998-11-30030 November 1998 Revs 8 Through 13 to TS Bases & Revs 12 Through 15 of Licensee Controlled Specs ML17284A7181998-08-0505 August 1998 Proposed Tech Specs SR 3.8.4.7,allowing Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 1 & 3,125 Vdc Batteries E-B1-1 & HPCS-B1-DG3 & Div 1,250 Vdc Battery E-B2-1 ML17284A7071998-07-17017 July 1998 Proposed Tech Specs Modifying SR 3.8.4.7 to Allow Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 2,125 Vdc,Battery E-B1-2 ML17292B2831998-03-0909 March 1998 Proposed Tech Specs Establishing Interim SLMCPR for Siemens Power Corp ATRIUM-9X Fuel Applicable to Cycle 14 Only ML17292B1321997-12-0404 December 1997 Proposed Tech Specs Modifying Min Critical Power Ratio Safety Limits ML17292B0281997-08-14014 August 1997 Proposed Tech Specs Revising TS 5.5.6 by Adding Note That Would Extend Interval Requirement to Perform Full Stroke Exercise Testing of TIP-V-6 Until 1998 Refueling Outage ML17292A9691997-08-12012 August 1997 Proposed Tech Specs Supporting Request for Enforcement Discretion for Period of 45 Days from TS Action 3.6.1.3.A Required Actions to Isolate Purge Line & Verify Penetration Flow Path Isolated Every 31 Days ML17292A9421997-07-16016 July 1997 Proposed Tech Specs Adding New Min Reactor Vessel Pressure Versus Reactor Vessel Metal Temp (P/T) Curves,Applicable Up to 12 EFPYs ML17292A8901997-06-0606 June 1997 Revised Tech Spec Page 2.0-1 Modified to Indicate That SLMCPR for ATRIUM-9X Fuel Applies Only to Cycle 13 & Corresponding Bases Pages ML17292A8651997-05-20020 May 1997 Proposed Tech Specs,Requesting Mod of Minimum Critical Power Ratio Safety Limits by 970615 ML17292A7631997-03-24024 March 1997 Rev 7 to Licensee Controlled Specs. ML17292A7621997-03-24024 March 1997 Rev 5 to TS Bases. ML17292A7581997-03-22022 March 1997 Proposed Tech Specs Modifying Response Time Testing SR for RPS Instrumentation,Primary Containment Isolation Actuation Instrumentation & ECCS Actuation Instrumentation ML17292A7531997-03-20020 March 1997 Proposed Tech Specs Re Response Time Testing Requirements ML17292A6591997-01-14014 January 1997 Proposed Tech Specs Reflecting Compilation of TS Change Requests Submitted to NRC in Ltrs Dtd 951208,960709 & 1212 ML17292A6341996-12-12012 December 1996 Proposed Tech Specs Requesting Conversion Based Upon NUREG-1434,rev 1 ML17292A5511996-10-15015 October 1996 Proposed Tech Specs Re Secondary Containment & SGTS to Reflect Revised Secondary Containment Drawdown & post- Accident Analyses Results ML17292A5411996-10-10010 October 1996 Proposed Tech Specs Requesting Addition of Section 2B(6) Re Storage of Byproduct,Source & Special Nuclear Materials ML17292A4501996-09-0606 September 1996 Proposed Tech Specs,Containing Corrections to Factual Statements & Proposed Info to Clarify Evaluations ML17292A4111996-08-0909 August 1996 Proposed Tech Specs,Revising TS Section 6.3 Re Unit Staff Qualifications,By Changing Operations Manager Qualification Requirements Associated W/Operations Knowledge from Meeting Ansi/Ans N18.1-1971 ML17292A3561996-07-0909 July 1996 Proposed Tech Specs,Revising Rev a, Including Changes in Vol 7.Proposed Rev Does Not Change Conclusion of NSHC or Environ Assessment Provided Rev a ML20107M3391996-04-24024 April 1996 Proposed Tech Specs,Modifying TS to Support Cycle 12, Scheduled to Begin Subsequent to Spring 1996 Outage ML17292A1511996-04-22022 April 1996 Proposed Tech Specs,Supplementing TS That Describes Administrative & Editorial Changes to Section 6.0, Administrative Controls. ML17291B2801996-03-19019 March 1996 Proposed Tech Specs Re Containment Leakage Testing ML17291B2491996-02-26026 February 1996 Proposed Tech Specs,Submitting Revised Copy of TS Bases Which Include Minor Changes & Clarifications Made Per Requirements of 10CFR50.59 ML17333A0201996-01-19019 January 1996 Proposed Tech Specs Re Primary Containment Leakage Testing ML17291B0941995-10-26026 October 1995 Proposed Tech Specs,Replacing Existing Reactor Recirculation Flow Control Sys W/Adjustable Speed Drive Sys ML17291A9911995-08-16016 August 1995 Proposed Tech Specs Page 3/4 4-4,incorporating Surveillance Notes in Front of Surveillances 4.4.1.2.1 & 4.4.1.2.2 for Jet Pump Operability to Clarify That Notes Apply to Each Surveillance ML17291A8441995-06-0606 June 1995 Proposed Tech Specs Section 6.0, Administrative Controls. ML17291A8401995-06-0606 June 1995 Proposed Tech Specs Index,Deleting Ref to Bases Pages ML17291A8371995-06-0606 June 1995 Proposed Tech Specs Section 6.9.3.2,adding Ref to Three TRs Describing Analytical Methods That May Be Used in Determining Reactor Core Operating Limits for Reload Licensing Applications ML17291A7561995-04-25025 April 1995 Proposed Tech Specs,Adding RWCU Sys High Blowdown Containment Isolation Trip Function & Associated LCO & SRs to Tables 3.3.2-1,3.3.2-2 & 4.3.2.1-1 ML17291A6541995-02-10010 February 1995 Proposed Tech Specs,Modifying Surveillance Acceptance Criteria from 10% to 20% for Individual Jet Pump diffuser- to-lower Plenum Differential Pressure Variations of Individual Jet Pump from Established Patterns ML17291A4811994-10-31031 October 1994 Proposed Tech Spec Relocating Safety/Relief Valve Position Indication Instrumentation Requirements ML17291A4781994-10-31031 October 1994 Proposed Tech Spec 3/4.1.3.1, Reactivity Control Sys. ML17291A4451994-10-12012 October 1994 Corrected Proposed TS Bases 3/4.2.6, Power/Flow Instability. ML17291A4221994-09-26026 September 1994 Proposed Tech Specs,Reflecting Use of Siemens Power Corp Staif Code for Stability Analysis,Per Ieb 88-007,Suppl 1 ML17291A3981994-09-18018 September 1994 Proposed TS Table 3.6.3-1 Re Primary Containment Isolation Valve Requirements ML17291A3191994-08-0808 August 1994 Proposed Tech Specs 4.0.5 Re Guideliness for Inservice Insp & Testing Program ML17291A2171994-07-12012 July 1994 Proposed Tech Specs for Relocation of TS Tables for Instrument Response Time Limits ML17291A2221994-07-0808 July 1994 Proposed TS W/Regard to Control Rod Scram Insertion Testing Under Emergency Circumstances ML17291A1561994-06-23023 June 1994 Proposed Tech Specs Re Supporting Hydrostatic Testing 1999-07-29
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML17284A9121999-10-13013 October 1999 Proposed Tech Specs 3.3.6.1,Table 3.3. 6.1-1, Primary Containment Isolation Instrumentation. ML17284A8761999-08-27027 August 1999 Replacement Page 9 of 9 to Attachment 4 of Procedure 13.10.6 ML17284A8491999-07-29029 July 1999 Proposed Tech Specs,Revising SR 3.5.2.2 Re Condensate Storage Tank Water Level ML17284A8421999-07-29029 July 1999 Proposed Tech Specs Revising SR of TS 3.8.4, DC Sources - Operating & SR 3.8.5.1 of TS 3.8.5, DC Sources - Shutdown. ML17284A8461999-07-29029 July 1999 Proposed Tech Specs,Revising Table 3.3.5.1-1, ECCS Instrumentation Items 1.a,2.a,4.a & E.A. ML17284A8521999-07-29029 July 1999 Proposed Tech Specs 3.4.9, RHR Shutdown Cooling Sys - Hot Shutdown. ML17333A0021999-04-20020 April 1999 Proposed Tech Specs Section 3.4.11,replacing Existing Reactor Pressure Temp Limit Curves by 000630 ML17292B6341999-04-0707 April 1999 Proposed Tech Specs Modifying MCPR Safety Limits to Allow Continued Power Operation at Plant Following Restart from R-14 RFO ML17292B5731999-03-0101 March 1999 ODCM for WNP-2 ML17292B4881998-12-17017 December 1998 Proposed Tech Specs SR 3.8.1.8,allowing Capability to Manually Transfer Between Preferred & Alternate Offsite Power Sources During Modes 1 & 2 by 990125 ML20198A7051998-11-30030 November 1998 Revs 8 Through 13 to TS Bases & Revs 12 Through 15 of Licensee Controlled Specs ML17284A7181998-08-0505 August 1998 Proposed Tech Specs SR 3.8.4.7,allowing Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 1 & 3,125 Vdc Batteries E-B1-1 & HPCS-B1-DG3 & Div 1,250 Vdc Battery E-B2-1 ML17284A7071998-07-17017 July 1998 Proposed Tech Specs Modifying SR 3.8.4.7 to Allow Performance Discharge Test to Be Performed in Lieu of Battery Svc Test for Div 2,125 Vdc,Battery E-B1-2 ML17284A6431998-05-29029 May 1998 Revised Plant Procedure Sys for Site Wide Procedures, Replacing Pages Located in Manual W/Pages in Package ML17292B2831998-03-0909 March 1998 Proposed Tech Specs Establishing Interim SLMCPR for Siemens Power Corp ATRIUM-9X Fuel Applicable to Cycle 14 Only ML17292B2591998-01-31031 January 1998 Offsite Dose Calculation Manual. ML17292B1321997-12-0404 December 1997 Proposed Tech Specs Modifying Min Critical Power Ratio Safety Limits ML17292B0281997-08-14014 August 1997 Proposed Tech Specs Revising TS 5.5.6 by Adding Note That Would Extend Interval Requirement to Perform Full Stroke Exercise Testing of TIP-V-6 Until 1998 Refueling Outage ML17292A9691997-08-12012 August 1997 Proposed Tech Specs Supporting Request for Enforcement Discretion for Period of 45 Days from TS Action 3.6.1.3.A Required Actions to Isolate Purge Line & Verify Penetration Flow Path Isolated Every 31 Days ML17292A9421997-07-16016 July 1997 Proposed Tech Specs Adding New Min Reactor Vessel Pressure Versus Reactor Vessel Metal Temp (P/T) Curves,Applicable Up to 12 EFPYs ML17292A8901997-06-0606 June 1997 Revised Tech Spec Page 2.0-1 Modified to Indicate That SLMCPR for ATRIUM-9X Fuel Applies Only to Cycle 13 & Corresponding Bases Pages ML17292A8651997-05-20020 May 1997 Proposed Tech Specs,Requesting Mod of Minimum Critical Power Ratio Safety Limits by 970615 ML17292A8301997-03-31031 March 1997 Wppss WNP-2 RPV Surveillance Matls Testing & Analysis. ML17292A7621997-03-24024 March 1997 Rev 5 to TS Bases. ML17292A7631997-03-24024 March 1997 Rev 7 to Licensee Controlled Specs. ML17292A7581997-03-22022 March 1997 Proposed Tech Specs Modifying Response Time Testing SR for RPS Instrumentation,Primary Containment Isolation Actuation Instrumentation & ECCS Actuation Instrumentation ML17292A7531997-03-20020 March 1997 Proposed Tech Specs Re Response Time Testing Requirements ML17292A6591997-01-14014 January 1997 Proposed Tech Specs Reflecting Compilation of TS Change Requests Submitted to NRC in Ltrs Dtd 951208,960709 & 1212 ML17292A6341996-12-12012 December 1996 Proposed Tech Specs Requesting Conversion Based Upon NUREG-1434,rev 1 ML17292A6161996-11-19019 November 1996 Rev 1 to WNP-2 IST Program Plan (Pumps & Valves) 2nd Interval (941213-041212). ML17292A5511996-10-15015 October 1996 Proposed Tech Specs Re Secondary Containment & SGTS to Reflect Revised Secondary Containment Drawdown & post- Accident Analyses Results ML17292A5411996-10-10010 October 1996 Proposed Tech Specs Requesting Addition of Section 2B(6) Re Storage of Byproduct,Source & Special Nuclear Materials ML17292A4501996-09-0606 September 1996 Proposed Tech Specs,Containing Corrections to Factual Statements & Proposed Info to Clarify Evaluations ML17292A4111996-08-0909 August 1996 Proposed Tech Specs,Revising TS Section 6.3 Re Unit Staff Qualifications,By Changing Operations Manager Qualification Requirements Associated W/Operations Knowledge from Meeting Ansi/Ans N18.1-1971 ML17292A3561996-07-0909 July 1996 Proposed Tech Specs,Revising Rev a, Including Changes in Vol 7.Proposed Rev Does Not Change Conclusion of NSHC or Environ Assessment Provided Rev a ML17292A7241996-05-31031 May 1996 Offsite Dose Calculation Manual. ML17292A2741996-04-25025 April 1996 Rev 0 to UT-WNP2-208V0, Exam Summary Sheet. ML20107M3391996-04-24024 April 1996 Proposed Tech Specs,Modifying TS to Support Cycle 12, Scheduled to Begin Subsequent to Spring 1996 Outage ML17292A1511996-04-22022 April 1996 Proposed Tech Specs,Supplementing TS That Describes Administrative & Editorial Changes to Section 6.0, Administrative Controls. ML17291B2801996-03-19019 March 1996 Proposed Tech Specs Re Containment Leakage Testing ML17291B2491996-02-26026 February 1996 Proposed Tech Specs,Submitting Revised Copy of TS Bases Which Include Minor Changes & Clarifications Made Per Requirements of 10CFR50.59 ML17333A0201996-01-19019 January 1996 Proposed Tech Specs Re Primary Containment Leakage Testing ML17291B1751995-12-31031 December 1995 Reactor Power Uprate Startup Test Rept, for WNP-2. W/951215 Ltr ML17291B0941995-10-26026 October 1995 Proposed Tech Specs,Replacing Existing Reactor Recirculation Flow Control Sys W/Adjustable Speed Drive Sys ML17291A9911995-08-16016 August 1995 Proposed Tech Specs Page 3/4 4-4,incorporating Surveillance Notes in Front of Surveillances 4.4.1.2.1 & 4.4.1.2.2 for Jet Pump Operability to Clarify That Notes Apply to Each Surveillance ML17291A9591995-07-28028 July 1995 Operations Instructions OI-23,Rev a to, Human Performance Improvement Program. ML20087E2831995-07-26026 July 1995 Performance Enhancement Strategy 1995 ML17291A8401995-06-0606 June 1995 Proposed Tech Specs Index,Deleting Ref to Bases Pages ML17291A8371995-06-0606 June 1995 Proposed Tech Specs Section 6.9.3.2,adding Ref to Three TRs Describing Analytical Methods That May Be Used in Determining Reactor Core Operating Limits for Reload Licensing Applications ML17291A8441995-06-0606 June 1995 Proposed Tech Specs Section 6.0, Administrative Controls. 1999-08-27
[Table view] |
Text
T 4
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1 ATTACHMENT 2 ,
j -
Proposed Amendments to Sections 5.3.1 and 6.9.3.2 of the WNP-2 Technical
' Specifications i
l 1
9604300335 960424 PDR ADOCK 05000397 p PDR
1
- . I DESIGN FEATURES 5.3 REACTOR CORE FUEL 253EM8t!ES
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5.3.1 The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a I rnatrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched
- uranium dioxide (UO2) as fuel material and water rods or channels. Limited substitutions
'of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved
)
applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number oflead fuel assemblies that have not completed representative testing may be placed in nonlimiting core positions.
& ^
i CONTROL ROD ASTEM8LIET 5.3.2 The reaccor core shall contain 185 cruciform shaped control red assemclies. The control materials shall be baron carnice, S.C. and hafnium.
5.a REACTOR COOLANT SYSTEM OESIGN 3RE!5URE AND rEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowancs for normal degradation pursuant to the applicaole surveillance requirements,
- b. For a pressure of:
- 1. 1250 psig on the suction side of the recirculation puma.
- 2. 1650 psig from the recirculation puso disenarge to the outlet ,
side of the discharge snutoff valve.
- 1. 1550 psig from the discharge shutoff valve to the jet pumps.
- c. For a tamperature of 5757.
YOUJME 5.4.2 The total water and steam volume of the reactor vessal and recirculation system is aooreximately 22,539 cubic feet at a nominal steam dome saturation temoerature of 545*F. .
WASHINGTON NUC:.f.AR UNIT Z 55 Amencment No. M
l
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ADMINISTRATIVE CONTROLS ' ' ' ' ' ' " " - ' * " " * " ' '
CORE OPERATING LINITS REPORT (Continued) 6.9.3.2 The analytical methods used to determine the core operating limits shall be those tcpical reports and those revisions anQor supplements of the topical report previously reviewed and approved by the NRC, which describe the methodology applicable to the current cycle. For WNP-2 the topical reports are:
- 1. ANF-ll2ii(P)(A), and Supplements 1 and 2, 'ANF8 Critical Power Correlation,* April 1990 ,
- 2. Letter, R. C. Jones (NRC) to R. A. Copeland (ANF), 'NRC Approval of ANF8 Additive Constants for ANF 9x9-9X BWR Fuel,' dated November 14, 1990
- 3. ANF-NF-524(P)(A), Revision 2 and Supplements 1 and 2, " Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors,"
November 1990
- 4. "O" ^; i[;iA)," ! , I. "..';ia l_^y, @ ~,j j,i i _ 7_j d ,'Ed S
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" '"' M "(") (".) , "- ? = ' , ": . ' ; f = ! , "L_.. "x ? nr ":th d; ?""":;;
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04,f. XN-NF-85-67(P)(A), Revision 1, ' Generic Nechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,' September 1986
, )d*. ANF-89 014(P)(A), Revision I and Supplements I and 2, " Advanced Nuclear Fuels Corporation Generic Nechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel,' October 1991 f, . IN-NF-81-22(P)(A), ' Generic Statistical Uncertainty Analysis Nethodology,' November 1983 0-[, Jf'. NEDE-240ll-P-A-10-US, ' General Electric Standard Application for Reactor Fuel,' U.S. Supplement, March 1991 Gy /f. NEDE-23785-1-PA, Evaluation of the loss-of-Coolant Revision 1. 'The Accident, Volume GESTR-LOCA III, SAFER /GESTRand SAFER Nodels for the Application Nethodology,' October 1984
\
- f. NEDO-20566A, ' General Electric Company Analytical Ndel for loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K,' September 1986 g . EMF-CC-074(P)(A), ' Volume 1 -- STAIF - A Computer Program for BWR Stability in the Frequency Domain Volume 2 -- STAIF A Computer Program for BWR Stability in the Frequency Domain, Code Qualification Report,*
July 1994.
- 11. CENPD-300-A, " Reference Safety Report For Boiling Water Reactor Reload Fuel", dated ...........
- 12. WPPSS-FTS-131(A), Revision 1, " Applications Topical Report For BWR Design And Analysis," dated March 1996 A A 6-21 Amendment ,o. . ,...,,,
! WASHINGTON NUCLEAR - UNIT 2
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ATTACHMENT 3 No Significant Hazards Consideration Determination
]
J l In accordance with the criteria for a significant hazards consideration established in 10 CFR 50.92, the Supply System has evaluated the proposed amendment to the WNP-2 Technical l
, Specifications and determined that it does not represent a significant hazards consideration.
The following discussion is provided in support of this conclusion.
. 1. Does the proposed amendment involve a significant increase in the probability or
] consequences of an accident previously evaluated?
l The proposed amendment consists of revision of sections 5.3.1 and 6.9.3.2 of the
. WNP-2 Technical Specifications. The proposed revision of Section 5.3.1 makes the description of the fuel assemblies consistent with the description previously submitted as part of the WNP-2 Improved Technical Specifications, and provides a clearer description of the fuel assemblies. Lead Fuel Assemblies (LFA) potentially included in
, non-limiting core positions as indicated in Section 5.3.1 would be included in the scope of the anaylsis performed using NRC approved methodologies to prepare the Core
- Operating Limit Reports (COLR). Thus, limitations pertinent to LFAs would be
- included in the scope of the COLR if such LFAs are included in the core. The proposed revision to Technical Specification Section 6.9.3.2 is submitted in accordance I with the guidance of Generic Letter 88-16 to identify the methodologies used to i develop the COLR that define cycle-specific parameter limits for each fuel cycle. The i revision deletes analytical methodologies no longer pertinent to determining WNP-2 i core operating limits, and adds reference to methodologies authored by ABB CENO i Fuel Operations (ABB) and the Supply System. The ABB methodology will be used to
! prepare the Cycle 12 and 13 COLR. The NRC approved Supply System methodology
- is being added to support potential future use. These methodologies are in accordance
- with accepted principles and have been determined to have the capability to correctly
! define operating limits for a core consisting of SPC and ABB fuel assemblies. The i
WNP-2 Technical Specification requires that core power distribution be maintained within limits defined by the COLR to assure that all applicable limits of the plant safety analysis are met. Therefore, the proposed amendment to the WNP-2 Technical Specification does not involve a significant increase in the probability or consequences of an accident previously evaluated because the reactor will be operated within limits of the plant safety analysis and the ability to mitigate the consequences of all accidents previously evaluated will be maintained.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed amendment consists of revision of sections 5.3.1 and 6.9.3.2 of the WNP-2 Technical Specification. The proposed revision of Section 5.3.1 makes the description of the fuel assemblies consistent with the description previously submitted as part of the WNP-2 Improved Technical Specifications and provides a clearer description of the fuel assemblies. Lead Fuel Assemblies potentially included in non-limiting core positions as indicated in Section 5.3.1 would be included in the scope of e
Page 1 of 2
, . .. l ATTACHMENT 3 the analysis perfomed using NRC approved methodologies to prepare the COLR.
Utilization of 10x10 fuel assemblies manufactured by ABB in the core together with i 9x9 fuel assemblies manufactured by SPC is similar to earlier core reloads involving concurrent use of SPC and 8x8 fuel assemblies manufactured by General Electric (GE). I i The 10x10 fuel assemblies are a standard design for boiling water reactors and have F been operating as LFAs in various locations in the WNP-2 reactor core for six operating cycles without adverse effects. Reload core designs are analyzed to j determine the operating limits to be included in the COLRs, to assure that the reactor l will be operated within the bounds of the plant safety analysis. The ABB methodology
- has been validated by comparison with other analytical methodologies, such that, i
consistent with acceptance criteria in Section 4.4 of the Standard Review Plan, there is more than 95% confidence that calculations performed using the ABB methodology i
, bound 95% of the calculations performed using methodologies previously approved by
- the NRC. The NRC approved Supply System methodology is being added to support potential future use. These methodologies are in accordance with accepted principles
! and have been previously approved by the staff. Therefore, the proposed amendment ;
does not create the possibility of a new or different kind of accident from any accident i i previously evaluated because the reactor has been previously operated safely with a mixture of different fuel assembly designs, the 10x10 fuel assemblies have been safely !
operated in other boiling water reactors and in the WNP-2 reactor as LFAs, and the !
! COLR developed with either the ABB or Supply System methodologies will provide operating safety limits such that the reactor will be operated within the bounds of the 4 plant safety analysis.
~
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
4 Plant safety limits are established through LCOs, limiting safety system settings, and i safety limits specified in the Technical Specification. There will be no changes to these settings and limits as a result of revising Section 5.3.1 or Section 6.9.3.2 except as reflected in the COLR. The reactivity and power distribution of the core will continue to be maintained within analyzed limits as defined by the COLR, and the COLR I
defined limits will be determined by an analytical methodology that assures that the j reactor will be operated within the bounds of the plant safety analysis. Therefore the proposed amendment to the WNP-2 Technical Specifications does not significantly reduce any margin of safety.
Page 2 of 2