ML17292B283

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Proposed Tech Specs Establishing Interim SLMCPR for Siemens Power Corp ATRIUM-9X Fuel Applicable to Cycle 14 Only
ML17292B283
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 03/09/1998
From:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17292B282 List:
References
NUDOCS 9803170389
Download: ML17292B283 (13)
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Text

REQUEST FOR AME NT MRRMCMCIIITICAL WRRRATICRARRTTLIMITR Attachment 2, Page 1 of 1 Marked Up Technical Specification Pages 980317038 9 q80309 05000397 PDR ADO PDR P

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs

2. 1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be ~ 25% RTP.

2. 1. 1.2 With the reactor steam dome pressure ) 785 psig and core flow a 10% rated core flow:

The HCPR for ATRIUH-9X fuel shall be a 1. 13 for two

~~ASLSVFA-% Pet recirculation loop operation or ~ 1. 14 for single r 'rculation loop operation. Whe or ~ ~

R shall be a 1.07 for two recirculation loop operation

.for single recirculation loop operation. The R limits for the ATRIUM-9X fuel are applicable to .

Cycle on+y.

2. 1. 1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2. 1.2 Reactor Coolant S stem Pressure SL Reactor steam dome pressure shall be ( 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hours:

2.2. 1 Restore compliance with all SLs; and 2.2 ' Insert all insertable control rods.

WNP-2 2.0-1 Amendment No. 449 151

eporting Requirements 5.6 5.6 Rep'orting Requirements 5.6.5 CORE OPERATING LIMITS REPORT COLR (continued)

1. ANF-1125(P)(A), and Supplements 1 and 2, "ANFB Critical Power Correlation," April 1990; 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; 9 ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel," October 1991; XN-NF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Hethodology," November 1983; F-u NEDE-23785-1-PA, Revision 1, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume III, SAFER/GESTR Application Methodology," October 1984; NED0-20566A, "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K," September 1986; mp~

mph.r-Program f~WR II~M', en'-Bema+a ,Gede-g~i¹i-cat-ien.

CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996; and continued WNP-2 5.0-21 Amendment No. 149

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT'OLR (continued)

WPPSS-FTS-131(A), Revision 1, "Applications Topical Report for BWR Design and Analysis," March 1996.

C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the.

NRC.

5.6.6 Post Accident Monitorin PAM Instrumentation Re ort When a report is required by Condition B or F of LCO 3.3.3. 1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

WNP-2 5.0-22 Amendment No. 149

REQUEST FOR AME NT MIIMIMCRITICAL WEE MTIE SARETT LIMITS , Page 1 of 1 Revised Technical Specification Pages

P I

r'

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs

2. 1. 1 Reactor Core SLs
2. l. 1. 1 With the reactor steam dome pressure < 785 psig or core flow <"10% rated core flow:

THERMAL POWER shall be z 25% RTP.

2. 1. 1.2 With the reactor steam dome pressure ~ 785 psig and core flow ~ 10% rated core flow:

The HCPR for ATRIUM-9X fuel shall be a 1. 13 for two recirculation loop operation or a 1. 14 for single recirculation loop operation. The HCPR for the ABB SVEA-96 fuel shall be ) 1.07 for two recirculation loop operation or a 1.09 for single recirculation loop operation. The HCPR limits for the ATRIUM-9X fuel are applicable to Cycle 14.

2. 1. 1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2. 1.2 Reactor Coolant S stem Pressure SL Reactor steam dome pressure shall be ~ 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hours:

2.2. 1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

WNP-2 2.0-1 Amendment No. 449 kR

f OReporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT COLR (continued)

1. ANF-1125(P)(A), and Supplements 1 and 2, "ANFB Critical Power Correlation," April 1990;
2. 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;

3. ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel," October 1991;
4. XN-NF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Methodology," November 1983;
5. NEDE-23785-1-PA, Revision 1, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume III, SAFER/GESTR Application Methodology," October 1984;
6. NED0-20566A, "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K," September 1986;
7. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996; and
8. WPPSS-FTS-131(A), Revision 1, "Applications Topical Report for BWR Design and Analysis," March 1996.

C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the

-NRC.

(continued)

WNP-2 5.0-21 Amendment No. 449

C t Reporting Requirements 5.6 5.8 Reporting Requirements (continued) 5.6.6 Post Accident Monitorin PAM Instrumentation Re ort When a report is required by Condition B or F of LCO 3.3.3. 1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation-channels of the Function -to OPERABLE status.

WNP-2 5.0-22 Amendment No. 449

4e l

C"

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