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Category:Code Relief or Alternative
MONTHYEARML22194A8402022-07-15015 July 2022 Correction Letter of Relief Request No. HR-I4R-220 Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs ML22181A1252022-07-0505 July 2022 Issuance of Relief Request No. HR I4R-220 Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs ML20091L9992020-04-13013 April 2020 Issuance of Relief Request HC-I4R-190 for the Fourth 10-Year Inservice Inspection Interval ML19220A1022019-09-0606 September 2019 Alternative Request VR-03 to Use ASME Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valve for the Fourth 10-Year Inservice Test Interval ML19136A0262019-07-11011 July 2019 Issuance of Relief Request No. HC-I3R-08, Revision 0, Use of the ASME Code LR-N18-0087, Submittal of Relief Request Associated with the Third Inservice Inspection (ISI) Interval2018-09-24024 September 2018 Submittal of Relief Request Associated with the Third Inservice Inspection (ISI) Interval ML17223A4832017-08-17017 August 2017 Relief from the Requirements of the ASME Code for Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations ML16343A0572016-12-20020 December 2016 Requests for Relief GR-01, PR-01, PR-02, VR-01, and VR-02, for the Fourth Inservice Testing Interval ML15281A1202015-11-0202 November 2015 Relief from the Requirements of the ASME Code ML0824700632008-10-16016 October 2008 Safety Evaluation of Relief Requests for the Third 10-Year Interval of the Inservice Inspection Program for Hope Creek Generating Station ML0515201772005-08-29029 August 2005 Evaluation of Relief Request HC-RR-I2-W01 ML0500302112005-08-0202 August 2005 Relief, HC-RR-12-30, MC5174 ML0420102502004-08-27027 August 2004 Evaluation of Relief Request HC-RR-12-023 ML0322504712003-08-26026 August 2003 Evaluation of Relief Request HC-RR-B08 ML0314204812003-05-15015 May 2003 Supplemental Information for Relief Request HC-RR-B11 ML0311300072003-04-14014 April 2003 Code Relief, Alternative to Required Volumetric Examination for Nozzles Where Plant Configuration Is Such That Visual Examination of Inner Radius May Be Performed on Essentially 100 Percent of Inner Radius ML0311300042003-04-14014 April 2003 Inservice Inspection Program Relief Request HC-RR-B12 ML0306901772003-02-20020 February 2003 Inservice Inspection Program Relief Request HC-RR-F02 ML0305903472003-02-20020 February 2003 Inservice Inspection Program Relief Request HC-RR-B08 ML0305902162003-02-20020 February 2003 Inservice Inspection Program Relief Request HC-RR-A02 ML0305902072003-02-20020 February 2003 Inservice Inspection Program Relief Request HC-RR-A06 2022-07-05
[Table view] Category:Letter
MONTHYEARIR 05000354/20230042024-02-0101 February 2024 Integrated Inspection Report 05000354/2023004 ML24030A8752024-02-0101 February 2024 Operator Licensing Examination Approval ML24009A1022024-01-26026 January 2024 Exemption from Select Requirements of 10 CF Part 73 (EPID L-2023-LLE-0045 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000354/20234012024-01-22022 January 2024 Material Control and Accounting Program Inspection Report 05000354/2023401 ML23341A1372024-01-16016 January 2024 Issuance of Amendment No. 235 Revise Trip and Standby Auto-Start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning ML23335A1122023-12-15015 December 2023 Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers ML23307A1532023-12-15015 December 2023 NRC Investigation Report No. 1-2023-001 ML23270C0072023-11-29029 November 2023 Notice of Proposed Amendment to Decommissioning Trust Agreement ML23324A3072023-11-17017 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000354/20230032023-11-0707 November 2023 Integrated Inspection Report 05000354/2023003 IR 05000272/20234022023-10-12012 October 2023 and Salem Nuclear Generating Station, Units 1 and 2 - Security Baseline Inspection Report 05000354/2023402, 05000272/2023402 and 05000311/2023402 (Cover Letter Only) LR-N23-0065, Submittal of 2023 Annual 10 CFR 50.46 Report2023-10-0202 October 2023 Submittal of 2023 Annual 10 CFR 50.46 Report LR-N23-0045, and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement2023-09-0808 September 2023 and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement ML23249A2612023-09-0606 September 2023 License Amendment Request to Modify the Salem and Hope Creek Exclusion Area Boundary IR 05000354/20230052023-08-31031 August 2023 Updated Inspection Plan for Hope Creek Generating Station (Report 05000354/2023005) ML23192A8212023-08-14014 August 2023 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 234, 347, and 329 Revise Technical Specifications to Delete Meteorological Tower Location IR 05000354/20230022023-08-0303 August 2023 Integrated Inspection Report 05000354/2023002 and Independent Spent Fuel Storage Installation Inspection Report 07200048/2023001 IR 05000354/20230102023-08-0303 August 2023 Biennial Problem Identification and Resolution Inspection Report 05000354/2023010 LR-N23-0052, Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.12023-07-31031 July 2023 Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.1 LR-N23-0042, Spent Fuel Cask Registration2023-07-12012 July 2023 Spent Fuel Cask Registration LR-N23-0046, Emergency Plan Document Revisions Implemented June 28, 20232023-07-10010 July 2023 Emergency Plan Document Revisions Implemented June 28, 2023 IR 05000354/20230112023-05-0101 May 2023 Commercial Grade Dedication Report 05000354/2023011 ML23121A1412023-05-0101 May 2023 Senior Reactor and Reactor Operator Initial License Examinations LR-N23-0034, 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station2023-04-27027 April 2023 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station LR-N23-0035, 2022 Annual Radioactive Effluent Release Report (ARERR)2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report (ARERR) IR 05000354/20230012023-04-26026 April 2023 Integrated Inspection Report 05000354/2023001 LR-N23-0010, License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location2023-04-21021 April 2023 License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location LR-N23-0009, License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains2023-04-18018 April 2023 License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains ML23087A1492023-04-17017 April 2023 NRC to PSEG Salem, Transmittal of the National Marine Fisheries Service'S March 24, 2023, Biological Opinion GAR-2020-02842 Concerning Salem and Hope Creek ML23089A0942023-04-17017 April 2023 NRC to PSEG Hope Creek, Transmittal of the National Marine Fisheries Service'S March 24, 2023, Biological Opinion GAR-2020-02842 Concerning Salem and Hope Creek ML23103A3232023-04-13013 April 2023 Submittal of Updated Final Safety Analysis Report, Rev. 26, Summary of Revised Regulatory Commitments for Hope Creek, Summary of Changes to PSEG Nuclear LLC, Quality Assurance Topical Report, NO-AA-10, Rev. 89 ML23095A3682023-04-12012 April 2023 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Threshold Determination for Proposed Transfer of Land Ownership LR-N23-0024, Submittal of Hope Creek Generating Station Technical Specification Bases Changes2023-03-29029 March 2023 Submittal of Hope Creek Generating Station Technical Specification Bases Changes LR-N23-0006, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-24024 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations ML23086A0912023-03-24024 March 2023 NMFS to NRC, Transmittal of Biological Opinion for Continued Operations of Salem and Hope Creek Nuclear Generating Stations LR-N23-0019, and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums2023-03-21021 March 2023 and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums ML23037A9712023-03-0909 March 2023 and Salem Nuclear, Unit Nos. 1 and 2 Issuance of Amendment Nos. 233, 344, and 325 Relocate Technical Specification Staff Qualification Requirements to the PSEG Quality Assurance Topical Report IR 05000354/20220062023-03-0101 March 2023 Annual Assessment Letter for Hope Creek Generating Station (Report 05000354/2022006) LR-N23-0016, and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments2023-02-28028 February 2023 and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments LR-N23-0018, Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges2023-02-27027 February 2023 Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges LR-N23-0012, Annual Property Insurance Status Report2023-02-24024 February 2023 Annual Property Insurance Status Report LR-N23-0014, Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.7172023-02-23023 February 2023 Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.717 IR 05000354/20220042023-01-24024 January 2023 Integrated Inspection Report 05000354/2022004 LR-N23-0011, In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage2023-01-19019 January 2023 In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage LR-N22-0096, and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination2023-01-0505 January 2023 and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination LR-N22-0094, Emergency Plan Document Revisions Implemented November 21, 20222022-12-14014 December 2022 Emergency Plan Document Revisions Implemented November 21, 2022 LR-N22-0091, Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments2022-12-0202 December 2022 Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments ML22335A0412022-12-0101 December 2022 Notification of Commercial Grade Dedication Inspection (05000354/2023011) and Request for Information IR 05000354/20220032022-11-0303 November 2022 Integrated Inspection Report 05000354/2022003 2024-02-01
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PSEG Nuclear LLC P.O Box 236, Hancocks Bndge, New Jersey 08038-0236 APR 1 4 2003 0 PSEG NuclearLLC LRN-03-0081 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 INSERVICE INSPECTION PROGRAM RELIEF REQUEST HC-RR-B11I HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSES NPF-57 DOCKET NOS. 50-354 Pursuant to 10CFR50.55a(a)(3), PSEG Nuclear, LLC (PSEG Nuclear) requests approval of the enclosed relief request. Approval for relief is requested in accordance with the alternative examination provisions of I OCFR50.55a(a)(3)(i). PSEG Nuclear proposes to use an alternative to the required volumetric examination for nozzles where plant configuration is such that visual examination of the inner radius may be performed on essentially 100 percent of the inner radius in lieu of the existing ASME-Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100. Compliance with the proposed alternatives will provide an adequate level of quality and safety for examination of the affected areas.
The attachment to this letter includes the proposed alternative and supporting justification for the relief. Based on the evaluation contained in the attachment, PSEG Nuclear has concluded that the proposed alternative provides an acceptable level of quality and safety. Accordingly, this proposal satisfies the requirements of 10 CFR 50.55a(a)(3)(i).
This relief request is applicable to PSEG Nuclear Hope Creek Generating Station.
PSEG Nuclear requests that the NRC approve this request by April 2003 in order to support Hope Creek refueling outage RFO 1 scheduled to commence April 12, 2003.
Should you have any questions regarding this request, please contact Mr. Howard Berrick at 856-339-1862.
S3 r elI G.
Salamon Manager - Nuclear Safety and Licensing
Attachment:
go4 ISI Relief Request HC-RR-B11 95-2168 REV 7/99
Document Control Desk APR 1 4 2003 LRN-03-0081 C Mr. H. Miller, Administrator Regional Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. G. Wunder Licensing Project Manager - Hope Creek Mail Stop 08B1 Washington, DC 20555-001 USNRC Senior Resident Inspector - Hope Creek (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box415 Trenton, NJ 08625
Document Control Desk Attachment 1 LRN-03-0081 Relief Request HC-RR-B1 I ASME Code Component Affected Alternative Exam Requirements for Inner Radius Examination of Class 1 Reactor Pressure Vessel Nozzles. [See Table I below]
Applicable ASME Code Edition and Addenda ASME Section Xl, 1989 Edition, is the code of record for PSEG Nuclear LLC (PSEG Nuclear) Hope Creek Nuclear Generating Station's Second Ten-Year ISI Program Interval.
Applicable Code Requirements Conduct ultrasonic examinations of Hope Creek Nuclear Generating Station Reactor Pressure Vessel (RPV) Nozzle Inside Radius Sections in accordance with ASME Section Xi 1989 Edition IWB-2500-1 requirements for Class 1 Examination Category B-D, Item B3.100, Figures IWB-2500-7 (a) through (d).
Proposed Alternative Pursuant to 10 CFR 50.55a(a)(3)(i), approval is requested to use the proposed alternative to the required volumetric examination for nozzles where plant configuration is such that visual examination of the inner radius may be performed on essentially 100 percent of the inner radius in lieu of the existing ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100.
Compliance with the proposed alternatives will provide an adequate level of quality and safety for examination of the affected areas.
PSEG Nuclear proposes to perform an enhanced VT-1, visual examination technique of the surface M-N shown in ASME Section Xl, Figures IWB-2500-7 (a) through (d) as an alternative to ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100 requiring volumetric examination (Ultrasonic, UT) of the Inner Radius of Class 1 Reactor Vessel Nozzles.
The enhanced remote visual examination will be performed upon the examination surface M-N to achieve essentially 100% coverage using 8x magnification video equipment to examine the inner radii. The resolution sensitivity for this remote examination will be established using a 1-mil diameter wire standard similar to that used for other reactor pressure vessel internal examinations intended to detect cracking.
Proposed Alternative InAccordance with 10 CFR 50.55a(a)(3)(i)
-- Alternative Provides Acceptable Level of Quality and Safety -
Page 1 of 4
Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B I Reactor vessel closure head vent and spray nozzles inner radii will receive direct visual examinations (VT-1) conducted in accordance with ASME Xl requirements, while the other remaining aforementioned components will receive enhanced visual examinations using the 1-mil diameter wire standard.
Crack-like surface flaws exceeding the acceptance criteria of Table IWB-3512-1 are unacceptable for continued service unless the reactor vessel meets the requirements of IWB-3142.2, IWB-3142.3 or IWB-3142.4.
Examinations proposed would be performed during the following refueling outages RFO11 (Spring 2003), RFO12 (Fall 2004), and RFO13 (Spring 2005).
Basis for Relief The following Hope Creek RPV Nozzle Inner radius exams listed below do not contain configurations that would impede visual examination of the nozzle inner radius area surface M-N.
Table I Hope Creek RPV Nozzle Inner Radius Exams Summary Number Examination Area Configuration 100408 RPVI-NIAIR - 0°- Recirculation Outlet Nozzle 100409 RPV1-NIBIR 180°- Recirculation Outlet Nozzle 100460 RPV1-N3AIR 720 -Main Steam Nozzle 100465 RPVI-N3BAIR 1080-Main Steam Nozzle 100470 RPV1-N3CAIR 252 0-Main Steam Nozzle 100475 RPV1-N3DAIR 288 0-Main Steam Nozzle 100520 RPVI-N6AIR Spray Head Nozzle 100525 RPV1-N6BIR Spare Spray Head Nozzle 100530 RPVI-N71R Head Vent Nozzle 100330 RPVI-N9A Capped CRD Hydraulic Return Nozzle Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
-- Alternative Provides Acceptable Level of Quality and Safety --
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Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B11I All nozzle forgings were nondestructively examined during fabrication and have been previously examined using ultrasonic techniques specific to the nozzle configuration. No indications of fabrication or service related cracking have been observed as result of these exams.
Nozzle inner radius examinations are the only non-welded areas requiring examination on the RPV. This requirement was deterministically made early in the development of ASME Section Xl, and applied to 100 percent of nozzles welded with full penetration welds. Fatigue cracking is the only applicable degradation mechanism for the nozzle inner radius region. For all nozzles other than feedwater, there is no significant thermal cycling during operation.
Therefore from a risk perspective there is no need to perform volumetric examination on any nozzles other tnan feedwater or operational CRD returns.
No service related cracking has been discovered in any of the BWR (boiling water reactor) fleet plant nozzles other than feedwater and operational CRD returns. The six feedwater nozzle inner radius sections will continue to be -
examined in accordance with UT techniques developed and qualified with GE-NE-523-A71-0594-A Revision 1 (the NRC has approved this report under TAC -
No. MA6787). PSEG Nuclear believes that application of a visual examination alternative for the listed nozzle inner radius regions ensures an acceptable level of quality and safety.
According to the NRC memorandum (Reference No. 1), the staff indicated that an ultrasonic examination could be replaced by a VT-1 visual examination of the proposed nozzle inspections on the basis that the surveillance is being maintained and a VT-1 visual examination is completed.
The implementation of this relief request should reduce vessel examination time by approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, which translates to significantly reduced personnel radiation exposure and cost savings.
Note: For Table IWB-3512-1, the depth of a crack indication is assumed to be one half of the measured length of the crack indication. As previously stated, crack-like surface flaws found exceeding the acceptance criteria of Table IWB-3512-1 are unacceptable for continued service unless the reactor vessel meets the requirements of IWB-3142.2, IWB-3142.3 or IWB-3142.4.
Duration of Proposed Alternative Hope Creek - Second Ten-Year Interval (ASME Xl 1989 Edition)
Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
- Alternative Provides Acceptable Level of Quality and Safety -
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Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B11I Precedence Previous relief has been granted to Detroit Edison, Fermi Unit 2 [NRC Safety Evaluation Reports TAC No. MB2166 and MB2755 dated October 5, 2001].
References
- 1. NRC Internal memorandum from K.R. Wichman (NRC) to W.H. Bateman (NRC) dated May 25, 2000; Subject The Third Meeting with the Industry to discuss the elimination of RPV Inner Radius Inspection (ML003718630).
- 2. Code Case N-648, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1.
Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
-- Alternative Provides Acceptable Level of Quality and Safety -
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