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Category:Report
MONTHYEARML23235A1722023-08-23023 August 2023
Re. Ginna Nuclear Power Plant, Transmittal of 2023 Owner'S Activity Report
NMP1L3469, Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits2022-06-30030 June 2022
Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits
ML21350A1142021-12-16016 December 2021
Annual Commitment Change Notification
ML21316A0512021-11-12012 November 2021
Notification of Deviation from Pressurized Water Reactor Owners Group (PWROG) Report WCAP-17451-P, Revision 1, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections
RS-21-093, R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2021-09-0101 September 2021
R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections
RS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021
Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping
ML20303A1752020-10-23023 October 2020
Proposed Relief Request from Section XI Repair/Replacement Documentation for Bolting Replacement of Pressure Retaining Bolting
ML20265A1982020-09-21021 September 2020
R. E. Ginna Nuclear Power Plant, Application to Revise Technical Specifications for Steam Generator Tube Inspection Frequency
ML17345A9902017-12-21021 December 2017
R. E. Ginna Nuclear Power Plant Flood Hazard Mitigation Strategies Assessment
ML17214A1182017-07-27027 July 2017
Transmittal of 2017 Owner'S Activity Report for the Plant
CY-16-002, ISFSI - Revision 5 of the Post-Shutdown Decommissioning Activities Report2016-01-14014 January 2016
ISFSI - Revision 5 of the Post-Shutdown Decommissioning Activities Report
ML15167A5052015-06-11011 June 2015
(Redacted Version) Response to Request for Additional Information Regarding the License Amendment Request (LAR) to Adopt NFPA 805, Attachment 3
ML15153A0262015-06-11011 June 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insight
RS-15-069, Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 22015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 2
ML15072A0112015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 2
RS-15-069, Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 2 of 22015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 2 of 2
RS-14-277, Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 12014-09-24024 September 2014
Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1
ML14170B0222014-06-26026 June 2014
Staff Assessment of Flooding Walkdown Reports Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Accident
CY-14-021, Post Shutdown Decommissioning Activities Report (Psdar), Revision 42014-04-0909 April 2014
Post Shutdown Decommissioning Activities Report (Psdar), Revision 4
ML14099A1962014-03-31031 March 2014
Constellation Energy Nuclear Group, LLC - Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task.
ML14071A4782014-02-21021 February 2014
Response to Nrc'S Request for Cashflow Statements Regarding Application for Order Approving Transfer of Operating Authority and Conforming License Amendments
CY-14-004, Independent Spent Fuel Storage Installation Nuclear Liability Insurance Coverage2014-02-19019 February 2014
Independent Spent Fuel Storage Installation Nuclear Liability Insurance Coverage
ML14007A7042014-02-19019 February 2014
R. E. Ginna Nuclear Power Plant - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
ML14037A2772014-02-0909 February 2014
Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for R.E. Ginna Nuclear Power Plant, TAC No.: MF1152
ML13294A0232013-10-16016 October 2013
Snubber Program Plan
ML13210A0342013-07-25025 July 2013
Supplemental Response to 10 CFR 50.54(f) Request for Recommendation 2.3, Seismic Information
ML13093A0652013-03-29029 March 2013
Transition Report to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition - Transition Report, Redacted Version. Enclosure 2
ML13066A1712013-02-28028 February 2013
R.E. Gina, Overall Integrated Plan for Mitigation Strategies for Beyond-Design-Basis External Events
ML12362A4512012-12-21021 December 2012
Supplemental Response to 10 CFR 50.54(f) Request for Information, Recommendation 2.3. Seismic, Attachment (1) Supplemental Seismic Walkdown Report, Table of Contents Through Attachment (2) Seismic Walkdown Checklists, Page B-90
ML12362A4522012-12-21021 December 2012
Supplemental Response to 10 CFR 50.54(f) Request for Information, Recommendation 2.3. Seismic, Attachment (3) Area Walk-By Checklists Through End
ML12277A0902012-10-12012 October 2012
Technical Letter Report on Aging Management Program Audits at Ginna and Nine Mile Point 1
ML12277A1742012-09-28028 September 2012
R. E. Ginna - License Renewal Aging Management, Submit Revised Reactor Vessel Internals Program Document in Accordance with RIS 2011-07
ML12080A1422012-03-16016 March 2012
Attachments 1 & 2, Structural Integrity Evaluation of Circumferential Indication in Ginna Bmi Nozzle No. A86, and Location of Indications in A86 Bmi Penetration
ML11343A6792011-12-0606 December 2011
Report of Facility Changes. Tests, and Experiments Conducted Without Prior Commission Approval
ML11363A0752011-10-0505 October 2011
Reactor Vessel Bottom Mounted Instrumentation Paint Cracking Analysis
ML11363A0762011-08-31031 August 2011
Evaluation of Leakage and Deposit Formation in Painted Full-Scale Bmi Mockups, Final Report, Revision 2, Swri Project No. 18.16196, Ceng Purchase Order No. 6610691
ML11166A1242011-06-0808 June 2011
Response to Second Request (Part 2) for Additional Information for Application for NRC Consent to Indirect License Transfer/Threshold Determination - Merger of Northeast Utilities and Nstar
ML1027306232010-09-22022 September 2010
R. E. Ginna - Transmittal of RCS Pressure and Temperature Limits Report (PTLR)
ML1019606182010-07-15015 July 2010
Exemption from 10 CFR 72.12 and 72.214 for Dry Spent Fuel Storage Activities - Haddam Neck Plant Independent Spent Fuel Storage Installation
ML1012704392010-05-0505 May 2010
Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants
ML1009200572010-03-29029 March 2010
Report of Facility Changes, Tests, and Experiments Conducted Without Prior Commission Approval
ML0915502712009-05-31031 May 2009
WCAP-17036-NP, Rev 0, Analysis of Capsule N from the R.E. Ginna Reactor Vessel Radiation Surveillance Program.
ML0914802612009-05-19019 May 2009
Transmittal of RCS Pressure and Temperature Limits Report (PTLR)
ML0906411032009-02-27027 February 2009
R. E. Ginna Nuclear Power Plant - License Renewal Aging Management Reactor Vessel Internals Program
ML0834501602008-12-31031 December 2008
DOE/RW-0596, Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites, December 2008
ML12220A1012008-11-24024 November 2008
Constellation Energy Group, Inc. Definitive Proxy Statement Schedule 14A
ML0827402682008-09-23023 September 2008
Transmittal of Steam Generator Examination Report for the R.E. Ginna Nuclear Power Plant Conducted in 2008
ML0821900132008-08-0707 August 2008
Monthly Operating Reports Second Quarter 2008
ML0807100412008-02-29029 February 2008
R.E. Ginna, Supplement Response to GL-04-002, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors
ML0731206132007-07-0303 July 2007
Orise Confirmatory Survey Dated 07/03/2007 for Haddam Neck (Cy)
2023-08-23
[Table view]
Category:Technical
MONTHYEARNMP1L3469, Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits2022-06-30030 June 2022
Constellation Energy Company, LLC, Request for Use of Honeywell Mururoa V4F1 R Supplied Air Suits
RS-21-093, R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2021-09-0101 September 2021
R. E. Ginna, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections
RS-21-001, Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping2021-01-0404 January 2021
Revised Proposed Alternative to Utilize Code Cases N-878 and N-880 for Carbon Steel Piping
ML20265A1982020-09-21021 September 2020
R. E. Ginna Nuclear Power Plant, Application to Revise Technical Specifications for Steam Generator Tube Inspection Frequency
RS-15-069, Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 22015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 2
ML15072A0112015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 1 of 2
RS-15-069, Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 2 of 22015-03-11011 March 2015
Enclosure 2: Areva, Inc., Rev. 1 to 51-9205719-001, R.E. Ginna Nuclear Power Plant Flood Hazard Reevaluation Report, Part 2 of 2
RS-14-277, Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 12014-09-24024 September 2014
Proposed Alternative to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1
ML14071A4782014-02-21021 February 2014
Response to Nrc'S Request for Cashflow Statements Regarding Application for Order Approving Transfer of Operating Authority and Conforming License Amendments
ML14007A7042014-02-19019 February 2014
R. E. Ginna Nuclear Power Plant - Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies)
ML14037A2772014-02-0909 February 2014
Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for R.E. Ginna Nuclear Power Plant, TAC No.: MF1152
ML13210A0342013-07-25025 July 2013
Supplemental Response to 10 CFR 50.54(f) Request for Recommendation 2.3, Seismic Information
ML13093A0652013-03-29029 March 2013
Transition Report to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition - Transition Report, Redacted Version. Enclosure 2
ML12277A0902012-10-12012 October 2012
Technical Letter Report on Aging Management Program Audits at Ginna and Nine Mile Point 1
ML12080A1422012-03-16016 March 2012
Attachments 1 & 2, Structural Integrity Evaluation of Circumferential Indication in Ginna Bmi Nozzle No. A86, and Location of Indications in A86 Bmi Penetration
ML11363A0752011-10-0505 October 2011
Reactor Vessel Bottom Mounted Instrumentation Paint Cracking Analysis
ML11363A0762011-08-31031 August 2011
Evaluation of Leakage and Deposit Formation in Painted Full-Scale Bmi Mockups, Final Report, Revision 2, Swri Project No. 18.16196, Ceng Purchase Order No. 6610691
ML0915502712009-05-31031 May 2009
WCAP-17036-NP, Rev 0, Analysis of Capsule N from the R.E. Ginna Reactor Vessel Radiation Surveillance Program.
ML0914802612009-05-19019 May 2009
Transmittal of RCS Pressure and Temperature Limits Report (PTLR)
ML0834501602008-12-31031 December 2008
DOE/RW-0596, Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites, December 2008
ML0821900132008-08-0707 August 2008
Monthly Operating Reports Second Quarter 2008
ML0718704972007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model, Attachment 4 Continued
ML0718705072007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model, Attachment 5 Through End
ML0718704942007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model, Cover Through Attachment 4
ML0718705032007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model. Attachment 4 Continued
ML0718704992007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model, Attachment 4 Continued
CY-07-087, CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model. Attachment 4 Continued2007-05-23023 May 2007
CY-HP-0202, Rev. 0, Calculation of Future Groundwater Dose Resulting from Residual Radioactivity in Concrete to Remain in Groundwater Saturated Zone Using the Cy Ltp Basement Fill Model. Attachment 4 Continued
ML0634004882006-11-15015 November 2006
Final Status Survey Final Report Phase IV, Appendix A9, Survey Unit Release Record 9106-0009, Discharge Canal
ML0634004402006-11-14014 November 2006
Final Status Survey Final Report Phase IV, Appendix A6, Survey Unit Release Record 9106-0006, Discharge Canal
ML0634004992006-11-13013 November 2006
Final Status Survey Final Report Phase IV, Appendix A12, Survey Unit Release Record 9106-0012, Discharge Canal
ML0612101732006-01-12012 January 2006
FEMA Final Exercise Report - R. E. Ginna (Dated 1/12/06)
ML0535602782005-12-31031 December 2005
Task 3 Groundwater Modeling Report, November 2005, Cover Letter Through Figure 22
ML0535602822005-12-31031 December 2005
Task 3 Groundwater Modeling Report, November 2005, Figure 23 Through Table E-1
ML0536201882005-11-17017 November 2005
Meeting with R. E. Ginna Nuclear Power Plant, LLC, Regarding Extended Power Uprate Amendment Application
ML0519300192005-07-0101 July 2005
R. E. Ginna Nuclear Power Plant, Transmittal of RCS Pressure and Temperature Limits Report (PTLR)
CY-05-054, Characterization Report for the West Section of the Excavation Associated with the Northeast Protected Area Grounds2005-02-24024 February 2005
Characterization Report for the West Section of the Excavation Associated with the Northeast Protected Area Grounds
CY-05-047, CY-HP-0193, Rev 0, Connecticut Yankee Decommissioning Health Physics Department Technical Support Document, Assessment of Existing Groundwater Dose for Phase II Release Areas of the Final Status Survey2005-02-22022 February 2005
CY-HP-0193, Rev 0, Connecticut Yankee Decommissioning Health Physics Department Technical Support Document, Assessment of Existing Groundwater Dose for Phase II Release Areas of the Final Status Survey
CY-05-022, License Termination Plan Supplemental Information - Survey Areas Potentially Affected by Groundwater Contamination and Capture Zone Analysis2005-01-31031 January 2005
License Termination Plan Supplemental Information - Survey Areas Potentially Affected by Groundwater Contamination and Capture Zone Analysis
CY-04-131, Estimates for Release of Radionuclides from Potentially Contaminated Concrete at the Haddam Neck Nuclear Plant.2004-12-0101 December 2004
Estimates for Release of Radionuclides from Potentially Contaminated Concrete at the Haddam Neck Nuclear Plant.
ML0435503962004-11-30030 November 2004
Task 2, Supplemental Characterization Report, Volume I - Text, Tables & Figures
ML0435504002004-11-30030 November 2004
Task 2, Supplemental Characterization Report, Volume III, Appendices 6 to End
ML0435503992004-11-30030 November 2004
Task 2, Supplemental Characterization Report, Volume III, Appendices 6, Pages C01-C150
ML0435503972004-11-30030 November 2004
Task 2, Supplemental Characterization Report, Volume II, Appendices 1-3
ML0435503982004-11-30030 November 2004
Task 2, Supplemental Characterization Report, Volume II, Appendices 4 & 5
CY-04-168, Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.20022004-09-16016 September 2004
Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002
ML0426704722004-08-31031 August 2004
Orise 04-1186, Confirmatory Survey of the Administration Building at the Connecticut Yankee Haddam Neck Plant, Haddam, Connecticut.
ML0421702772004-07-27027 July 2004
Revised Final Report - Confirmatory Survey of Open Land Area Survey Units at Connecticut Yankee Haddam Neck Plant, Haddam, Connecticut
ML0411403232004-04-15015 April 2004
R. E. Ginna, Report of Facility Changes, Tests, and Experiments Conducted Without Prior Commission Approval
ML0326811302003-09-25025 September 2003
Summary of the U.S. Nuclear Regulatory Commission (NRC) Staff'S Review of the R.R. Ginna Nuclear Power Plant (Ginna) Steam Generator Tube Inspection Report Dated July 2002
ML0310703292003-04-11011 April 2003
R. E. Ginna Nuclear Power Plant - Response to LRA RAIs 4.2.1-1 & 4.2.2-1
2022-06-30
[Table view] |
Text
t Safety Analysis of Proposed Ilodification t
of'ressurizer Instrument Terminal Blocks Presented to PORC and NSARB on Jan. 31, 1978.
1.0 SCOPE OF 'ANALYSIS This analysis addresses the effects on plant safety of replacing the existing Buchanan terminal blocks used in the. pressurizer pressure and level indication circuits with Nestinghouse Type 542247 (805432) terminal blocks.
The reason for making this change of components is to provide a terminal block which has demonstrated capability to function in the containment environment resulting from a LOCA or MSLB. The existing Buchanan terminal blocks have no qualification documentation. The subject terminal R
F blocks are located in the Containment Vessel basement, ele-k vation 236', outside the pressurizer shield wall. The circuits involved are those associated with PT's-429, 480, 431, and 449, and LT's 428, 426, 427, 433.
2.0 'EFERENCES 2.1 IE Bulletin No. 78-02, January 30, 1978.
2.2 R.E. Ginna FSAR, Appendix 6-E.
2.3 Connecticut Yankee Atomic Power Company.
Letter to L.D. Davison from C.R. Pitman, dated January 28, 1978; Subject,'rankl'in'nstitut'e(FQ Te's't on Westin house Terminal'l'o'ck.
2.4 Connecticut Yankee, Plant Design Change Request No. 270, Technical Review, signed by L.D. Davison
I 2.5 westinghouse Electric Corp., Test Report on the Effect of a LOCA on the'Electrical Performance of Four Terminal Blocks, Report 4 PEW-TR-83, dated September 13, 1977.
2 6 Memo from R. Mattson, Director, Division of Systems Safety, ONRR, USNRC, et al, from R. Tedesco, Assistant Director for Plant Systems, DDS, USNRC.
3 0 SAFETY ANALYSIS 3 1 A review has been made of all events analyzed in the Ginna Station FSAR and the events requiring analysis by USNRC R.G. 1.70. The events related to this modification are:
- 1. Major and minor fires
- 2. Seismic events
- 3. The loss of coolant accident
- 4. The Main Steam Line Break Accident 3.2 The modification does not increase the possibility of or impact of a fire.
3 3 The modification is designated seismic Category I and therefore operability during and following the SSE must, be assured. This is implemented by properly securing the new terminal blocks in a workmanlike fashion. The terminal blocks are lightweight and therefore seismic stresses are negligible.
3 4 Loss of coolant accident. The pressurizer .pressure and level indication circuits are not required to function following a large break. Typically, containment pressure initiates safety injection and no credit is taken for reactor trip.
For smaller breaks, the pressurizer instrumentation may be required for system control. The instrumentation, in any case, is required for trip initiation, within the first half hour of a break. Qualification for this time period is assured as is described in the following paragraphs.
Pressure, temperature, radiation. The following table provides a comparison of test results from Connecticut Yankee (refs. 2.3 and 2.4) and Westinghouse (ref. 2.5) with the Ginna environmental design parameters (ref. 2.2).
C.Y. RG&E Pressure, psig 40 106 to 91 60 Temperature, 'F 285 340 to 329 286 Radiation, rad 2 x 107 2 x 107 rods n reached until 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> The C.Y.'ests were of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration while the N tests were of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> duration. Thus, the tests performed by Westinghouse provide assurance that the instrumentation will survive the pressure, temperature, and radiation levels'for
- a. sufficient length of time to perform their intended function.
Steam environment. References 2.3, 2.4, and 2.5 provide assur-ance that the terminal strips will function for the required length of time in the Ginna steam environment. A review of Reference 2.5 indicates that the terminal strips to be installed, westinghouse Type 542247, have functioned for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 340 F and subsequently functioned under borated spray water.
Chemical spray. The installation at Ginna will be in enclosed cabinets, thereby providing protection against spray. Assurance is provided in the installation that sufficient cabinet vent area exists so that the cabinet will not collapse due to the sudden external pre'ssure.
The cabinets will be thoroughly inspected to assure that there are no openings in the cabinets at a level which would subject the terminals to caustic spray.
Submergence. Xnasmuch as the transmitters and terminal strips are located in the basement, submergence is addressed.
The terminal blocks are located above the lowest transmitter.
The bottom of the lowest transmitter is approximately 23" above the floor level. The volume of the containment sumps is 4050 ft3. Since the total pressurizer volume is 800 ft3 (steam plus water volume) and based upon a review of pressure transient data, in the event of a major loss of coolant accident low level and low pressure trips would be actuated well before submergence would occur. For very small leaks, sump level would provide the operator with sufficient time to take appropriate action prior to submergence. For example, a 70 gpm leak, the maximum flow of the changing pumps, would not fill the sumps for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Steam line break inside containment. Pressurizer pressure and level trips are required for certain trips following a steam line break. Qualification for a period of time sufficient to initiate the trips is assured as described in the following paragraphs.
351 Pressure. .The Ginna FSAR demonstrates that the peak pressure from a steam break is less than the peak pressure from a loss of coolant accident. The discussion in paragraph 3.4.1 assures acceptable performance.
3.5 2 Radiation. Since no fuel failures or release of primary coolant is predicted for a steam break, paragraph 3.4.1 assures acceptability of the terminal strips.
3 5 3 Steam and chemical sprays. The discussion in paragraph 3.4.2 and 3.4.3 is applicable and bounds the steam break t
conditions. Therefore, acceptability of the terminal strips is assured.
3 5 4 Submergence. Calculations have shown that the entire water volume of the secondary side, including hot wells, 'feedwater heaters, piping, air ejector, and steam generators, at normal operation, is 103,315 gallons. Zf one assumes a basement floor free area of 6000 ft2 (a lower bound on the best 1 estimate of 6667 ft2), the water level in the containment basement'is below the lowest transmitter even after the entire secondary side inventory is dumped to containment. Prior to emptying the secondary side, the transmitters will have performed their function.
355 Temperature. A recent NRC memo (ref 2.5) indicates that peak containment temperature following steam line break may exceed, for a brief period of time (about 100 secs.), the LOCA peak temperature. Ref 2.5 reports that this temperature may, for westinghouse PWR's, be as high as 340'F. Paragraph 3.4.1 and
ref 2.5 provide assurance that the transmitter will function for the required period of time.. Additional assurance of acceptable operation is obtained as follows.
The pressurizer instrumentation is located remotely from the steam lines. Thus, temperatures in the vicinity of the pressurizer instrumentation should be lower than those near the steam lines. Heat transfer consideration will cause the temperature of the terminal strips to lag the steam-air. mixture temperature in the immediate vicinity of the terminal block, which in turn should lag the temperature outside the cabinet. Finally, the trip signals may be generated early in the- temperature transient. Thus, there is reasonable assurance that the terminal blocks will provide 1
their necessary function in the post steam break temperature environment.
3.6 Exposure to other hazards. lt" has been confirmed by inspection that there is no chemical or water piping close to the pressurizer instrument enclosures which could result in a spray or deluge. Effects from containment spray are addressed in section 3.4.3.
3.7 Alternative indication. Zf necessary, a standard pressure gauge can be installed on the primary coolant sample line, in the sampling room outside of containment, to display primary pressure directly. The wide range pressurizer level indi-cation circuit (LT 433) is functionally independent of the three pressurizer control system level transmitters (LT 426, LT 427, LT 428), but is located in one of the enclosures
(with PT 431) and utilizes a terminal'lock similar to the others.
Sample lines from both liquid and steam space in the pressurizer are accessible outside containment. Ef necessary these sample lines could be used to monitor pressurizer level and pressure.
4 p PRELIMINARY SAFETY EVALUATION The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, previously evaluated in the safety analysis report will not be increased by the proposed modification.
The possibility of an accident or malfunction of a different tp~e than any evaluated previously in the safety analysis will not be created by the proposed modification.
The margin of safety as fefined in the basis for any tech-nical specification wi13. not be reduced bv the proposed modification.P The proposed modification does not involve an unreviewed safety question or require a Technical Specification change.
