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Category:OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING
MONTHYEARML18107A4991999-08-25025 August 1999 Application for Amends to Licenses DPR-70 & DPR-75,proposing to Authorize Licensee to Perform Single Cell Charging of Connected Cells in Operable Class 1E Batteries ML18107A5141999-08-25025 August 1999 Suppl to 971114 Application for Amends to Licenses DPR-70 & DPR-75,deleting Footnotes Added as one-time Changes & Proposing Change Identified Subsequent to 971114 Submittal That Corrects Error Issued in Amend 69 for Unit 2 ML18107A4631999-07-29029 July 1999 Application for Amends to Licenses DPR-70 & DPR-75,replacing Surveillance 4.6.1.1a with Appropriate NUREG-1431 Requirements ML18107A4581999-07-23023 July 1999 Application for Amends to Licenses DPR-70 & DPR-75 to Change TS Surveillance Requirement 4.8.1.1.2 D 7 by Removing Restriction to Perform Test Every 18 Months During Shutdown ML18107A4221999-07-0202 July 1999 Application for Amends to Licenses DPR-70 & DPR-75 to Relocate TS 3/4.3.4 Lco,Surveillance Requirements & Bases from TS to Ufsar,Iaw GL 95-10 ML18107A3771999-06-10010 June 1999 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS 3/4.6.1, Containment Integrity, by Clarifying When Verification of Primary Containment Integrity May Be Performed by Administrative Means ML18107A3571999-06-0404 June 1999 Application for Amends to Licenses DPR-70,DPR-75 & NPF-57 Re Transfer of PSEG Ownership Interests & Licensed Operating Authorities to New,Affiliated Nuclear Generating Company, PSEG Nuclear Llc.Proprietary Encls Withheld ML18107A2081999-04-14014 April 1999 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.9.12, Fhvas, to Ensure Consistency Between TS Requirements for Fhavs & Sys Design Basis,Per NUREG-1431. Page 4 of 4 in Attachment 2 of Incoming Submittal Omitted ML18106B1571999-03-31031 March 1999 Suppl to 990115 Application for Amend to License DPR-70, Allowing Plant Operation to Continue to Thirteenth Refueling Outage (1R13),currently Scheduled to Begin on 990918 ML18106B0621999-02-0808 February 1999 Application for Amends to Licenses DPR-70 & DPR-75, Respectively.Proposed Amend Modifies ECCS Subsystems Surveillance Requirement 4.5.3.2.b by Adding Option of Providing Necessary Isolation Function When Testing ML18106B0501999-02-0202 February 1999 Application for Amends to Licenses DPR-70 & DPR-75,revising Max Permissible Enrichment of Stored New Fuel Described in TS Section 5.6.1.1.Rept Entitled, Criticality Analysis of Salem Units 1 & 2 Fresh Fuel Racks, Encl ML18106B0221999-01-15015 January 1999 Application for Amend to License DPR-70,proposing one-time Changes to Certain Unit 1 TS Surveillance Requirements for Fuel Cycle 13,currently Scheduled to Begin on 990918 ML18106A9161998-10-12012 October 1998 Application for Amend to License DPR-75,requesting one-time Changes to Certain Unit 2 TS SR for Fuel Cycle 10.Basis for Requested Changes,Analysis Supporting No Significant Hazards Determination & Marked Up TS Pages,Encl ML18106A8921998-09-29029 September 1998 Application for Amends to Licenses DPR-72 & DPR-75,modifying TS 3/4.9.4 to Permit Use of Equivalent Methods to Obtain Containment Closure During Refueling Operations for Containment Equipment Hatch ML18106A8761998-09-17017 September 1998 Application for Amends to Licenses DPR-70 & DPR-75,modifying TSs 3/4.8.2.2,3/4.8.2.4 & 3/4.8.2.6 to Address Movement of Irradiated Fuel & Deleting LCO Re Requirement to Establish Containment Integrity ML18106A8351998-08-12012 August 1998 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS for Containment Air Locks for Salem Units 1 & 2 ML18106A7911998-07-30030 July 1998 Application for Amends to Licenses DPR-70 & DPR-75,revising Acceptance Criteria for CR Emergency Air Conditioning Sys from Maintaining CR at 1/8-inch Positive Pressure W/Respect to Adjacent Areas to Following Phrase ML18106A4271998-03-26026 March 1998 Application for Amend to License DPR-70,changing Applicability Statement from Mode 3 to Modes 1 & 2 ML18106A4221998-03-26026 March 1998 Application for Amends to Licenses DPR-70 & DPR-75, Respectively,Incorporating Note at Bottom of Salem Unit 2 T3/4.8.2 Into Salem 1 TS & Adding Asterisks to Word Inverters. ML18106A2521998-01-0808 January 1998 Supplemental Application for Amends to Licenses DPR-70 & DPR-75,respectively Adding Asterisk to Word Operable in LCO & Associated Note at Bottom of Page ML18106A2111997-12-15015 December 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS to Adopt Option B of 10CFR50,App J,For Type B & C Testing & Modifying Existing TS Wording for Previous Adoption of Option B on Type a Testing ML18106A1941997-12-11011 December 1997 Application for Amend to License DPR-70,requesting one-time Exemption to Allow Containment Purging in Modes 3 & 4 ML18102B6771997-11-14014 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,providing TS Surveillance Requirements to Codify Existing Procedural Commitments for SW Accumulator Vessels ML18102B6731997-11-14014 November 1997 Application for Amends to Licenses DPR-70 & DPR-75, Correcting Editorial & Administrative Errors in Ts,Which Existed Since Initial Issuance ML18102B6611997-11-0404 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.8.1 Re EDG & Deleting SR 4.8.1.1.2.d.1,calling for Diesels to Be Subjected to Insp Based on Vendor Recommendations ML18102B6571997-11-0404 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising Surveillance Requirement 4.6.2.1.b to Verify That on Recirculation Flow,Containment Spray Pumps Develop Differential Pressure of Greater than or Equal to 204 Psid ML18102B6371997-10-29029 October 1997 Application for Amend to License DPR-75,revising TS 3/4.4.6, Steam Generators, Making One Time Change to Unit 2 SG Insp Schedule to Require Next Insp within 24 Months of Mode 2 for Unit 2 Fuel Cycle 10 ML18102B6411997-10-29029 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising SR 4.8.1.1.2.d.2 Proposed in Util 960925 Submittal Re EDG Testing ML18102B6471997-10-24024 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS 3/4.7.7, Auxiliary Building Exhaust Air Filtration Sys & Bases ML18102B6441997-10-24024 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising Containment Hydrogen Analyzer SRs of TS 4.6.4.1 ML18102B6291997-10-21021 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,extending Applicability of Power Range Neutron Flux TS Trip to Require Operability & Surveillance Testing in Mode 3,when Reactor Trip Sys Breakers in Closed Position ML18102B6181997-10-14014 October 1997 Application for Amend to License DPR-70,modifying Pressure Isolation Valve Lco,Lco Action Statement,Surveillance Requirement & Table 4.4-4 to Be Consistent W/Salem Unit 2 ML18102B6061997-10-0606 October 1997 Application for Amend to License DPR-70,changing TS for LCOs 3.1.3.1, Movable Control Assemblies & 3.1.3.2.1, Position Indication Sys. ML18102B5551997-08-27027 August 1997 Application for Amends to Licenses DPR-70 & DPR-75,providing Revised TS Bases Pages for TS 3.7.3, CCW Sys, to State That CCW Pumps Required to Be Operable to Meet LCO ML18102B5281997-08-19019 August 1997 Application for Amend to License DPR-75,providing Addl Operational Flexibility to Allow Orderly Resumption of Startup & Preclude Unwarranted Power Transients at Plant Unit 2 ML18102B5321997-08-18018 August 1997 Application for Amend to License DPR-75,incorporating Plant Computer (P-250) Into TS Bases for Movable Control Assemblies ML18102B4871997-08-0101 August 1997 Application for Amends to Licenses DPR-70 & DPR-75,rewording TS Section 4.2.1 to State That Util Will Adhere to Section 7 Incidental Take Statement Approved by Natl Marine Fisheries Svc ML18102B3041997-05-14014 May 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising SR 4.7.6.1.d.1 to Indicate That Specified Acceptable Filter Dp Is to Be Measured Across Filter Housing & to Reflecting Filter Dp Acceptance Value of Greater than 2.70 Wg ML18102B2611997-05-0101 May 1997 Suppl Info to Util 970211 Application for Amends to Licenses DPR-70 & DPR-75,revising TS for Chilled Water Sys - Auxiliary Bldg Subsystem.Ref to Radiation Monitoring Signal Removed from Bases Section ML18102B2581997-05-0101 May 1997 Application for Amend to License DPR-75,revising TS Section 3/4.7.7 to Reflect Extended Allowed Outage Times & Creating New TS Section 3/4.7.11 to Address Importance of Switchgear & Penetration Area Ventilation Sys ML18102B0071997-04-30030 April 1997 Application for Amends to Licenses DPR-40,DPR-56,DPR-70 & DPR-75,requesting Consent of NRC for Indirect Transfer of Control of Interest in Plants That Will Occur Under Proposed Merger ML18102B2511997-04-28028 April 1997 Submits Suppl to 970131 & 0314 Applications for Amends to Licenses DPR-70 & DPR-75,revising TS on Pressurizer Power Operated Relief Valves.Suppl Issued to State That Conclusion of NSHC Remains Valid ML18102B0221997-04-25025 April 1997 Application for Amends to Licenses DPR-70 & DPR-75, Eliminating Flow Path from RHR Sys to RCS Hot Legs as Specified in LCO 3.5.2.c.2 ML18102A9791997-04-11011 April 1997 Application for Amends to Licenses DPR-70 & DPR-75, Increasing Cw Flow to 2550 Gpm for 31 Day Surveillance Fans Started & Operated in Low Speed ML18102A8951997-03-0404 March 1997 Application for Amends to Licenses DPR-70 & DPR-75,modifying ECCS Surveillance Test Acceptance Criteria for Centrifugal Charging & SI Pumps ML18102A8491997-02-11011 February 1997 Application for Amends to Licenses DPR-70 & DPR-75, Requesting Addition of New TS 3/4.7.10, Chilled Water Sys ML18102A8361997-01-31031 January 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.4.3 to Ensure That Automatic Capability of Power Operated Relief Valve to Relieve Pressure Is Maintained When Valves Are Isolated by Closure of Block Valves ML18102A8021997-01-31031 January 1997 Application for Amends to Licenses DPR-70 & DPR-75, Requesting Rev to TS Re Containment Systems Air Temp to Ensure Representative Average Air Temp Is Measured ML18102A7861997-01-23023 January 1997 Application for Amend to License DPR-75,revising TS Bases 3/4.3.3.1 Re Radiation Monitoring Instrumentation Channel 2R41 ML18102A7361997-01-0707 January 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising Eighteen Month Surveillance Performed to Measure RCS Total Flow Rate to Account for Necessary Plant Conditions for Performing Test in Mode 1 & Current Extended Plant Outages 1999-08-25
[Table view] Category:TEXT-LICENSE APPLICATIONS & PERMITS
MONTHYEARML18107A4991999-08-25025 August 1999 Application for Amends to Licenses DPR-70 & DPR-75,proposing to Authorize Licensee to Perform Single Cell Charging of Connected Cells in Operable Class 1E Batteries ML18107A5141999-08-25025 August 1999 Suppl to 971114 Application for Amends to Licenses DPR-70 & DPR-75,deleting Footnotes Added as one-time Changes & Proposing Change Identified Subsequent to 971114 Submittal That Corrects Error Issued in Amend 69 for Unit 2 ML18107A4631999-07-29029 July 1999 Application for Amends to Licenses DPR-70 & DPR-75,replacing Surveillance 4.6.1.1a with Appropriate NUREG-1431 Requirements ML18107A4581999-07-23023 July 1999 Application for Amends to Licenses DPR-70 & DPR-75 to Change TS Surveillance Requirement 4.8.1.1.2 D 7 by Removing Restriction to Perform Test Every 18 Months During Shutdown ML18107A4221999-07-0202 July 1999 Application for Amends to Licenses DPR-70 & DPR-75 to Relocate TS 3/4.3.4 Lco,Surveillance Requirements & Bases from TS to Ufsar,Iaw GL 95-10 ML18107A3771999-06-10010 June 1999 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS 3/4.6.1, Containment Integrity, by Clarifying When Verification of Primary Containment Integrity May Be Performed by Administrative Means ML18107A3571999-06-0404 June 1999 Application for Amends to Licenses DPR-70,DPR-75 & NPF-57 Re Transfer of PSEG Ownership Interests & Licensed Operating Authorities to New,Affiliated Nuclear Generating Company, PSEG Nuclear Llc.Proprietary Encls Withheld ML18107A2081999-04-14014 April 1999 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.9.12, Fhvas, to Ensure Consistency Between TS Requirements for Fhavs & Sys Design Basis,Per NUREG-1431. Page 4 of 4 in Attachment 2 of Incoming Submittal Omitted ML18106B1571999-03-31031 March 1999 Suppl to 990115 Application for Amend to License DPR-70, Allowing Plant Operation to Continue to Thirteenth Refueling Outage (1R13),currently Scheduled to Begin on 990918 ML18106B0621999-02-0808 February 1999 Application for Amends to Licenses DPR-70 & DPR-75, Respectively.Proposed Amend Modifies ECCS Subsystems Surveillance Requirement 4.5.3.2.b by Adding Option of Providing Necessary Isolation Function When Testing ML18106B0501999-02-0202 February 1999 Application for Amends to Licenses DPR-70 & DPR-75,revising Max Permissible Enrichment of Stored New Fuel Described in TS Section 5.6.1.1.Rept Entitled, Criticality Analysis of Salem Units 1 & 2 Fresh Fuel Racks, Encl ML18106B0221999-01-15015 January 1999 Application for Amend to License DPR-70,proposing one-time Changes to Certain Unit 1 TS Surveillance Requirements for Fuel Cycle 13,currently Scheduled to Begin on 990918 ML18106A9161998-10-12012 October 1998 Application for Amend to License DPR-75,requesting one-time Changes to Certain Unit 2 TS SR for Fuel Cycle 10.Basis for Requested Changes,Analysis Supporting No Significant Hazards Determination & Marked Up TS Pages,Encl ML18106A8921998-09-29029 September 1998 Application for Amends to Licenses DPR-72 & DPR-75,modifying TS 3/4.9.4 to Permit Use of Equivalent Methods to Obtain Containment Closure During Refueling Operations for Containment Equipment Hatch ML18106A8761998-09-17017 September 1998 Application for Amends to Licenses DPR-70 & DPR-75,modifying TSs 3/4.8.2.2,3/4.8.2.4 & 3/4.8.2.6 to Address Movement of Irradiated Fuel & Deleting LCO Re Requirement to Establish Containment Integrity ML18106A8351998-08-12012 August 1998 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS for Containment Air Locks for Salem Units 1 & 2 ML18106A7911998-07-30030 July 1998 Application for Amends to Licenses DPR-70 & DPR-75,revising Acceptance Criteria for CR Emergency Air Conditioning Sys from Maintaining CR at 1/8-inch Positive Pressure W/Respect to Adjacent Areas to Following Phrase ML18106A4271998-03-26026 March 1998 Application for Amend to License DPR-70,changing Applicability Statement from Mode 3 to Modes 1 & 2 ML18106A4221998-03-26026 March 1998 Application for Amends to Licenses DPR-70 & DPR-75, Respectively,Incorporating Note at Bottom of Salem Unit 2 T3/4.8.2 Into Salem 1 TS & Adding Asterisks to Word Inverters. ML18106A2521998-01-0808 January 1998 Supplemental Application for Amends to Licenses DPR-70 & DPR-75,respectively Adding Asterisk to Word Operable in LCO & Associated Note at Bottom of Page ML18106A2111997-12-15015 December 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS to Adopt Option B of 10CFR50,App J,For Type B & C Testing & Modifying Existing TS Wording for Previous Adoption of Option B on Type a Testing ML18106A1941997-12-11011 December 1997 Application for Amend to License DPR-70,requesting one-time Exemption to Allow Containment Purging in Modes 3 & 4 ML18102B6771997-11-14014 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,providing TS Surveillance Requirements to Codify Existing Procedural Commitments for SW Accumulator Vessels ML18102B6731997-11-14014 November 1997 Application for Amends to Licenses DPR-70 & DPR-75, Correcting Editorial & Administrative Errors in Ts,Which Existed Since Initial Issuance ML18102B6611997-11-0404 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.8.1 Re EDG & Deleting SR 4.8.1.1.2.d.1,calling for Diesels to Be Subjected to Insp Based on Vendor Recommendations ML18102B6571997-11-0404 November 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising Surveillance Requirement 4.6.2.1.b to Verify That on Recirculation Flow,Containment Spray Pumps Develop Differential Pressure of Greater than or Equal to 204 Psid ML18102B6371997-10-29029 October 1997 Application for Amend to License DPR-75,revising TS 3/4.4.6, Steam Generators, Making One Time Change to Unit 2 SG Insp Schedule to Require Next Insp within 24 Months of Mode 2 for Unit 2 Fuel Cycle 10 ML18102B6411997-10-29029 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising SR 4.8.1.1.2.d.2 Proposed in Util 960925 Submittal Re EDG Testing ML18102B6471997-10-24024 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,modifying TS 3/4.7.7, Auxiliary Building Exhaust Air Filtration Sys & Bases ML18102B6441997-10-24024 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising Containment Hydrogen Analyzer SRs of TS 4.6.4.1 ML18102B6291997-10-21021 October 1997 Application for Amends to Licenses DPR-70 & DPR-75,extending Applicability of Power Range Neutron Flux TS Trip to Require Operability & Surveillance Testing in Mode 3,when Reactor Trip Sys Breakers in Closed Position ML18102B6181997-10-14014 October 1997 Application for Amend to License DPR-70,modifying Pressure Isolation Valve Lco,Lco Action Statement,Surveillance Requirement & Table 4.4-4 to Be Consistent W/Salem Unit 2 ML18102B6061997-10-0606 October 1997 Application for Amend to License DPR-70,changing TS for LCOs 3.1.3.1, Movable Control Assemblies & 3.1.3.2.1, Position Indication Sys. ML18102B5551997-08-27027 August 1997 Application for Amends to Licenses DPR-70 & DPR-75,providing Revised TS Bases Pages for TS 3.7.3, CCW Sys, to State That CCW Pumps Required to Be Operable to Meet LCO ML18102B5281997-08-19019 August 1997 Application for Amend to License DPR-75,providing Addl Operational Flexibility to Allow Orderly Resumption of Startup & Preclude Unwarranted Power Transients at Plant Unit 2 ML18102B5321997-08-18018 August 1997 Application for Amend to License DPR-75,incorporating Plant Computer (P-250) Into TS Bases for Movable Control Assemblies ML18102B4871997-08-0101 August 1997 Application for Amends to Licenses DPR-70 & DPR-75,rewording TS Section 4.2.1 to State That Util Will Adhere to Section 7 Incidental Take Statement Approved by Natl Marine Fisheries Svc ML20148K0841997-06-0404 June 1997 Amends 194 & 177 to Licenses DPR-70 & DPR-75,respectively, Changing TSs 3.4.3 & 3.4.5, Relief Valves, to Ensure That Automatic Capability of PORVs to Relieve Pressure Is Maintained When Valves Are Isolated ML18102B3041997-05-14014 May 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising SR 4.7.6.1.d.1 to Indicate That Specified Acceptable Filter Dp Is to Be Measured Across Filter Housing & to Reflecting Filter Dp Acceptance Value of Greater than 2.70 Wg ML18102B2611997-05-0101 May 1997 Suppl Info to Util 970211 Application for Amends to Licenses DPR-70 & DPR-75,revising TS for Chilled Water Sys - Auxiliary Bldg Subsystem.Ref to Radiation Monitoring Signal Removed from Bases Section ML18102B2581997-05-0101 May 1997 Application for Amend to License DPR-75,revising TS Section 3/4.7.7 to Reflect Extended Allowed Outage Times & Creating New TS Section 3/4.7.11 to Address Importance of Switchgear & Penetration Area Ventilation Sys ML18102B0071997-04-30030 April 1997 Application for Amends to Licenses DPR-40,DPR-56,DPR-70 & DPR-75,requesting Consent of NRC for Indirect Transfer of Control of Interest in Plants That Will Occur Under Proposed Merger ML18102B2511997-04-28028 April 1997 Submits Suppl to 970131 & 0314 Applications for Amends to Licenses DPR-70 & DPR-75,revising TS on Pressurizer Power Operated Relief Valves.Suppl Issued to State That Conclusion of NSHC Remains Valid ML18102B0221997-04-25025 April 1997 Application for Amends to Licenses DPR-70 & DPR-75, Eliminating Flow Path from RHR Sys to RCS Hot Legs as Specified in LCO 3.5.2.c.2 ML18102A9791997-04-11011 April 1997 Application for Amends to Licenses DPR-70 & DPR-75, Increasing Cw Flow to 2550 Gpm for 31 Day Surveillance Fans Started & Operated in Low Speed ML18102A8951997-03-0404 March 1997 Application for Amends to Licenses DPR-70 & DPR-75,modifying ECCS Surveillance Test Acceptance Criteria for Centrifugal Charging & SI Pumps ML18102A8491997-02-11011 February 1997 Application for Amends to Licenses DPR-70 & DPR-75, Requesting Addition of New TS 3/4.7.10, Chilled Water Sys ML18102A8361997-01-31031 January 1997 Application for Amends to Licenses DPR-70 & DPR-75,revising TS 3/4.4.3 to Ensure That Automatic Capability of Power Operated Relief Valve to Relieve Pressure Is Maintained When Valves Are Isolated by Closure of Block Valves ML18102A8021997-01-31031 January 1997 Application for Amends to Licenses DPR-70 & DPR-75, Requesting Rev to TS Re Containment Systems Air Temp to Ensure Representative Average Air Temp Is Measured ML18102A7861997-01-23023 January 1997 Application for Amend to License DPR-75,revising TS Bases 3/4.3.3.1 Re Radiation Monitoring Instrumentation Channel 2R41 1999-08-25
[Table view] |
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.. , ... Public Service Electric and Gas Company Joseph J. Hagan Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President
-Nuclear Operations fEB 03 1994 NLR-N94016 LCR 94-06 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR AMENDMENT REACTOR COOLANT SYSTEM FLOW RATE SALEM GENERATING STATION UNIT NO. 2 DOCKET NO. 50-311 In accordance with the requirements of 10CFR50.90, Public Service Electric & Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating License DPR-75 for Salem Generating Station Unit No. 2. In accordance with 10CFR50.91(b)
(1) requirements, a copy of this request has been sent to the State of New Jersey. The proposed amendment modifies Technical Specification 2.2, Limiting Safety System Settings, Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints, Functional Unit 12 Loss of Flow. The Reactor Coolant System (RCS) Loop design flow is reduced 1% to 86,430 gpm per loop. The Note (*) associated with the Trip Setpoint and the Allowable Value for Loss of Flow is changed to reflect the new value. The proposed amendment also modifies the Technical Specification 3.2.5, Power Distribution Limits, Table 3.2-1 -DNB Parameters.
The RCS minimum required total flow rate is reduced 1% to 353,700 gpm. Table 3.2-1 is modified for Reactor Coolant System flow to provide a limit of greater than or equal to 353,700 gpm. Attachment 1 includes a description, justification, and significant hazards analysis for the proposed change. Attachment 2 contains the Technical Specification pages revised with pen and ink changes. 9402220069 940203 PDR :ADOCK 05000311 P PDR -----,'. '1 Lo\
Document Control Desk* NLR-N94016 LCR 94-06 2 FEB 0 3 1994 PSE&G is requesting a 60 day implementation period after amendment approval.
Should there be any questions with regard to this submittal, please do not hesitate to contact us. c Mr. J. c. Stone Licensing Project Manager Mr. c. Marschall Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Manager IV gan sident -Operations New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 Ref: NLR-N94016 STATE OF NEW JERSEY SS. COUNTY OF SALEM J. J. Hagan, being duly sworn according to law deposes and says: I am Vice President
-Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Salem Generating Station, Unit No. 2, are true to the best of my knowledge, information and belief. me 1994 KIMBERLY JO BROWN My commission expires on ATTACHMENT 1 NLR-N94016 LCR 94-06 REACTOR COOLANT SYSTEM FLOW RATE I. DESCRIPTION OF THE PROPOSED CHANGE A. Change Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints as follows: 1. For Functional Unit 12, Loss of Flow, change the Note to read: "*Design flow is 86,430 gpm per loop." B. Change Table 3.2-1 DNB Parameters as follows: 1. Change the third parameter from "Reactor Coolant System" to read, "Reactor Coolant System Total Flow Rate". 2. Change the limit for Reactor Coolant System Total Flow Rate to read:
gpm. II. REASON FOR THE PROPOSED CHANGES Salem Unit No.2 has experienced a decrease in calculated RCS total flow over the past several refueling cycles. This has not been confirmed by RCS elbow tap data, which tends to indicate flow has remained basically constant.
PSE&G is investigating the differences that have occurred between the calculated RCS flow and the elbow tap indications.
Following the Unit 2 eighth refueling outage, RCS total flow was calculated to be slightly above the minimum required by Technical Specification 3.2.5. Recently, a review of the flow calculation procedure identified a non-conservatism that may reduce the calculated flow by approximately 1000 gpm. PSE&G believes the decrease in calculated RCS total flow is based on changes in the indications and inputs to the calculation, not actual RCS flow. PSE&G is investigating the low RCS flow calculation to resolve the discrepancy between calculated flow and elbow tap indications.
The impetus for the proposed revision is the small margin between the calculated RCS total flow and the Technical Specification limit. III. JUSTIFICATION FOR THE PROPOSED CHANGES Technical Specification
2.2 Limiting
Safety System Settings -Reactor Trip System Instrumentation Setpoints requires the Trip NLR-N94016 Attachment 1 Setpoint and Allowable Value for Functional Unit 12 Loss of Flow to be based on the design RCS Flow per loop. If the setpoint is less conservative than the required value from Table 2.2-1, the channel must be declared inoperable and the appropriate Action Statement from Chapter 3/4.3 Instrumentation must be entered. Design RCS flow identified for calculation of this setpoint is 87,300 gpm per loop. This flow is being reduced 1% to 86,430 gpm per loop. Since the loss of flow setpoint must be greater than or equal to 90% of the specified the design flow, reducing the design flow value would not require a physical change to the setpoint.
Technical Specification
3.2.5 Power
Distribution Limits -DNB Limits requires Reactor Coolant System (RCS) Total Flow Rate to be greater than or equal to 357,200 gpm. If RCS flow is less than the required limit, flow must be restored to within limits within two hours or Thermal Power must be reduced to less than 5% of Rated Thermal Power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The flow is determined by precision heat balance measurements at least once per 18 months and the elbow tap flow indications are correlated to the calculated value. PSE&G does not re-calibrate flow tap indications, but provides administrative limits that correlate to the calculated RCS flow values. The elbow tap meters provide continuous flow indication to ensure total RCS flow is greater than that required by Technical Specification 4.2.5.1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The Loss of Flow Reactor Trip is set at of design RCS flow per loop (87,300 gpm) and thus prevents operation with significant flow reductions.
The RCS total flow rate is limited to greater than or equal to 357,200 gpm to ensure DNB limits are not exceeded.
The proposed Technical Specification revision reduces the RCS loop flow rate to 86,430 gpm and RCS total flow rate minimum value to 353,700 gpm. This flow rate is a 1% reduction in the minimum required RCS loop and total flow rates. EFFECTS OF REDUCED RCS FLOW ON FSAR ANALYSIS PSE&G has confirmed that adequate margin exists in the LOCA, Non-LOCA and Containment analyses to justify the proposed Technical Specification changes. The accidents which rely on adequate RCS flow have been evaluated for the 1% reduction.
LOCA Analyses The following UFSAR LOCA analyses were evaluated for effects of a 1% reduction in RCS flow: Large Break LOCA (UFSAR Section 15.4.1) Small Break LOCA (UFSAR Section 15.3.1) Steam Generator Tube Rupture (UFSAR Section 15.4.4)
NLR-N94016 Attachment 1 The Large Break LOCA analyses assume an RCS flow of 345,600 gpm. The Small Break LOCA and Steam Generator Tube Rupture analyses assume a flow of 330,000 gpm. These values conservatively bound the proposed reduction in RCS flow. Therefore, the proposed change has no adverse impact on these analyses.
The peak cladding temperature and hydrogen generation criteria of 10CFR50.46, and the offsite dose criteria of lOCFRlOO would continue to be met. Long Tenn LOCA (Containment Integrity)
Analyses (UFSAR Section 15. 4. 8 .1) Because the proposed reduction in RCS flow does not affect the Tavg limits, the overall energy available in the RCS would not increase, and the limiting LOCA cases for mass and energy releases would not be affected.
However, because there is a slight increase in reactor vessel delta-T, the hot leg break cases were evaluated to assess the effect of the redistribution of available energy towards the hot leg side. This evaluation concludes that the containment pressure resulting from a double ended hot leg break would increase by an upper bounding limit of 0.15 psi. The hot leg breaks would remain bounded by the limiting double-ended pump suction break cases, which are not adversely affected by the proposed reduction in RCS flow. Subcompartment Analyses (UFSAR Section 15.4.8.3)
The proposed reduction in RCS flow is estimated to result in a 0.4 degree F reduction in vessel inlet temperature.
Reduced temperature results in increased fluid density. The penalty associated with the change in temperature would result in approximately a 0.1% increase in critical flow. The licensing basis analysis model is not readily available to allow performance of a sensitivity evaluation for the 0.1% increase in critical flow. However, review of the current subcompartment analysis results show that evaluations were performed for breaks as large as 100 square inches at the reactor pressure vessel inlet and outlet piping. Based on piping displacements resulting from LOCA, and gap sizes for pipe whip restraints, it has been determined that the largest break consistent with the RCS piping configuration, is 75 square inches at the vessel inlet and outlet locations, and a single-ended break at all other RCS locations.
This reduction in break size offsets any penalty associated with the reduced RCS flow. Note that the reduction in break size is based on a mechanistic evaluation of RCS piping, but does not rely upon leak-before break technology.
A Salem specific leak-before-break submittal was made to the NRC on July 6, 1993, justifying further relaxations in primary loop pipe break postulations for the purposes of evaluating dynamic effects.
NLR-N94016 Attachment 1 Blowdown Reactor Vessel and Loop Forces (UFSAR Section 3.9.1.5} The forces created by a postulated RCS pipe break are a function of RCS operating conditions, including reactor vessel inlet and outlet temperatures.
The proposed 1% reduction in RCS flow would have an effect on RCS temperature.
However, the impact on operating temperature is small enough to be considered negligible relative to the calculation of forces from a postulated RCS pipe break. Post LOCA. Long Term Subcriticality (Related to UFSAR Section 15.4.1) Hot Leg Switchover Time and Adequate Core Cooling Flow Verification (UFSAR Sections 6.3.2 and 15.4.1) A review of the analyses associated with these transients indicates that the reduced RCS flow would have no adverse impact on the assumptions for the respective calculations.
Non-LOCA Analyses The following current Salem Non-LOCA analyses were evaluated for the effects of a 1% reduction in RCS flow: Excessive Load Increase (UFSAR Section 15.2.11) Four cases of a 10% step load increase from nominal full power conditions are analyzed, based on automatic versus manual rod control, and minimum versus maximum reactivity feedback parameters.
Evaluation of this transient using the proposed reduction in RCS flow has concluded that the Departure From Nucleate Boiling Ratio (DNBR) design basis is met. Excessive Heat Removal Due to Feedwater System Malfunctions (UFSAR Section 15.2.10) Two cases for this transient are analyzed; a full power case is evaluated to show that the DNBR remains above the safety limit, and a zero power case is shown to be bounded by the Rod Withdrawal From Subcritical (RWFS) transient.
Steam generator overfill is considered also evaluated for this transient.
Evaluation of the proposed reduction in RCS flow has determined that the DNBR for the full power case would remain above the safety limit, the zero power case remains bounded by RWFS, and steam generator overfill would not be affected.
Accidental Depressurization of the Main Steam System (UFSAR Section 15.2.13) This event is initiated by the full opening of a single steam dump, relief or safety valve from hot zero power conditions.
A reduction in RCS flow potentially decreases the minimum DNBR.
NLR-N94016 Attachment 1 Evaluation of this event shows that the DNBR analysis of record remains valid given the proposed reduction in RCS flow. Major Secondary Side Pipe Rupture (UFSAR Section 15.4.2) A reduction in RCS flow would potentially decrease the minimum DNBR for this event, which considers a double-ended rupture of the main steam piping at hot zero power, with and without offsite power. The DNB penalty would be partially offset because the lower flow would reduce the primary to secondary heat transfer, resulting in a reduced power increase.
Evaluation of this event concludes that sufficient DNB margin exists to offset the reduced flow penalty, without taking credit for the offsetting reduction in power increase.
Steam Line Break Mass/Energy Release (UFSAR 15.4.8.2)
A decrease in RCS flow reduces primary to secondary heat transfer, thereby reducing steam pressure and temperature during normal operation.
Any reduction in secondary temperature and pressure would tend to lessen the mass/energy release following a steam line break. Therefore, the proposed reduction in RCS flow would not adversely affect the steamline break mass/energy releases.
Loss of External Electrical Load and/or Turbine Trip (UFSAR Section 15.2.13) Four cases of a total loss of steam demand at full power, without a direct reactor trip, are analyzed, based on automatic rod control versus no rod control, and minimum versus maximum reactivity feedback parameters.
Evaluation of this transient using the proposed reduction in RCS flow has concluded that the Departure From Nucleate Boiling Ratio (DNBR) design basis is met. An evaluation of the maximum primary and secondary system pressures following this transient was also performed.
This evaluation concluded the pressures would not be significantly affected by the proposed flow reduction.
Loss of Offsite Power (UFSAR Section 15.2.9) An evaluation of the proposed reductions in RCS flow for the Loss of Offsite Power event shows that natural circulation core cooling would not be significantly affected.
The DNBR remains above the safety limit, the licensing basis criteria for primary and secondary system pressure continues to be met, and pressurizer filling would not occur. Loss of Normal Feedwater (UFSAR Section 15.2.8) Evaluation of this event shows that sufficient margin exists relative to primary and secondary peak pressures and pressurizer NLR-N94016 Attachment 1 filling, such that the conclusions stated in the UFSAR remain valid. Feedwater System Pipe Break (UFSAR Section 15.4.3) Evaluation of this event shows that the proposed reduction in RCS flow would result in a small decrease in steam generator mass, and no significant impact on peak hot leg temperatures.
The current licensing basis analyses contain sufficient margin to accommodate the decay heat removal penalty associated with the proposed reduction in RCS flow. Partial Loss of Forced RCS Flow (UFSAR Section 15.2.5) Complete Loss of Forced RCS Flow (UFSAR Section 15.3.4) Evaluation of these transients shows that the effects of the proposed reduction in RCS flow on DNB and RCS pressure can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid. Reactor Coolant Pump Shaft Seizure (Locked Rotor) Evaluation of this transient shows that the effects of the proposed reduction in RCS flow on peak fuel clad temperature and RCS pressure can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid. Rod Withdrawal from Subcritical Condition (UFSAR Section 15.2.1) Rod Withdrawal at Power (UFSAR Section 15.2.2) Rod Cluster Control Assembly Misalignment (UFSAR Section 15.2.3 and 15.3.5) Startup of an Inactive Loop (UFSAR Section 15.2.6) Evaluation of these transients shows that the effects of the proposed reduction in RCS flow on DNB can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid. Uncontrolled Boron Dilution (UFSAR Section 15.2.4) Thermal design flow is not an input to the boron dilution analysis.
Therefore, the conclusions presented in the UFSAR are not affected by the proposed reduction in RCS flow. Rupture of a Control Rod Drive Mechanism Housing (UFSAR Section 15 .4. 7) In order to demonstrate that gross fuel damage would not occur, the core would remain in a coolable geometry, and the RCS would remain intact, the following more restrictive criteria are applied to this event:
NLR-N94016 Attachment 1 1) The average fuel pellet enthalpy at the hot spot is less than 200 cal/gm (360 Btu/lbm) . 2) Fuel melt at the hot spot is limited to less than the innermost 10% of the fuel pellet. 3) Peak RCS pressure is less than that which would cause stresses to exceed the Faulted Condition Limits. Evaluation of this transient shows that the effects of the proposed reduction in RCS flow on peak fuel clad temperature and RCS pressure can be accommodated by existing margins, such that the above criteria would continue to be met, and the conclusions in the UFSAR remain valid. Spurious Operation of Safety Injection at Power (UFSAR Section 15.2.14) Since the transient conditions of this event are not significantly altered by the proposed reduction in RCS flow, the conclusions in the UFSAR remain valid. Accidental Depressurization of the RCS (UFSAR Section 15.2.12) Since the transient conditions of this event are not significantly altered by the proposed reduction in RCS flow, and the OT-delta-T setpoint is not changed, the conclusions in the UFSAR remain valid. IV. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION In accordance with lOCFRS0.92, PSE&G has reviewed the proposed changes and concluded the proposed changes do not involve a significant hazards consideration because the changes would not: 1. Involve a significant increase in the probability or consequences of an accident previously analyzed.
No component modification, system realignment, or change in operations will occur which could affect the probability of any accident or transient.
The proposed reduction in RCS loop and total flow rates will not change the probability of a challenge to any Engineered Safeguard Feature or other device. The consequences of previously analyzed accidents have been found to remain within acceptable licensing basis limits when the reduced flow rates are assumed. The system transient response is not affected by the initial RCS flow assumption, unless the initial assumption is so low as to impair the steady-state core cooling capability or steam generator heat transfer capability.
This is clearly not the case with a 1% reduction in RCS flow. The proposed change to the wording of the parameter title on Table 3.2-1 is editorial for clarity. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously analyzed.
NLR-N94016 Attachment 1 2. Create the possibility of a new or different kind of accident.
No component modification, system realignment, or change in operating procedure will occur which could create the possibility of a new event not previously considered.
The proposed reduction in RCS loop and total flow rates will not initiate any new events. Therefore, the proposed changes would not create the possibility of a different or new kind of accident.
- 3. Involve a significant reduction in a margin of safety. The proposed decrease in RCS loop and total flow rates has been analyzed and found to have an insignificant effect on the applicable transient analyses found in the FSAR. The proposed change to the wording of the parameter title on Table 3.2-1 is editorial for clarity. Therefore, the proposed changes would not involve a significant reduction in any margin of safety. Therefore, based on the information presented above, PSE&G has concluded there is no significant hazards consideration.