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| {{#Wiki_filter:Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 . Vice President and Chief Nuclear Officer November 28, 1988 NLR-N88195 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen: | | {{#Wiki_filter:Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 . |
| RESPONSE TO NRC GENERIC LETTER 88-11 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Generic Letter 88-11 forwards Regulatory Guide 1.99, Revision 2 for implementation by reactor licensees. | | Vice President and Chief Nuclear Officer November 28, 1988 NLR-N88195 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen: |
| Therein, the Commission requests that the results of analyses performed*in accordance with Rev. 2 of the guide be submitted for Commission review. A proposed schedule for implementation of any actions required as a result of the changes in methodology contained in Rev. 2 is also requested. | | RESPONSE TO NRC GENERIC LETTER 88-11 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Generic Letter 88-11 forwards Regulatory Guide 1.99, Revision 2 for implementation by reactor licensees. Therein, the Commission requests that the results of analyses performed*in accordance with Rev. 2 of the guide be submitted for Commission review. A proposed schedule for implementation of any actions required as a result of the changes in methodology contained in Rev. 2 is also requested. Additionally, for Pressurized Water Reactor plants, the Commission identifies the need to review the Low Temperature Overpressure Protection system for potential changes resulting from pressure/temperature limits established using Revision 2 of the guide. This latter issue is applicable to Salem Units 1 and 2 and is addressed in Enclosure 1 to this transmittal. |
| Additionally, for Pressurized Water Reactor plants, the Commission identifies the need to review the Low Temperature Overpressure Protection system for potential changes resulting from pressure/temperature limits established using Revision 2 of the guide. This latter issue is applicable to Salem Units 1 and 2 and is addressed in Enclosure 1 to this transmittal. | | Accordingly, Public Service Electric and Gas Company hereby forwards Enclosures 1 and 2 in response to the subject Generic Letter. Enclosure 1 pertains to Salem Units 1 and 2. Enclosure 2 is specific to Hope Creek. |
| Accordingly, Public Service Electric and Gas Company hereby forwards Enclosures 1 and 2 in response to the subject Generic Letter. Enclosure 1 pertains to Salem Units 1 and 2. Enclosure 2 is specific to Hope Creek. If there are any questions regarding the enclosed information, please feel free to contact us. Sincerely, Enclosures | | If there are any questions regarding the enclosed information, please feel free to contact us. |
| -,
| | Sincerely, Enclosures |
| Document Control Desk C Mr. G. W. Rivenbark Licensing Project Manager Mr. J. C. Stone Licensing Project Manager Ms. K. Halvey Gibson 2 Senior Resident Inspector | | |
| -Salem (Acting) Mr. G. W. Meyer Senior Resident Inspector | | Document Control Desk 2 11-28-88 C Mr. G. W. Rivenbark Licensing Project Manager Mr. J. C. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector - Salem (Acting) |
| -Hope Creek Mr. W. T. Russell, Administrator Region I Ms. J. Moon, Interim Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 11-28-88 ENCLOSURE 1 PSE&G RESPONSE TO THE NRC GENERIC LETTER 88-11 SALEM UNITS 1 & 2 (1) Results Summary PSE&G has assessed the impact of Regulatory Guide 1.99, Rev. 2 and the attached Regulatory Guide 1.99, Rev. 2, on the pressure temperature limits contained in the Technical Specifications for Salem Units 1 and 2. This review was completed along with the analyses performed to monitor the changes in reactor vessel material fracture toughness based on removal of surveillance capsule Z during Salem Unit l's 7th refueling outage and removal of capsule U during Salem Unit 2's 3rd refueling outage. Results of the technical analysis which was conducted using the methods described in Regulatory Guide 1.99, Rev. 2, were submitted to the NRC previously in Westinghouse reports WCAP 11544 and 11955. These reports are specific to Salem Units 1 and 2. The heatup and cooldown curves generated as part of the Westinghouse analyses referenced above, reflect the pressure/temperature (P-T) limits for Salem Units 1 and 2 and are based on the requirements of Appendix G to 10CFR50 and Regulatory Guide 1.99, Rev. 2. The P-T curves for Salem Unit 1 were prepared by Westinghouse in September 1988. The P-T curves for Salem Unit 2 were prepared by Westinghouse in November 1987 using the draft version of Regulatory Guide 1.99, Rev. 2 available at that time. Westinghouse has reviewed the Salem Unit 2 P-T curves and concluded that they are in compliance with the final version of Regulatory Guide 1. 99, Rev. 2. Additionally, the low temperature overpressure protection (LTOP) system for Salem Units 1 and 2 has been reviewed to determine the impact of the revised P-T curves on LTOP setpoint, enable temperature, system hardware, and applicable operating procedures. | | Mr. G. W. Meyer Senior Resident Inspector - Hope Creek Mr. W. T. Russell, Administrator Region I Ms. J. Moon, Interim Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 |
| The results of that review are summarized in the following paragraphs. | | |
| At Salem Units 1 and 2, this system is referred to as the Pressurizer Overpressure Protection System (POPS). The POPS is a two train system which uses separate and independent pressure transmitters to open two pressurizer relief valves if RCS pressure exceeds a preset value of 375 psig. The POPS is required to be armed whenever the RCS is below 312°F. The POPS relief valves protect the RCS from pressure transients exceeding the limits of Appendix G to 10CFR50 when one or more RCS cold leg temperature is at or below 312°F. Either POPS Power Operated Relief Valve (PORV) has adequate relieving capacity to protect the RCS from 1 e Enclosure 1 (Continued) overpressurization as a result of the limiting heat input or mass input cases; i.e. (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50°F above RCS cold leg temperature or (2) the start of a safety injection pump and its injection into the water solid RCS. Several provisions presently exist for prevention of pressure transients when the RCS temperature is below 312°F. Current Technical Specification 3.4.1.3 for startup of an RCP requires that a steam bubble must be established in the pressurizer prior to pump start or the SG/RCS delta-T be verified to be less than 50°F. Also, Technical Specification 3.5.3 allows a maximum of one safety injection pump to remain operable and power to all inoperable safety injection pumps must be removed by racking out the power supply breakers when the RCS temperature is below 312°F. Also the shutdown procedure requires that a steam bubble be maintained in the pressurizer during plant cooldowns. | | ENCLOSURE 1 PSE&G RESPONSE TO THE NRC GENERIC LETTER 88-11 SALEM UNITS 1 & 2 (1) Results Summary PSE&G has assessed the impact of Regulatory Guide 1.99, Rev. |
| The Residual Heat Removal (RHR) System is put into service once the RCS temperature is below 350°F but above 312°F. The RHR system provides relief capacity comparable to that of a POPS valve. However, no credit has been taken in the low temperature overpressure analysis for this relief capacity. | | 2 and the attached Regulatory Guide 1.99, Rev. 2, on the pressure temperature limits contained in the Technical Specifications for Salem Units 1 and 2. This review was completed along with the analyses performed to monitor the changes in reactor vessel material fracture toughness based on removal of surveillance capsule Z during Salem Unit l's 7th refueling outage and removal of capsule U during Salem Unit 2's 3rd refueling outage. Results of the technical analysis which was conducted using the methods described in Regulatory Guide 1.99, Rev. 2, were submitted to the NRC previously in Westinghouse reports WCAP 11544 and 11955. |
| Based on the above analysis, it is determined that changes to the pressurizer overpressure protection system are not required as a result of the revised P-T limits. (2) Schedule For Implementation of R.G. 1.99, Rev. 2 As indicated previously, R.G. 1.99, Rev. 2 has been used for the preparation of revised P-T curves and operating limits for Salem Units 1 and 2. PSE&G will submit License Amendment Requests to incorporate new heatup and cooldown curves in the Technical Specifications of Salem Units 1 & 2 by December 1988. As part of those License Amendment requests, appropriate changes to the Bases section will also be provided. | | These reports are specific to Salem Units 1 and 2. |
| Any required modifications to the Salem Updated Final Safety Analysis Report will be submitted as part of the next routine annual update presently scheduled for July 1989. 2 ENCLOSURE 2 PSE&G RESPONSE TO NRC GENERIC LETTER 88-11 HOPE CREEK (1) Results Summary The impact of implementing R.G. 1.99 Rev. 2 can best be determined by comparing the RTNDT adjusted reference temperature (ART) values based on R.G. 1.99, Rev. 1 and R.G. 1.99, Rev. 2. Table 1 shows the ART values at 4 EFPY and at 32 EFPY for each beltline material in Hope Creek. These EFPY values are selected to be representative of conditions existing early in the operating life and near the end of the operating life of the plant. The following conclusions are drawn from the results in this table: (a) The R.G. 1.99, Rev. 2 ART values at 32 EFPY for this unit are below 200°F, which is the allowable limit in 10CFR50, Appendix G. Therefore, implementation of R.G. 1.99, Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing. (b) The ART value which applies to the pressure-temperature (P-T) curves in the Technical Specifications is 39°F at 32 EFPY as determined using R.G. 1.99, Rev. 1 methods. The maximum R.G. 1.99, Rev. 2 ART value in Table 1 is 65.8°F at 32 EFPY. Therefore, at 32 EFPY the R.G. 1.99, Rev. 1 P-T curves are less conservative than P-T curves that would be generated with R.G. 1.99, Rev. 2. However, the current P-T curves are applicable up to 6.7 EFPY even if the ART is calculated according to R.G. 1.99, Rev. 2 methods. Therefore, the P-T curves presently contained in the Technical Specifications are conservative for several more years of operation. (c) The worst case LPCI nozzle and weld are also included in this beltline region analysis due to their predicted fluence value at 32 EFPY. Since the maximum ART of 13.1°F at 32 EFPY based on Rev. 2, is less than the 40°F RTNDT value applied to the limiting vessel discontinuity curves, the LPCI nozzle is bounded by the limiting vessel discontinuity curves. Therefore, the vessel discontinuity limits on the P-T curves will not be changed through 32 EFPY of operation. | | The heatup and cooldown curves generated as part of the Westinghouse analyses referenced above, reflect the pressure/temperature (P-T) limits for Salem Units 1 and 2 and are based on the requirements of Appendix G to 10CFR50 and Regulatory Guide 1.99, Rev. 2. The P-T curves for Salem Unit 1 were prepared by Westinghouse in September 1988. The P-T curves for Salem Unit 2 were prepared by Westinghouse in November 1987 using the draft version of Regulatory Guide 1.99, Rev. 2 available at that time. Westinghouse has reviewed the Salem Unit 2 P-T curves and concluded that they are in compliance with the final version of Regulatory Guide |
| (2) Schedule For Implementation Of R.G. 1.99, Rev. 2 Generic Letter 88-11 requires that R.G. 1.99, Rev. 2 be implemented within two outages after May, 1988. Based on the results of our analyses, the following implementation schedule is proposed: (a) The current P-T curves are conservative for up to 6.7 EFPY of operation. | | : 1. 99, Rev. 2. |
| However, because the 1 e Enclosure 2 (Continued) implementation of R.G. 1.99, Rev. 2 results in calculated 32 EFPY ART values more conservative than those of R.G. 1.99, Rev. 1 at 32 EFPY, the P-T curves will be revised within two refueling outages (but not later than 6.7 EFPY). Further, since a neutron dosimeter capsule has been removed for testing, revision of the P-T curves will be combined with the results of this testing effort. In this way, implementation of R.G. 1.99, Rev. 2 will be based upon verified, plant specific fluence predictions. (b) Upon completion of the revisions to the P-T curves, appropriate changes will be made to Section 5.3 and Appendix 5A of the Hope Creek Updated FSAR, and also to Section 3/4.4.6 and the corresponding Bases section of the Hope Creek Technical Specifications. | | Additionally, the low temperature overpressure protection (LTOP) system for Salem Units 1 and 2 has been reviewed to determine the impact of the revised P-T curves on LTOP setpoint, enable temperature, system hardware, and applicable operating procedures. The results of that review are summarized in the following paragraphs. |
| The Technical Specification changes will be submitted prior to completion of the second Refueling outage. Changes to the Updated FSAR will be incorporated as part of the next routine annual update following submittal of the Technical Specification changes. 2 | | At Salem Units 1 and 2, this system is referred to as the Pressurizer Overpressure Protection System (POPS). The POPS is a two train system which uses separate and independent pressure transmitters to open two pressurizer relief valves if RCS pressure exceeds a preset value of 375 psig. The POPS is required to be armed whenever the RCS is below 312°F. |
| ...........
| | The POPS relief valves protect the RCS from pressure transients exceeding the limits of Appendix G to 10CFR50 when one or more RCS cold leg temperature is at or below 312°F. Either POPS Power Operated Relief Valve (PORV) has adequate relieving capacity to protect the RCS from 1 |
| \f * --' * "' Table 1 COMPARISON OF REV 1 AND REV 2 ART VALUES FOR HOPE CREEK 4 EFPY Rev 1 Rev 2 Rev 1 32 EFPY Rev 2 Beltline Component ART c*r> ART C-F> ART C°F) ART {°F) Plates: 5K3025-l 26.1 29.8 39.0 64.2 5K2608-l 22.0 24.5 27.5 42.2 5K2698-l 22.8 25.2 29.8 45.0 5K2963-l -4.7 2.5 4.9 29.7 5K2530-l 24.9 33.5 35.6 65.0 5K3238-1 15.2 23.4 30.2 59.3 51(3230-1 | | |
| -4.l 2.5 6.6 29.7 6C35-l -4.0 5.4 8.9 41. 3 6C45-l 5.7 15.5 14.3 47.0 Beltline Welds: 510-01205 | | e (Continued) overpressurization as a result of the limiting heat input or mass input cases; i.e. (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50°F above RCS cold leg temperature or (2) the start of a safety injection pump and its injection into the water solid RCS. |
| -33.0 -9.2 -20.1 58.0 053040/1125-02205 | | Several provisions presently exist for prevention of pressure transients when the RCS temperature is below 312°F. |
| -24.l 0.1 -13.4 65.8 519-01205 | | Current Technical Specification 3.4.1.3 for startup of an RCP requires that a steam bubble must be established in the pressurizer prior to pump start or the SG/RCS delta-T be verified to be less than 50°F. Also, Technical Specification 3.5.3 allows a maximum of one safety injection pump to remain operable and power to all inoperable safety injection pumps must be removed by racking out the power supply breakers when the RCS temperature is below 312°F. |
| -46.3 -47.1 -41.3 -41.0 504-01205 | | Also the shutdown procedure requires that a steam bubble be maintained in the pressurizer during plant cooldowns. |
| -28.0 -29.l -22.5 -23.0 055733/1810-02205 | | The Residual Heat Removal (RHR) System is put into service once the RCS temperature is below 350°F but above 312°F. |
| -35.4 -27.9 10.6 053040/1810-02205 | | The RHR system provides relief capacity comparable to that of a POPS valve. However, no credit has been taken in the low temperature overpressure analysis for this relief capacity. |
| -44.6 -36.9 -36.7 1.6 LPCI Nozzles and Weld: 19468-1 (nozzle) -8.0 7.1 10024-1 (nozzle) -6.1 13.1 001-01205 (weld) -32.4 -31. 5}} | | Based on the above analysis, it is determined that changes to the pressurizer overpressure protection system are not required as a result of the revised P-T limits. |
| | (2) Schedule For Implementation of R.G. 1.99, Rev. 2 As indicated previously, R.G. 1.99, Rev. 2 has been used for the preparation of revised P-T curves and operating limits for Salem Units 1 and 2. PSE&G will submit License Amendment Requests to incorporate new heatup and cooldown curves in the Technical Specifications of Salem Units 1 & 2 by December 1988. As part of those License Amendment requests, appropriate changes to the Bases section will also be provided. Any required modifications to the Salem Updated Final Safety Analysis Report will be submitted as part of the next routine annual update presently scheduled for July 1989. |
| | 2 |
| | |
| | ENCLOSURE 2 PSE&G RESPONSE TO NRC GENERIC LETTER 88-11 HOPE CREEK (1) Results Summary The impact of implementing R.G. 1.99 Rev. 2 can best be determined by comparing the RTNDT adjusted reference temperature (ART) values based on R.G. 1.99, Rev. 1 and R.G. |
| | 1.99, Rev. 2. Table 1 shows the ART values at 4 EFPY and at 32 EFPY for each beltline material in Hope Creek. These EFPY values are selected to be representative of conditions existing early in the operating life and near the end of the operating life of the plant. The following conclusions are drawn from the results in this table: |
| | (a) The R.G. 1.99, Rev. 2 ART values at 32 EFPY for this unit are below 200°F, which is the allowable limit in 10CFR50, Appendix G. Therefore, implementation of R.G. |
| | 1.99, Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing. |
| | (b) The ART value which applies to the pressure-temperature (P-T) curves in the Technical Specifications is 39°F at 32 EFPY as determined using R.G. 1.99, Rev. 1 methods. |
| | The maximum R.G. 1.99, Rev. 2 ART value in Table 1 is 65.8°F at 32 EFPY. Therefore, at 32 EFPY the R.G. |
| | 1.99, Rev. 1 P-T curves are less conservative than P-T curves that would be generated with R.G. 1.99, Rev. 2. |
| | However, the current P-T curves are applicable up to 6.7 EFPY even if the ART is calculated according to R.G. 1.99, Rev. 2 methods. Therefore, the P-T curves presently contained in the Technical Specifications are conservative for several more years of operation. |
| | (c) The worst case LPCI nozzle and weld are also included in this beltline region analysis due to their predicted fluence value at 32 EFPY. Since the maximum ART of 13.1°F at 32 EFPY based on Rev. 2, is less than the 40°F RTNDT value applied to the limiting vessel discontinuity curves, the LPCI nozzle is bounded by the limiting vessel discontinuity curves. Therefore, the vessel discontinuity limits on the P-T curves will not be changed through 32 EFPY of operation. |
| | (2) Schedule For Implementation Of R.G. 1.99, Rev. 2 Generic Letter 88-11 requires that R.G. 1.99, Rev. 2 be implemented within two outages after May, 1988. Based on the results of our analyses, the following implementation schedule is proposed: |
| | (a) The current P-T curves are conservative for up to 6.7 EFPY of operation. However, because the 1 |
| | |
| | e (Continued) implementation of R.G. 1.99, Rev. 2 results in calculated 32 EFPY ART values more conservative than those of R.G. 1.99, Rev. 1 at 32 EFPY, the P-T curves will be revised within two refueling outages (but not later than 6.7 EFPY). Further, since a neutron dosimeter capsule has been removed for testing, revision of the P-T curves will be combined with the results of this testing effort. In this way, implementation of R.G. 1.99, Rev. 2 will be based upon verified, plant specific fluence predictions. |
| | (b) Upon completion of the revisions to the P-T curves, appropriate changes will be made to Section 5.3 and Appendix 5A of the Hope Creek Updated FSAR, and also to Section 3/4.4.6 and the corresponding Bases section of the Hope Creek Technical Specifications. The Technical Specification changes will be submitted prior to completion of the second Refueling outage. Changes to the Updated FSAR will be incorporated as part of the next routine annual update following submittal of the Technical Specification changes. |
| | 2 |
| | |
| | \f * -- ' * "' |
| | Table 1 COMPARISON OF REV 1 AND REV 2 ART VALUES FOR HOPE CREEK 4 EFPY 32 EFPY Rev 1 Rev 2 Rev 1 Rev 2 Beltline Component ART c*r> ART C-F> ART C°F) ART {°F) |
| | Plates: |
| | 5K3025-l 26.1 29.8 39.0 64.2 5K2608-l 22.0 24.5 27.5 42.2 5K2698-l 22.8 25.2 29.8 45.0 5K2963-l -4.7 2.5 4.9 29.7 5K2530-l 24.9 33.5 35.6 65.0 5K3238-1 15.2 23.4 30.2 59.3 51(3230-1 -4.l 2.5 6.6 29.7 6C35-l -4.0 5.4 8.9 41. 3 6C45-l 5.7 15.5 14.3 47.0 Beltline Welds: |
| | 510-01205 -33.0 -9.2 -20.1 58.0 053040/1125-02205 -24.l 0.1 -13.4 65.8 519-01205 -46.3 -47.1 -41.3 -41.0 504-01205 -28.0 -29.l -22.5 -23.0 055733/1810-02205 -35.4 -27.9 -26~9 10.6 053040/1810-02205 -44.6 -36.9 -36.7 1.6 LPCI Nozzles and Weld: |
| | 19468-1 (nozzle) -8.0 7.1 10024-1 (nozzle) -6.1 13.1 001-01205 (weld) -32.4 -31. 5}} |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M7571999-10-22022 October 1999 Advises That Attachment 1 to ,Marked as Proprietary,Re Safety Limit MCPR & Fuel Vendor Change Will Be Withheld from Public Disclosure Per 10CFR2.790(b)(5) ML20217M2101999-10-19019 October 1999 Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates ML20217H8471999-10-18018 October 1999 Discusses Completion of Licensing Action for GL 98-01 & Suppl 1, Yr 2000 Readiness of Computer Sys at Npps, to All Holders of Operating Licenses for NPPs ML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML20217K8441999-10-15015 October 1999 Submits Revised Commitment to NRC Bulletin 90-01,Suppl 1 for Hope Creek Generating Station ML20217H9771999-10-13013 October 1999 Forwards SRO & RO Initial Exam Rept 50-354/98-302,suppl Rept on 990125-29,mtg Meeting on 990322,990429-30 & 0617-18 in-office Review & 990720 Telcon on Appeal Results.Overall, 11 of 16 Applicants Received NRC Licenses ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML20217C4391999-10-0606 October 1999 Informs That Util Authorized to Administer Initial NRC Retake Written Exam to Applicant Listed,During Week of 991011 ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 ML20217A9601999-10-0404 October 1999 Forwards Errata Redressing Deficiencies & Correcting Two Typos to Ufsar,Rev 10.Incorporate Attached Pages/Figures Into Controlled Copies of UFSAR ML20217A6861999-10-0101 October 1999 Forwards Insp Rept 50-354/99-05 on 990711-0829.Four Violations Occurred Re Areas of Fire Protection,Operation at Reduced Feedwater Inlet Temp & safety-related Battery Charging Operation & Being Treated as NCVs LR-N990430, Forwards Rev 10 to Hope Creek Generating Station Ufsar,Iaw 10CFR50.71(e).Details Re Each Change Also Attached to Facilitate NRC Review1999-09-28028 September 1999 Forwards Rev 10 to Hope Creek Generating Station Ufsar,Iaw 10CFR50.71(e).Details Re Each Change Also Attached to Facilitate NRC Review IR 05000272/19990071999-09-28028 September 1999 Forwards Insp Repts 50-272/99-07 & 50-306/99-07 on 990721- 0831.One Potentially Safety Significant Issue Identified Dealing with Control Room Special Ventilation System.Four Addl Issues of Low Safety Significance Identified ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) ML18107A5421999-09-22022 September 1999 Forwards Discharge Monitoring Rept for Salem Generating Station for Aug 1999.Rept Is Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection 05000354/LER-1999-009, Forwards LER 99-009-00, License Condition Violation - Min FW Temp Limits. Commitments Made by Util Encl1999-09-20020 September 1999 Forwards LER 99-009-00, License Condition Violation - Min FW Temp Limits. Commitments Made by Util Encl ML20217K7781999-09-16016 September 1999 Forwards Discharge Monitoring Rept for Hope Creek Generating Station for Month of Aug 1999. Rept Is Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML20212B4021999-09-13013 September 1999 Submits Supplemental Info Related to Hope Creek License Change Request (LCR) H98-08,submitted to NRC on 981230, Re Flood Protection TS Changes ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML20211N5421999-09-0808 September 1999 Forwards Amend 121 to License NPF-57 & Safety Evaluation. Amend Revises TSs by Relocating Procedural Details of RETS to Offsite Dose Calculation Manual LR-N990395, Provides Comments on NRC Ltr Dtd 990714, Closure of TAC Number MA1194 - Response to RAIs to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity, for Hope Creek Generating Sation. Revised GE Report Encl Also1999-09-0101 September 1999 Provides Comments on NRC Ltr Dtd 990714, Closure of TAC Number MA1194 - Response to RAIs to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity, for Hope Creek Generating Sation. Revised GE Report Encl Also ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML20211B5341999-08-20020 August 1999 Forwards RAI Re 2nd 10-yr ISI Interval Relief Requests Re Plant.Info Requested to Be Provided within 60 Days of Receipt of Ltr ML18107A4861999-08-19019 August 1999 Forwards NPDES Discharge Monitoring Rept, for Salem Generating Station for Month of Jul 1999.Rept Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML20210R4911999-08-11011 August 1999 Forwards Insp Rept 50-354/99-04 on 990530-0711.No Violations Noted.Inspectors Reviewed Performance Indicators Submitted as Part of Pilot Program for New Regulatory Oversight Process & Verified Data ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML20210H9241999-07-26026 July 1999 Informs That State of Nj Dept of Environ Protection Has No Comments on Licensee 990517 Request for Amend to TS by Adding TS 3.3.10, Instrumentation of OPRM Sys ML20210F3271999-07-22022 July 1999 Forwards SE Granting Relief Requests RR-B1,RR-C1,RR-D1 & RR-B3 Re First 10-year Interval for ISI Program at Hope Creek ML18107A4531999-07-20020 July 1999 Forwards Discharge Monitoring Rept for Salem Generating Station, for June 1999.Rept Is Required by & Prepared for EPA & Nj Dept of Environ Protection ML20210D3971999-07-16016 July 1999 Forwards Discharge Monitoring Rept for Hope Creek Generating Station, for June 1999.Rept Is Required by & Prepared for EPA & Nj Dept of Environ Protection 05000354/LER-1999-007, Forwards LER 99-007-00,re License Condition Violation - Class-1E Battery Charging Operation.Commitments Made by Util Encl1999-07-14014 July 1999 Forwards LER 99-007-00,re License Condition Violation - Class-1E Battery Charging Operation.Commitments Made by Util Encl ML20209G2831999-07-14014 July 1999 Disclosure Closure of TAC MA1194 Re Licensee Response to RAI to GL 92-01,Rev 1,Suppl 1, Rc Structural Integrity, for Plant LR-N990250, Provides Proposed Alternative & Supporting Justification for Relief from Augmented Inservice Requirements of 10CFR50.55a(g) for Volumetric Exam of RPV Circumferential Welds1999-07-0909 July 1999 Provides Proposed Alternative & Supporting Justification for Relief from Augmented Inservice Requirements of 10CFR50.55a(g) for Volumetric Exam of RPV Circumferential Welds ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML20217K8441999-10-15015 October 1999 Submits Revised Commitment to NRC Bulletin 90-01,Suppl 1 for Hope Creek Generating Station ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML20217A9601999-10-0404 October 1999 Forwards Errata Redressing Deficiencies & Correcting Two Typos to Ufsar,Rev 10.Incorporate Attached Pages/Figures Into Controlled Copies of UFSAR LR-N990430, Forwards Rev 10 to Hope Creek Generating Station Ufsar,Iaw 10CFR50.71(e).Details Re Each Change Also Attached to Facilitate NRC Review1999-09-28028 September 1999 Forwards Rev 10 to Hope Creek Generating Station Ufsar,Iaw 10CFR50.71(e).Details Re Each Change Also Attached to Facilitate NRC Review ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) 05000354/LER-1999-009, Forwards LER 99-009-00, License Condition Violation - Min FW Temp Limits. Commitments Made by Util Encl1999-09-20020 September 1999 Forwards LER 99-009-00, License Condition Violation - Min FW Temp Limits. Commitments Made by Util Encl ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML20212B4021999-09-13013 September 1999 Submits Supplemental Info Related to Hope Creek License Change Request (LCR) H98-08,submitted to NRC on 981230, Re Flood Protection TS Changes ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations LR-N990395, Provides Comments on NRC Ltr Dtd 990714, Closure of TAC Number MA1194 - Response to RAIs to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity, for Hope Creek Generating Sation. Revised GE Report Encl Also1999-09-0101 September 1999 Provides Comments on NRC Ltr Dtd 990714, Closure of TAC Number MA1194 - Response to RAIs to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity, for Hope Creek Generating Sation. Revised GE Report Encl Also ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML20210H9241999-07-26026 July 1999 Informs That State of Nj Dept of Environ Protection Has No Comments on Licensee 990517 Request for Amend to TS by Adding TS 3.3.10, Instrumentation of OPRM Sys 05000354/LER-1999-007, Forwards LER 99-007-00,re License Condition Violation - Class-1E Battery Charging Operation.Commitments Made by Util Encl1999-07-14014 July 1999 Forwards LER 99-007-00,re License Condition Violation - Class-1E Battery Charging Operation.Commitments Made by Util Encl LR-N990250, Provides Proposed Alternative & Supporting Justification for Relief from Augmented Inservice Requirements of 10CFR50.55a(g) for Volumetric Exam of RPV Circumferential Welds1999-07-0909 July 1999 Provides Proposed Alternative & Supporting Justification for Relief from Augmented Inservice Requirements of 10CFR50.55a(g) for Volumetric Exam of RPV Circumferential Welds ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments ML18107A4181999-06-30030 June 1999 Submits Response to NRC Request for Info Re Y2K Readiness at Npps,Per GL 98-01,suppl 1.Disclosure Encl LR-N990316, Responds to NRC Request for Info Re Y2K Readiness at Npps, Per GL 98-01,suppl 1.Disclosure Encl1999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Npps, Per GL 98-01,suppl 1.Disclosure Encl ML18107A4131999-06-25025 June 1999 Provides Further Clarification of Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting Station SBO & Loca/Loop Loading Requirements,Per Util 990426 Ltr & Discussion with NRC ML18107A4061999-06-21021 June 1999 Provides Supplemental Info to Proposed License Amend, Modifying TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Info Re Acceptance Criterion Discussed ML20209B6441999-06-21021 June 1999 Offers No Comments on Licensee 990529 Request for Revs to Plant Radiological Effluent Ts,Per GL 89-01 ML20196E6471999-06-21021 June 1999 Forwards Revised marked-up TS Page for HCGS License Change Requests H99-02 & H99-05,dtd 990329 & 0524,respectively. Revised Pages Do Not Alter Conclusions Reached in 10CFR50.92 No Significant Hazards Analysis Previously Submitted ML20196E9631999-06-17017 June 1999 Informs That Util Has Made Change to Commitment Stated in NRC Ser,Suppl 5.Commitment That Has Been Changed Is Item Number 1 of First Paragraph on Page 9-3 of Ser,Suppl 5 LR-N990295, Submits Change 1 to Relief Request RR-A4,which Clarifies Requirements Re Snubber Visual Insps.Request Was Submitted as Part of Plant Second Interval ISI Program on 9905111999-06-16016 June 1999 Submits Change 1 to Relief Request RR-A4,which Clarifies Requirements Re Snubber Visual Insps.Request Was Submitted as Part of Plant Second Interval ISI Program on 990511 05000354/LER-1999-006, Forwards LER 99-006-00 Re Esfa B Channel Primary Containment Isolation Signal Actuation.Attachment a Lists Commitments Util Making to NRC Re LER1999-06-15015 June 1999 Forwards LER 99-006-00 Re Esfa B Channel Primary Containment Isolation Signal Actuation.Attachment a Lists Commitments Util Making to NRC Re LER ML18107A3691999-06-11011 June 1999 Forwards Corrected Monthly Operating Rept for Apr 1999 for Salem Generating Station,Unit 1.Original Submittal Contained Typo for year-to-date Value for Numbers of Hours Generator Was on Line (Service Hours) ML18107A3641999-06-0404 June 1999 Requests Enforcement Discretion for TS 3/4.6.2.3 Re Containment Cooling Sys for Salem Generating Station,Unit 1 ML18107A3561999-06-0303 June 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S99-05 & Has No Comments ML18107A3611999-05-27027 May 1999 Forwards Responses to NRC 990301 & 990323 RAIs for Salem & Hope Creek Generating Stations Relating to GL 96-05 ML18107A3301999-05-24024 May 1999 Forwards Suppl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety Related Power Operated Gate Valves. Encl Contains Methodology Used in Determination of Pressure Locking Susceptibility of PORVs Block Valves ML18107A3291999-05-20020 May 1999 Forwards Redacted Response to NRC 990322 RAI Re Notification of Licensed Operator That Tested Positive for Alcohol. Attachment 2 Withheld,Per 10CFR2.790(a)(6) ML18107A3031999-05-18018 May 1999 Provides Summary of Changes to NRC Commitments That Have Been Made But Not Reported by Other Means,Iaw with NEI Process for Managing NRC Commitments ML18107A2891999-05-13013 May 1999 Forwards Rev 36 to Pse&G Nuclear Business Unit Emergency Plan. Rev 36 Incorporates Changes to Section 1-3,6 & 7 & 9-17.Attached Copy Includes All Sections of EP for Completeness 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML18095A4881990-09-17017 September 1990 Requests Regional Waiver of Compliance from Tech Spec 3.6.2.3, Containment Cooling Sys. Waiver Requested in Order to Allow Replacement of Containment Fan Cooler Unit Motor #22 W/O Requiring Plant Shutdown ML18095A4901990-09-13013 September 1990 Provides Supplemental Info Applicable to Clarification of 10CFR50,App R Exemption Request Re Fire Suppression Sys for Panel 335,per NRC Request ML20059H5391990-09-10010 September 1990 Forwards Data Re Bailey Solid State Logic Modules (Sslm) Reliability Program for Period Ending June 1990.Quarterly Summaries of Sslm Data Will Be Provided to End of June 1991 ML20059E6821990-09-0404 September 1990 Forwards Info Re Temporary Mod to Security Plan Concerning Protected Area.Info Withheld ML20059E6791990-09-0404 September 1990 Forwards Security Upgrade Project Status Rept,Per Regulatory Effectiveness Review.Rept Withheld (Ref 10CFR73.21) ML18095A4621990-08-31031 August 1990 Provides Revised Response to Generic Ltr 89-13, Svc Water Problems Affecting Safety-Related Equipment. Only HXs Exhibiting Unsatisfactory Test Results Will Be Inspected & Possibly Cleaned ML18095A4641990-08-31031 August 1990 Forwards Revised Response to NRC Bulletin 88-004 Re Potential pump-to-pump Interaction.Util Pursuing Permanent Solution to Issue & Will Implement Appropriate Permanent Field Change by End of Unit 1 10th Refueling Outage ML20059E7021990-08-30030 August 1990 Forwards RERR-9, Hope Creek Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Rev 11 to Odcm.Summary of Change to ODCM & Rationale Provided in Part G of RERR-9.W/o Rev 11 to ODCM ML20059D2531990-08-30030 August 1990 Forwards Hope Creek Generating Station Semiannual Radioactive Effluent Release Rept RERR-9 for Jan-June 1990 & Rev 11 to ODCM for Hope Creek Generating Station ML18095A4431990-08-30030 August 1990 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept,Jan-June 1990 & Rev 6 to Odcm. ML18095A4531990-08-30030 August 1990 Forwards RERR-28, Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Revised Odcm.W/O Revised ODCM ML18095A4391990-08-29029 August 1990 Forwards Semiannual Rept Re fitness-for-duty Performance Data for 6-month Period Ending 900630,per 10CFR26.71(d).Rept Includes Testing Results,Random Testing Program Results & Confirmed Positive Tests for Specific Substances ML18095A4421990-08-28028 August 1990 Clarifies 900710 Request for Amends to Licenses DPR-70 & DPR-75,changing Sections I & M.Under Proposed Change,Section I Should Be Changed to Read Section 2.J for License DPR-70 & Section M Changed to Read Section 2.N for License DPR-75 ML20059D2191990-08-27027 August 1990 Provides Change to Schedule for Implementing Generic Ltr 88-11.Revised pressure-temp Curves Will Be Submitted Prior to Completion of Second Refueling Outage.Existing pressure- Temp Curves Conservative to 32 EFPY ML20059D2491990-08-24024 August 1990 Requests Regional Waiver of Compliance from Tech Specs 4.0.3 & 4.8.1.1.2.f.2.Waiver Allows Sufficient Time to Perform Surveillance Requirement W/O Requiring Unit Shutdown.Request Involves No Irreversible Environ Consequencies ML20059C3051990-08-24024 August 1990 Notifies of Minor Structural Enhancements to Spent Fuel Racks Issued by Util 891101 Proposed Change to Tech Spec 5.6.3 ML20059B6611990-08-22022 August 1990 Confirms That 10 Anchor/Darling Model S350W Swing Check Valves Installed at Plant,Per NRC Bulletin 89-002.All 18 Valves Inspected & Retaining Block Studs Replaced W/Upgraded Matl.No Crack Noted on Any Studs Which Were Replaced ML20059C2861990-08-21021 August 1990 Provides Correction to 900810 Response to Request for Addl Info Re Util Request for Restatement of OL Expiration Dates ML18095A4151990-08-10010 August 1990 Forwards Response to Request for Addl Info Re Reinstatement of OL Expiration Dates Based on Original Issuance of Ols. Advises That Correct Expiration Date for OL Proposed to Be 200418 ML18095A4091990-08-0909 August 1990 Forwards Responses to NRC Comments Re Plant Simulator Certification for 10CFR55.45(b)(2),per 891228 Ltr ML18095A4061990-08-0808 August 1990 Forwards Corrected marked-up Pages for Tech Spec Table 3.3-11 Re Subcooling Margin Monitor & Reactor Vessel Level Instrumentation Sys,Per 900223 Ltr.Administrative Changes Made as Indicated ML18095A3861990-07-30030 July 1990 Forwards Listing of Station Blackout Major Audit Items Resolution Scope,Per Station Blackout Schedule Commitment ML18095A3761990-07-26026 July 1990 Forwards Decommissioning Repts & Certification of Financial Assurance for Plants ML18095A3661990-07-26026 July 1990 Forwards Decommissioning Repts for Hope Creek,Peach Bottom & Salem Nuclear Generating Stations ML18095A3721990-07-24024 July 1990 Forwards Rept & Certification of Financial Assurance for Decommissioning for Plants,Per 10CFR50.75 ML18095A3611990-07-18018 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. ML18095A3621990-07-18018 July 1990 Forwards Corrected Tech Spec Page 3/4 3-5 for License Change Request 89-12 Submitted on 891227 & 900521 ML18095A3741990-07-18018 July 1990 Provides Supplemental Info Re Facility sub-cooling Margin Monitor ML18095A3751990-07-18018 July 1990 Provides Status of Commitments Made to NRC by Util in 900109 Ltr Re NUREG-0737,Item II.D.1,per 900628 Telcon ML20055G1571990-07-17017 July 1990 Requests Waiver of Compliance from Tech Spec 6.3.1 Re Requirement That Facility General Manager (Gm) Hold Senior Operator License.Waiver Will Temporarily Allow Individual to Fill Position While Gm Attends off-site Mgt Program ML18095A3591990-07-13013 July 1990 Corrects Typo in 900702 Response to Generic Ltr 90-04 Re Schedule for Completion of Remaining Open Items ML18095A3471990-07-11011 July 1990 Responds to NRC 900611 Ltr Re Violations Noted in Insp Repts 50-272/90-14 & 50-311/90-14.Corrective Actions:Directive from Radiation Protection Mgt to All Radiation Protection Personnel Issued Re Control of Compliance Agreement Sheets ML20055G4721990-07-11011 July 1990 Requests Extension Until Sept 1990 to Submit Level 1 Analysis,Per Generic Ltr 88-20 Re Individual Plant Exam Completion ML18095A3451990-07-10010 July 1990 Forwards Addl Info Re License Change Request 89-03 Re Reactor Trip Sys Instrumentation ML18095A3461990-07-10010 July 1990 Responds to NRC 900608 Ltr Re Violations Noted in Insp Repts 50-272/90-12 & 50-311/90-12.Corrective Actions:Assessment of ECCS & Component Performance Undertaken & ECCS Flow Testing Procedure Upgraded to Address Human Factors ML18095A3481990-07-10010 July 1990 Submits Supplemental Rept Identifying Root Cause of Missed Commitment & Corrective Actions to Assure Future Compliance Re Implementation of Mods to Facility PASS ML18095A3491990-07-10010 July 1990 Forwards Jn Steinmetz of Westinghouse 900614 Ltr Re Reassessment of Util Response to Bulletin 88-002 ML18095A3441990-07-0909 July 1990 Provides Written Notification Re Change in Calculated Peak Clad Temp,Per 900606 Verbal Notification ML20055E5181990-07-0303 July 1990 Suppls Request for Noncode Temporary Repair of Svc Water Sys.Ultrasonic Exam of Area Surrounding Patch Will Include Samples,In Accessible Areas,Out to Diameter of 14 Inches ML18095A3281990-07-0202 July 1990 Responds to NRC 900530 Ltr Re Violations Noted in Insp Repts 50-272/90-09 & 50-311/90-09.Corrective Actions:Util Intends to Use Nuclear Shift Supervisor as Procedure Reader & EOP, Rev 2 Under Development ML18095A3301990-07-0202 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues.Table Describing Status of Generic Safety Issue Implementation Encl ML18095A3391990-06-29029 June 1990 Forwards Correction to 890913 License Change Request 88-09, Consisting of Tech Spec Page 3/4 4-13 ML18095A3241990-06-28028 June 1990 Forwards Retyped Pages to 871224 License Change Request 87-15 & Modified,Per 900620 Ltr ML18095A3231990-06-28028 June 1990 Responds to NRC 900518 Ltr Re Violations Noted in Insp Repts 50-272/90-10,50-311/90-10 & 50-354/90-07.Two Noncited Violations Disputed.Util fitness-for-duty Program Exceeds Part 26 Requirements for Positive Blood Alcohol Limits ML18095A3221990-06-28028 June 1990 Provides Supplemental Info Re 900223 Proposed Revs to Tech Specs for Reactor Vessel Level Instrumentation Sys.Tables 3.3-11a & 3.3-11b Should Be Combined Into Single Table ML18095A3211990-06-26026 June 1990 Requests 30-day Extension Until 900730 to Provide Completion Schedule to Resolve Audit Findings Re Station Blackout ML20055D0351990-06-26026 June 1990 Forwards Addl Info Re Svc Water Sys Noncode Temporary Repair,Per 900621 Application.Util Will Perform Ultrasonic Testing of Pipe Wall Adjacent to Fillet Weld Repair & on Patch.Insps Will Be Performed at 90-day Intervals ML18095A3141990-06-25025 June 1990 Provides Schedule Change for Implementation of Control Room Mods.Schedule Modified to Address Overhead Annunciator Human Engineering Discrepancies During Phase III ML18095A3161990-06-25025 June 1990 Forwards Supplemental Info Re Response to Generic Ltr 88-14. All Committed Actions Complete as of 900613 ML18095A3201990-06-25025 June 1990 Responds to NRC 900524 Ltr Re Violations Noted in Insp Repts 50-272/90-11 & 50-311/90-11.Corrective Actions:All Overdue Operations & Maint Procedure Files Reviewed for Outstanding Rev Requests & Procedure Upgrade Program Initiated 1990-09-04
[Table view] |
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Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 .
Vice President and Chief Nuclear Officer November 28, 1988 NLR-N88195 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:
RESPONSE TO NRC GENERIC LETTER 88-11 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Generic Letter 88-11 forwards Regulatory Guide 1.99, Revision 2 for implementation by reactor licensees. Therein, the Commission requests that the results of analyses performed*in accordance with Rev. 2 of the guide be submitted for Commission review. A proposed schedule for implementation of any actions required as a result of the changes in methodology contained in Rev. 2 is also requested. Additionally, for Pressurized Water Reactor plants, the Commission identifies the need to review the Low Temperature Overpressure Protection system for potential changes resulting from pressure/temperature limits established using Revision 2 of the guide. This latter issue is applicable to Salem Units 1 and 2 and is addressed in Enclosure 1 to this transmittal.
Accordingly, Public Service Electric and Gas Company hereby forwards Enclosures 1 and 2 in response to the subject Generic Letter. Enclosure 1 pertains to Salem Units 1 and 2. Enclosure 2 is specific to Hope Creek.
If there are any questions regarding the enclosed information, please feel free to contact us.
Sincerely, Enclosures
Document Control Desk 2 11-28-88 C Mr. G. W. Rivenbark Licensing Project Manager Mr. J. C. Stone Licensing Project Manager Ms. K. Halvey Gibson Senior Resident Inspector - Salem (Acting)
Mr. G. W. Meyer Senior Resident Inspector - Hope Creek Mr. W. T. Russell, Administrator Region I Ms. J. Moon, Interim Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625
ENCLOSURE 1 PSE&G RESPONSE TO THE NRC GENERIC LETTER 88-11 SALEM UNITS 1 & 2 (1) Results Summary PSE&G has assessed the impact of Regulatory Guide 1.99, Rev.
2 and the attached Regulatory Guide 1.99, Rev. 2, on the pressure temperature limits contained in the Technical Specifications for Salem Units 1 and 2. This review was completed along with the analyses performed to monitor the changes in reactor vessel material fracture toughness based on removal of surveillance capsule Z during Salem Unit l's 7th refueling outage and removal of capsule U during Salem Unit 2's 3rd refueling outage. Results of the technical analysis which was conducted using the methods described in Regulatory Guide 1.99, Rev. 2, were submitted to the NRC previously in Westinghouse reports WCAP 11544 and 11955.
These reports are specific to Salem Units 1 and 2.
The heatup and cooldown curves generated as part of the Westinghouse analyses referenced above, reflect the pressure/temperature (P-T) limits for Salem Units 1 and 2 and are based on the requirements of Appendix G to 10CFR50 and Regulatory Guide 1.99, Rev. 2. The P-T curves for Salem Unit 1 were prepared by Westinghouse in September 1988. The P-T curves for Salem Unit 2 were prepared by Westinghouse in November 1987 using the draft version of Regulatory Guide 1.99, Rev. 2 available at that time. Westinghouse has reviewed the Salem Unit 2 P-T curves and concluded that they are in compliance with the final version of Regulatory Guide
- 1. 99, Rev. 2.
Additionally, the low temperature overpressure protection (LTOP) system for Salem Units 1 and 2 has been reviewed to determine the impact of the revised P-T curves on LTOP setpoint, enable temperature, system hardware, and applicable operating procedures. The results of that review are summarized in the following paragraphs.
At Salem Units 1 and 2, this system is referred to as the Pressurizer Overpressure Protection System (POPS). The POPS is a two train system which uses separate and independent pressure transmitters to open two pressurizer relief valves if RCS pressure exceeds a preset value of 375 psig. The POPS is required to be armed whenever the RCS is below 312°F.
The POPS relief valves protect the RCS from pressure transients exceeding the limits of Appendix G to 10CFR50 when one or more RCS cold leg temperature is at or below 312°F. Either POPS Power Operated Relief Valve (PORV) has adequate relieving capacity to protect the RCS from 1
e (Continued) overpressurization as a result of the limiting heat input or mass input cases; i.e. (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50°F above RCS cold leg temperature or (2) the start of a safety injection pump and its injection into the water solid RCS.
Several provisions presently exist for prevention of pressure transients when the RCS temperature is below 312°F.
Current Technical Specification 3.4.1.3 for startup of an RCP requires that a steam bubble must be established in the pressurizer prior to pump start or the SG/RCS delta-T be verified to be less than 50°F. Also, Technical Specification 3.5.3 allows a maximum of one safety injection pump to remain operable and power to all inoperable safety injection pumps must be removed by racking out the power supply breakers when the RCS temperature is below 312°F.
Also the shutdown procedure requires that a steam bubble be maintained in the pressurizer during plant cooldowns.
The Residual Heat Removal (RHR) System is put into service once the RCS temperature is below 350°F but above 312°F.
The RHR system provides relief capacity comparable to that of a POPS valve. However, no credit has been taken in the low temperature overpressure analysis for this relief capacity.
Based on the above analysis, it is determined that changes to the pressurizer overpressure protection system are not required as a result of the revised P-T limits.
(2) Schedule For Implementation of R.G. 1.99, Rev. 2 As indicated previously, R.G. 1.99, Rev. 2 has been used for the preparation of revised P-T curves and operating limits for Salem Units 1 and 2. PSE&G will submit License Amendment Requests to incorporate new heatup and cooldown curves in the Technical Specifications of Salem Units 1 & 2 by December 1988. As part of those License Amendment requests, appropriate changes to the Bases section will also be provided. Any required modifications to the Salem Updated Final Safety Analysis Report will be submitted as part of the next routine annual update presently scheduled for July 1989.
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ENCLOSURE 2 PSE&G RESPONSE TO NRC GENERIC LETTER 88-11 HOPE CREEK (1) Results Summary The impact of implementing R.G. 1.99 Rev. 2 can best be determined by comparing the RTNDT adjusted reference temperature (ART) values based on R.G. 1.99, Rev. 1 and R.G.
1.99, Rev. 2. Table 1 shows the ART values at 4 EFPY and at 32 EFPY for each beltline material in Hope Creek. These EFPY values are selected to be representative of conditions existing early in the operating life and near the end of the operating life of the plant. The following conclusions are drawn from the results in this table:
(a) The R.G. 1.99, Rev. 2 ART values at 32 EFPY for this unit are below 200°F, which is the allowable limit in 10CFR50, Appendix G. Therefore, implementation of R.G.
1.99, Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing.
(b) The ART value which applies to the pressure-temperature (P-T) curves in the Technical Specifications is 39°F at 32 EFPY as determined using R.G. 1.99, Rev. 1 methods.
The maximum R.G. 1.99, Rev. 2 ART value in Table 1 is 65.8°F at 32 EFPY. Therefore, at 32 EFPY the R.G.
1.99, Rev. 1 P-T curves are less conservative than P-T curves that would be generated with R.G. 1.99, Rev. 2.
However, the current P-T curves are applicable up to 6.7 EFPY even if the ART is calculated according to R.G. 1.99, Rev. 2 methods. Therefore, the P-T curves presently contained in the Technical Specifications are conservative for several more years of operation.
(c) The worst case LPCI nozzle and weld are also included in this beltline region analysis due to their predicted fluence value at 32 EFPY. Since the maximum ART of 13.1°F at 32 EFPY based on Rev. 2, is less than the 40°F RTNDT value applied to the limiting vessel discontinuity curves, the LPCI nozzle is bounded by the limiting vessel discontinuity curves. Therefore, the vessel discontinuity limits on the P-T curves will not be changed through 32 EFPY of operation.
(2) Schedule For Implementation Of R.G. 1.99, Rev. 2 Generic Letter 88-11 requires that R.G. 1.99, Rev. 2 be implemented within two outages after May, 1988. Based on the results of our analyses, the following implementation schedule is proposed:
(a) The current P-T curves are conservative for up to 6.7 EFPY of operation. However, because the 1
e (Continued) implementation of R.G. 1.99, Rev. 2 results in calculated 32 EFPY ART values more conservative than those of R.G. 1.99, Rev. 1 at 32 EFPY, the P-T curves will be revised within two refueling outages (but not later than 6.7 EFPY). Further, since a neutron dosimeter capsule has been removed for testing, revision of the P-T curves will be combined with the results of this testing effort. In this way, implementation of R.G. 1.99, Rev. 2 will be based upon verified, plant specific fluence predictions.
(b) Upon completion of the revisions to the P-T curves, appropriate changes will be made to Section 5.3 and Appendix 5A of the Hope Creek Updated FSAR, and also to Section 3/4.4.6 and the corresponding Bases section of the Hope Creek Technical Specifications. The Technical Specification changes will be submitted prior to completion of the second Refueling outage. Changes to the Updated FSAR will be incorporated as part of the next routine annual update following submittal of the Technical Specification changes.
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Table 1 COMPARISON OF REV 1 AND REV 2 ART VALUES FOR HOPE CREEK 4 EFPY 32 EFPY Rev 1 Rev 2 Rev 1 Rev 2 Beltline Component ART c*r> ART C-F> ART C°F) ART {°F)
Plates:
5K3025-l 26.1 29.8 39.0 64.2 5K2608-l 22.0 24.5 27.5 42.2 5K2698-l 22.8 25.2 29.8 45.0 5K2963-l -4.7 2.5 4.9 29.7 5K2530-l 24.9 33.5 35.6 65.0 5K3238-1 15.2 23.4 30.2 59.3 51(3230-1 -4.l 2.5 6.6 29.7 6C35-l -4.0 5.4 8.9 41. 3 6C45-l 5.7 15.5 14.3 47.0 Beltline Welds:
510-01205 -33.0 -9.2 -20.1 58.0 053040/1125-02205 -24.l 0.1 -13.4 65.8 519-01205 -46.3 -47.1 -41.3 -41.0 504-01205 -28.0 -29.l -22.5 -23.0 055733/1810-02205 -35.4 -27.9 -26~9 10.6 053040/1810-02205 -44.6 -36.9 -36.7 1.6 LPCI Nozzles and Weld:
19468-1 (nozzle) -8.0 7.1 10024-1 (nozzle) -6.1 13.1 001-01205 (weld) -32.4 -31. 5