ML032820167
| ML032820167 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/31/2003 |
| From: | Kuo P NRC/NRR/DRIP/RLEP |
| To: | Mecredy R Rochester Gas & Electric Corp |
| Kuo PT, NRR/DRIP/RLEP, 415-1183 | |
| Shared Package | |
| ML032820021 | List: |
| References | |
| Download: ML032820167 (6) | |
Text
D-1 APPENDIX D REFERENCES This appendix contains a listing of references used in the preparation of the Safety Evaluation Report prepared during the review of the license renewal application for Ginna, Docket Number 50-244.
American Society of Mechanical Engineers (ASME)
ASME Boiler and Pressure Vessel Code,Section III, Subsection NF.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.
ASME Boiler and Pressure Vessel Code as modified by Code Case N-481.
CSUPP-ASME(CS)-EXT (Structural carbon steel used in NSSS pipe and component supports that is outdoors, i.e., exposed to the weather).
Bulletins (BL)
NRC Enforcement Bulletin (IEB)79-01B NRC Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors, July 26, 1988.
NRC Bulletin 96-02, Movement of Heavy Loads Over pent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment, dated April 11, 1996.
NRC Bulletin 2002-01, Reactor Pressure Vessel Head Degradation and RCS Pressure Boundary Integrity.
NRC Bulletin 2002-02, Reactor Pressure Vessel Head Penetration Nozzle Inspection Programs.
Code of Federal Regulations ( 10 CFR) 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants."
10 CFR 50.55a, "Codes and Standards".
10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events".
10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients without Scram (ATWS) Events for Light-water-cooled Nuclear Power Plants".
10 CFR 50.63, "Loss of All Alternating Current Power".
D-2 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
10 CFR 50 Appendix G, Fracture Toughness Requirements.
10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions.
10 CFR Part 51, Subpart A, Appendix B, "Environmental Effect of Renewing the Operating License of a Nuclear Power Plant."
10 CFR Parts 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants.
10 CFR 100, "Reactor Site Criteria".
Correspondence U.S. NRC Letter dated August 19, 1985 from John A. Zwolinski to R. W. Kober.
USNRC Letter Johson to Mecredy, Ginna Flaw Indication in the Reactor Vessel Inlet Nozzle Weld - 1989 Reactor Vessel Examination (TAC No. 71906), July 7, 1989.
Electric Power Research Institute (EPRI) and Materials Reliability Program (MRP)
EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants".
EPRI NSAC-20L-R2, Recommendations for an Effective Flow-Accelerated Corrosion Program.
EPRI TR-101108, Boric Acid Corrosion Evaluation Program, Phase 1 - Task 1 Report.
EPRI TR-102134, PWR Secondary Water Chemistry Guideline - Revision 3, May 1993.
EPRI TR-103834-P1-2, Effects of Moisture on the Life of Power Plant Cables, August 1994.
EPRI TR-104213, Bolted Joint Maintenance and Application Guide.
EPRI TR-104748, Boric Acid Corrosion Guidebook.
EPRI TR 107396, Closed Cooling Water Chemistry Guidelines, October 1997.
EPRI 10003057, Table B-3, License Renewal Electrical Handbook, November 2001.
Executive Orders NRC Order EA-03-009, Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors issued on February 11, 2003.
D-3 Generic Letters (GL)
GL 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, March 17, 1988.
GL 89-13, Service Water System Problems Affecting Safety-Related Equipment, July 18, 1989.
GL 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations, April 1, 1997.
Information Notices (IN)
IN 87-44, Thimble Tube Thinning in Westinghouse Reactors.
IN 90-04, Cracking of the Upper Shell-to-Transition Cone Girth Welds in Steam Generators, January 26, 1990.
IN 93-95, Storm-Related Loss of Offsite Power Events Due to Salt Buildup on Switchyard Insulators.
IN 97-46, Unisolable Crack in High-Pressure Injection Piping.
IN 2001-09, Main Feedwater System Degradation in Safety-Related ASME Code Class 2 Piping Inside the Containment of a Pressurized Water Reactor.
Miscellaneous License Renewal Meeting Minutes, Interoffice Correspondence from B. Hunn to G. Wrobel September 23, 2002.
Nuclear Energy Institute (NEI)
NEI 95-10, Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule, Revision 3, issued in March 2001.
NEI 97-06, "Steam Generator Program Guidelines", 1997.
NUREG Reports NUREG-0588 (Category II), Interim Staff Position on Environmental Qualification of Safety-related Electrical Equipment.
NUREG-0612, Control of Heavy Loads, March 2,1983.
NUREG-0737, Clarification of TM Action Plan Requirements.
D-4 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), June 1987.
NUREG-0821, Integrated Plant Safety Assessment Systematic Evaluation Program, R. E.
Ginna Nuclear Power Plant, Final Report, December 1982.
NUREG-1339, Resolution of Generic Safety Issue 29: Bolting Degradation of Failure in Nuclear Power Plants, 1990.
NUREG-1437, Supplement 14, Generic Environmental Impact Statement for License Renewal of Nuclear Plant Regarding R. E. Ginna Nuclear Power Plant, June 25, 2003.
NUREG-1760, Aging Assessment of Safety-Related Fuses Used in Low-and-Medium -Voltage Applications in Nuclear Power Plants," May 2002.
NUREG-1774, A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002," dated July 2003.
NUREG-1800, Standard Review Plan for the Review of License Renewal Application for Nuclear Power Plants," July 2001.
NUREG-1801, Generic Aging Lessons Learned (GALL) Report, July 2001.
NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue on Fatigue Design Curves of Austenitic Stainless Steels, April 1999.
NUREG/CR-5729, Multivariable Modeling of Pressure Vessel and Piping J-R Data.
NUREG/CR-6260, Application of NUREG/CR 5999, Interim Fatigue Curves to Selected Nuclear Power Plant Components.
NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, March 1998.
Regulatory Guides (RG)
RG 1.65, Materials and Inspections for Reactor Vessel Closure Studs, October 1973.
RG-1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Rev. 3 RG 1.99, Revision 2, Radiation Embrittlement of Reactor Pressure Vessel Materials, May 1988.
RG 1.127,Inspection of Water-Control Structures Associated with Nuclear Power Plants,"
March 1978.
RG 1.161, Evaluation of Reactor Pressure Vessels With Charpy Upper-Shelf Energy Less Than 50 ft-lb.
D-5 Reports Framatone Report BAW-2425, Revision 1, Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of R. E. Ginna for Extended Life Through 54 Effective Full Power Years, June 2002.
Readiness of Plant Infrastructure to Support a License Renewal Effort Report, September 19, 2000.
Self Assessment 2002-0044, Ginna Station License Renewal Application, July 19, 2002.
System/Structure Scoping Report, LRSP-AUXFEED, Auxiliary Feedwater (LR-18), April 18, 2002.
System/Structure Scoping Report, LRSP-CCW, Component Cooling Water (LR-06), May 14, 2002.
Containment Building Tendon Investigation, GAI Report 2347 (3.5.2.3.1.2 )
Rochester Gas & Electric (RG&E)
Nuclear Safety Audit Review Board, Minutes of Meeting 245 July 25, 2002 Plant Operations Review Committee, Minutes of Meeting 2002-0043 July 23, 2002 PWR Primary Water Chemistry Guidelines, Revision 3, November, 1995.
Sandia Aging Management Guidelines for Electrical Cable and Terminations.
ND-PRO, "Procedures, Instructions and Guidelines".
Engineering Guideline EG-014, Data Retrieval to Begin License Renewal Project, Revision 0.
Engineering Guideline EG-015, License Renewal Issues Management, Revision 0.
Engineering Guideline EG-017, Ginna Operating Experience Failure Data Retrieval, Revision 0.
Engineering Guideline EG-012, Scoping and Screening and Mechanical AMRs, Revision 1.
Engineering Procedure EP-3-S-0712, License Renewal Project Guideline, Revision 0.
Engineering Procedure EP-3-S-0713, Scoping and Screening for License Renewal, Revision 1.
Engineering Procedure EP-3-S-0714, Mechanical Aging Management Review for License Renewal, Revision 1.
Engineering Procedure EP-3-S-0901, Records and Document Control, Revision 7.
D-6 Engineering Procedure EP-3-S-0715, Electrical Aging Management Review for License Renewal.
Engineering Procedure EP-3-S-0716, Civil Aging Management Review for License Renewal.
Engineering Procedure EP-3-S-0718, Electrical and I&C Integrated Plant Assessment Documents for License Renewal, Revision 0.
Ginna Station License Renewal Audit AINT-2001-0015-CJK November 6, 2001.
IP-PRO-1, Interface Procedures Writers Guide, Revision 8.
IP-CAP-1, Abnormal Condition Tracking Initiation or Notification (Action) Report, Revision 14.
IP-QAP-1, Structure, System, and Component Safety Classifications, Revision 4.
ND-CAP, Corrective Action Program, Revision 7.
ND-DES, Design Control, Revision 6.
Technical Reports ALTRAN Technical Report 99124TR001 (2.3.3.10.2)
Technical Evaluation Report C5506-551, Franklin Research Center, March 29, 1985.
US Nuclear Regulatory Commission (NRC) Inspection Reports Inspection Report 50-244/2003-008 (ML032340358)
Westinghouse Topical Reports (WCAP)
WCAP-7410-L, "Environmental Testing of Engineered Safety Feature Related Equipment (NSSS
- Non-Standard Scope)".
WCAP-7733, "Reactor Vessel Weld Cladding - Base Metal Interaction," July 1971.
WCAP-12928, "Structural Evaluation of the Robert E. Ginna Pressurizer Surge Line, Considering the Effect of Thermal Stratification," May 1991.
WCAP-14422, Rev 2-A, "License Renewal Evaluation: Aging Management for Reactor Coolant Supports," December 2000.
WCAP-14535A, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination," SER published, September 1996.
WCAP-14535A, ""Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination,"
Republication November 1996.
D-7 WCAP-14574-A, "License Renewal Evaluation: Aging Management Evaluation for Pressurizers,"
December 2000.
WCAP-14575-A, "Aging management Evaluation for Class I Piping and Associated Pressure Boundary Components," December 2000.
WCAP-14756-A, "Aging management Evaluation for Pressurized Water Reactor Containment Structure," May 2001.
WCAP-14577, Rev. 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals," March 2001.
WCAP-15338, "A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants," March 2000.
WCAP-15837, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the R. E. Ginna Nuclear Power Plant for the License Renewal Program," April 2002.
WCAP-15873, "A Demonstration of the Applicability of ASME Code Case N-481 to the Primary Loop Casings of R. E. Ginna Nuclear Power Plant for the License Renewal Program," April 2002.
WCAP-15885, "R.E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," Rev 0, May 2002.