ML052500572
| ML052500572 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/19/2005 |
| From: | Spina J Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP1L 1976, TAC MC3272, TAC MC3273 | |
| Download: ML052500572 (32) | |
Text
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..Constellation Energy:
A Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 13093 August 19,2005 NMP1L 1976 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Nine Mile Point Units 1 and 2 Docket Nos. 50-220 and 50-410 Facility Operating License Nos. DPR-63 and NPF-69 License Renewal Application, Reformatted Sections, 3.1.2., 3.2.2., 3.3.2., 3.4.2.,
3.5.2., 3.6.2.
(TAC Nos. MC3272 and MC3273)
Gentlemen:
By letter NMP 1 L 1962 dated July 14, 2005, Nine Mile Point Nuclear Station, LLC (NMPNS) submitted an amended License Renewal Application (LRA) for Nine Mile Point Units 1 and 2.
During subsequent discussions with the staff it was agreed that the format used by NMPNS in the amended application could be better structured to facilitate the staffs review. NMPNS has reformatted the referenced sections and appended them to this letter as Attachment 1.
There are no new commitments contained in this submittal. If you have any questions about this submittal, please contact David Dellario, NMP License Renewal Project Manager, at (315) 349-7141.
Very truly yours, Jaips A. Spina Vfce President Nine Mile Point JASIMSL/sac
,I(oq
Page 2 NMP1L 1976
Enclosure:
, Reformatted Sections, 3.1.2., 3.2.2., 3.3.2., 3.4.2., 3.5.2., 3.6.2.
cc:
(w/ Enclosure)
Mr. S. Collins, NRC Regional Administrator, Region I Mr. N. B. (Tommy) Le, License Renewal Project Manager, NRR (3 sets) cc:
(w/o Enclosure)
Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. T. G. Colburn, Senior Project Manager, NRR Mr. John P. Spath, NYSERDA
Attachment I Reformatted Sections, 3.1.2., 3.2.2., 3.3.2., 3.4.2., 3.5.2., 3.6.2.
Section 3.1.2
NINE MILE POINT NUCLEAR STATION
,: LICENSE RENEWAL APPLICATION ITECHNICAL INFORMATION 3.1.2.C FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the LRA. For the Reactor Vessel, Internals, and Reactor Coolant Systems, those programs are addressed in the following subsections.
3.1.2.C.1 CUMULATIVE FATIGUE DAMAGE (BWRIPWR)
In accordance with NUREG-1800 paragraph 3.1.2.2.1, fatigue is a time-limited aging analysis (TLAA) as defined in 10CFR54.3. TLAAs are required to be evaluated in accordance with 10 CFR 54.21 (c)(1). The evaluation of this TLAA is addressed separately in Section 4.3.
3.1.2.C.2 LOSS OF MATERIAL DUE TO PITTING AND CREVICE CORROSION (BWR/PWR)
In accordance with NUREG-1800 Section 3.1.2.2.2, (Item 1), this item is applicable only to pitting and crevice corrosion in the PWR steam generator shell assembly. Therefore, this item is not applicable to NMPNS.
In accordance with NUREG-1800 Section 3.1.2.2.2, (Item 2), this item addresses loss of material due to pitting and crevice corrosion in BWR isolation condenser components.
NMP1 has emergency (isolation) condensers (ECs). The design of the emergency condensers features end bells that are welded to the EC shell, which are not designed to be removed. Therefore, eddy current testing of the tubing is not possible. Loss of material is managed by a combination of several programs. The Water Chemistry Control Program, described in Appendix B2.1.2, controls chemical contaminants in both the tube and shell side water such that the conditions that would promote pitting and crevice corrosion are prevented. The EC tube side, which is ASME Class 2, is subject to a system inservice pressure test under the ASME Section Xl, Subsections IWB, IWC, and IWD Inservice Inspection Program (Appendix B2.1.1). The pressure test would detect a tube leak caused by pitting or crevice corrosion. The EC shell is ASME Class 3 and is subject to a functional test under the Inservice Pressure Testing Program under the ASME Section Xl, Subsections IWB, IWC, and IWD Inservice Inspection Program. The functional test would detect loss of material due to pitting and crevice corrosion if the corrosion caused a throughwall leak of the EC shell.
For additional verification that a tube leak does not exist, NMP1 will implement an online tube leakage test. The test will be performed by isolating the makeup and drain valves to the emergency condenser tube side, and monitoring the shell side level for 24 to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to ensure the AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION
," LICENSE RENEWAL APPLICATION.
TECHNICAL INFORMATION water level is not increasing. A water level increase on the shell side during the test would indicate tube leakage. The online test will be incorporated as a new activity in the Preventive Maintenance Program (Appendix B2.1.32).
The new activity will be implemented prior to the period of extended operation.
The Preventive Maintenance Program is also credited for managing loss of material due to pitting and crevice corrosion because it includes the temperature monitoring performed on the Emergency Cooling System, including the heat exchangers. Continuous radiation monitoring of the EC shell side vents is also performed, which would provide indication of a tube leak.
Since none of the activities described above would detect loss of material due to pitting and crevice corrosion before a leak occurred, these activities will be supplemented by a visual inspection for cracking and loss of material of the accessible outer surfaces of the peripheral tubes, tube sheet, and emergency condenser shell. This activity will also be incorporated into the Preventive Maintenance Program.
NMP2 does not have isolation condensers, therefore Item 2 is not applicable.
3.1.2.C.3 LOSS OF FRACTURE TOUGHNESS DUE TO NEUTRON IRRADIATION EMBRITTLEMENT (BWRIPWR)
In accordance with NUREG-1800 Section 3.1.2.2.3, (Item 1), certain aspects of neutron irradiation embrittlement are TLAAs as defined in 10 CFR 54.3. TLAAs are required to be evaluated in accordance with 10 CFR 54.21 (c)(1). The evaluation of this TLAA is addressed separately in Section 4.2.
In accordance with NUREG-1800 Section 3.1.2.2.3, (Item 2), this item addresses loss of fracture toughness due to neutron irradiation embrittlement which could occur in the reactor vessel.
The NMPNS Reactor Vessel Surveillance Program manages loss of fracture toughness of reactor pressure vessel beltline materials for NMP1 and NMP2, and is described in Appendix B2.1.19. The program will be consistent with NUREG-1 801 after enhancements are incorporated. One of the enhancements will be to incorporate the requirements and elements of the Integrated Surveillance Program (ISP) as documented in BWRVIP-116 and approved by the NRC, or an NRC approved plant-specific program, into the Reactor Vessel Surveillance Program. With respect to withdrawal schedules, under the ISP, neither NMP1 nor NMP2 is identified as a host AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION a; LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION plant. The representative materials for the limiting RPV plate and Vweld materials, and their associated withdrawal schedules, are identified in BWRVIP-1 16.
In accordance with NUREG-1800, (Item 3), this item pertains to PWR baffle/former bolts only, therefore is not applicable to NMPNS.
3.1.2.C.4 CRACK INITIATION AND GROWTH DUE TO THERMAL AND MEHANICAL LOADING OR STRESS CORROSION CRACKING (BWR/PWR)
In accordance with NUREG-1800 Section 3.1.2.2.4, (Item 1), this item addresses crack initiation and growth due to thermal and mechanical loading or SCC (including intergranular stress corrosion cracking {IGSCCI) that could occur in small-bore reactor coolant system and connected system piping less than NPS 4.
1 For NMP1 and NMP2, the subject small-bore piping is managed by the ASME Section Xl, Subsection IWB, IWC, and IWD Inservice Inspection Program (Appendix B2.1.1), the Water Chemistry Control Program (Appendix B2.1.2) and the One-Time Inspection Program (Appendix B2.1.20).
Additionally, for small bore piping and fittings in the NMP1 CRD System and NMP2 Reactor Vessel Instrumentation, Reactor Recirculation, and CRD Systems that are not part of the ASME Section Xl, Subsection IWB, IWC, and IWD Inservice Inspection Program, NMPNS only credits the Water Chemistry Control and One-Time Inspection Programs.
For the small bore piping, whether included in the ASME Section Xl, Subsection IWB, IWC, and IWD Inservice Inspection Program or not the inspections conducted under the One-Time Inspection Program will consist of NDE using methods with a demonstrated capability.to detect cracks on the inside surfaces of the piping, or destructive examinations. Both nondestructive and destructive examinations will be performed on a sample of the piping population.
In accordance with NUREG-1800 Section 3.1.2.2.4, (Item 2), this item pertains to crack initiation and growth due to thermal and mechanical loading or SCC (including IGSCC) that could occur in BWR reactor vessel flange leak detection line and BWR jet pump sensing line.
For NMP1 and NMP2, cracking of the vessel flange leak detection lines is managed by the ASME Section Xl, Subsections IWB, IWC, and IWD' Inservice Inspection Program (Appendix B2.1.1), One-Time Inspection Program (Appendix B2.1.20) and Water Chemistry Control Program AGING MANAGEMENT REVIEW
a;,.,
I,' is NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION (Appendix B2.1.2). The inspections conducted under the One-Time Inspection Program will consist of NDE using methods with a demonstrated capability to detect cracks on the inside surfaces of the piping, or destructive examinations. Both nondestructive and destructive examinations will be performed on a sample of the piping population. A portion of the NMP2 vessel flange leak detection line is carbon steel and is not subject to cracking. Loss of material of the carbon steel portion is managed by the Water Chemistry Control Program and the One-Time Inspection Program.
NMP1 does not have a jet pump sensing line and for NMP2, the jet pump sensing lines are not in scope for license renewal. Therefore, the aging effect of cracking is not applicable for jet pump sensing lines at NMPNS.
In accordance with NUREG-1800 Section 3.1.2.2.4, (Item 3), this Item addresses crack initiation and growth due to thermal and mechanical loading or SCC (including IGSCC) that could occur in BWR isolation condenser components.
NMP1 has emergency (isolation) condensers (ECs). The design of the emergency condensers features end bells that are welded to the EC shell, which are not designed to be removed. Therefore, eddy current testing of the tubing is not possible. Cracking is managed by a combination of several programs. The Water Chemistry Control Program, described in Appendix B2.1.2, controls chemical contaminants in both the tube and shell side water such that the conditions that would promote cracking are prevented. The EC tube side, which is ASME Class 2, is subject to a system inservice pressure test under the ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program (Appendix B2.1.1). The pressure test would detect a tube leak caused by cracking. The EC shell is ASME Class 3 and is subject to a functional test under the Inservice Pressure Testing Program under the ASME Section Xl, Subsections IWB, IWC, and IWD Inservice Inspection Program. The functional test would detect cracking due to SCC or cyclic loading if the crack caused a throughwall leak of the EC shell.
For additional verification that a tube leak does not exist, NMP1 will implement an online tube leakage test. The test will be performed by isolating the makeup and drain valves to the emergency condenser tube side, and monitoring the shell side level for 24 to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to ensure the water level is not increasing. A water level increase on the shell side during the test would indicate tube leakage. The online test will be incorporated as a new activity in the Preventive Maintenance Program (Appendix B2.1.32).
The new activity will be implemented prior to the period of extended operation.
The Preventive Maintenance Program is also credited for cracking because it includes the temperature monitoring performed on the Emergency AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWALAPPLICATION TECHNICAL INFORMATION Cooling System, including the heat exchangers. Temperature monitoring can provide early indication of a tube leak. Continuous radiation monitoring of the EC shell side vents is also performed, which would provide indication of a tube leak.
Since none of the activities described above would detect crack initiation or stress corrosion cracking before a leak occurred, these activities will be supplemented by a visual inspection for cracking from the accessible outer surfaces of the peripheral tubes, tube sheet, and emergency condenser shell. This activity will also be incorporated into the Preventive Maintenance Program.
NMP2 does not have isolation condensers, therefore Item 3 is not applicable.
3.1.2.C.5 CRACK GROWTH DUE TO CYCLIC LOADING (PWR)
In accordance with NUREG-1 300 Section 3.1.2.2.5, this item pertains to the reactor vessel shell and reactor coolant system of PWRs only, therefore is not applicable to NMPNS.
3.1.2.C.6 CHANGES IN DIMENSION DUE TO VOID SWELLING (PWR)
In accordance with NUREG-1 800 Section 3.1.2.2.6, this item pertains to PWR reactor vessel internals only, therefore is not applicable to NMPNS.
3.1.2.C.7 CRACK INITIATION AND GROWTH DUE TO STRESS CORROSION CRACKING OR PRIMARY WATER STRESS CORROSION CRACKING (PWR)
In accordance with NUREG-1800 Section 3.1.2.2.7, (Item 1), this item pertains to stress corrosion cracking (SCC) and primary water stress corrosion cracking (PWSCC) of PWR components only, therefore is not applicable to NMPNS.
In accordance with NUREG-1800 Section 3.1.2.2.7, (Item 2), this item pertains to SCC crack initiation and growth in PWR cast austenitic stainless steel (CASS) components only, therefore is not applicable to NMPNS.
In accordance with NUREG-1800 Section 3.1.2.2.7, (Item 3), this item pertains to PWSCC of PWR pressurizer instrument nozzles and heater sheaths and sleeves only, therefore is not applicable to NMPNS.
3.1.2.C.8 CRACK INITIATION AND GROWTH DUE TO STRESS CORROSION CRACKING OR IRRADIATION-ASSISTED STRESS CORROSION CRACKING (PWR)
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION
..LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION In accordance with NUREG-1 800 Section 3.1.2.2.8, this item pertains to PWR reactor internal baffle/former bolts only, therefore is not applicable to NMPNS.
3.1.2.C.9 LOSS OF PRELOAD DUE TO STRES RELAXATION (PWR)
In accordance with NUREG-1800 Section 3.1.2.2.9, this item pertains to PWR reactor internal baffle/former bolts only, therefore is not applicable to NMPNS.
3.1.2.C.10 LOSS OF SECTION THICKNESS DUE TO EROSION (PWR)
In accordance with NUREG-1800 Section 3.1.2.2.10, this item pertains to PWR steam generator components only, therefore is not applicable to NMPNS.
3.1.2.C.1 I CRACK INITIATION AND GROWTH DUE TO PWSCC, ODSCC, OR INTERGRANULAR ATTACK OR LOSS OF MATERIAL DUE TO WASTAGE AND PITTING CORROSION OR LOSS OF SECTION THICKNESS DUE TO FRETTING AND WEAR OR DENTING DUE TO CORROSION OF CARBON STEEL TUBE SUPPORT PLATE (PWR)
In accordance with NUREG-1 800 Section 3.1.2.2.11, this item pertains to PWR steam generator tubes, sleeves and plugs only, therefore is not applicable to NMPNS.
3.1.2.C.12 LOSS OF SECTION THICKNESS DUE TO FLOW-ACCELERATED CORROSION In accordance with NUREG-1800 Section 3.1.2.2.12, this item pertains to PWR steam generator tube supports only, therefore is not applicable to NMPNS.
3.1.2.C.13 LIGAMENT CRACKING DUE TO CORROSION (PWR)
In accordance with NUREG-1800 Section 3.1.2.2.13, this item pertains to PWR steam generator tube support plates only, therefore is not applicable to NMPNS.
3.1.2.C.14 LOSS OF MATERIAL DUE TO FLOW-ACCELERATED CORROSION (PWR)
In accordance with NUREG-1800 Section 3.1.2.2.14, this item pertains to PWR steam generator feedwater inlet rings and supports only, therefore is not applicable to NMPNS.
3.1.2.C.1 5 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section B13.3 for further discussion of this topic.
AGING MANAGEMENT REVIEW
Section 3.2.2
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.2.2.C FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1 801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the license renewal application. For the Engineered Safety Features, those programs are addressed in the following subsections.
3.2.2.C.1 CUMULATIVE FATIGUE DAMAGE In accordance with NUREG-1800, Section 3.2.2.2.1, fatigue is a TLAA as defined in 10 CFR 54.3. TLAAs are required to be evaluated in accordance with 10 CRF 54.21(c)(1). The evaluation of this TLAA is addressed separately in Section 4.3.
3.2.2.C.2 LOSS OF MATERIAL DUE TO GENERAL CORROSION In accordance with NUREG-1 800, Section 3.2.2.2.2, (Item 1), this subsection discusses loss of material, at locations with stagnant flow conditions, due to general corrosion of pumps, valves, piping, and fittings associated with some of the BWR emergency core cooling systems and with lines to the suppression chamber and to the drywell and suppression chamber spray system.
For NMP1, the applicable systems are the Containment Spray, Core Spray, Emergency Cooling, and Main Steam (for automatic depressurization)
Systems. The aging effect is managed by the combination of the Water.
Chemistry Control (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs.
For NMP2, the applicable systems are the High Pressure Core Spray, Low Pressure Core Spray, Reactor Core Isolation Cooling and Residual Heat Removal Systems. The aging effect is managed by the combination of the Water Chemistry Control (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs.
In accordance with NUREG-1800, Section 3.2.2.2.2, (Item 2), this subsection discusses loss of material due to general corrosion of components in the standby gas treatment, containment isolation, and emergency core cooling systems (ECCS).
For NMPI, the applicable systems are the Containment Spray, Core Spray, Emergency Cooling, Reactor Building Ventilation (for standby gas treatment) and Main Steam (for automatic pressurization) Systems. The aging effect for internal surfaces is managed by the One-Time Inspection AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION
. LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION (Appendix B2.1.20), Preventive Maintenance (Appendix B2.1.32) or Open-Cycle Cooling Water (Appendix B2.1.10) Program. The aging effect for external surfaces of carbon steel components in the ECCS systems is managed by the Systems Walkdown Program( Appendix B2.1.33).
For NMP2, the applicable systems are the Hydrogen Recombiner, Reactor Core Isolation Cooling, Standby Gas Treatment, and Main Steam (for automatic depressurization) Systems. The aging effect for internal surfaces is managed by the One-Time Inspection Program. The aging effect for external surfaces of carbon steel components in ECCS systems is managed by the Systems Walkdown Program (Appendix B2.1.33).
3.2.2.C.3 LOCAL LOSS OF MATERIAL DUE TO PITTING AND CREVICE CORROSION In accordance with NUREG-1800, Section 3.2.2.2.3, (Item 1), this subsection discusses loss of material, at locations with stagnant flow conditions, due to pitting and crevice corrosion of pumps, valves, piping, and fittings associated with some of the BWR emergency core cooling systems and with lines to the suppression chamber and to the drywell and suppression chamber spray system.
For NMP1, the applicable systems are the Containment Spray, Core Spray, and Emergency Cooling Systems. The aging effect is managed by the combination of the Water Chemistry Control (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs.
For NMP2, the applicable systems are the High Pressure Core Spray, Low Pressure Core Spray, Reactor Core Isolation Cooling and Residual Heat Removal Systems. The aging effect is managed by the combination of the Water Chemistry Contro! (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs.
In accordance with NUREG-1800, Section 3.2.2.2.3, (Item 2), this subsection discusses loss of material due to pitting and crevice corrosion of components in the standby gas treatment, containment isolation, and emergency core cooling systems.
For NMPI, the applicable systems are the Containment Spray, Core Spray, Emergency Cooling, and Main Steam (for automatic depressurization)
Systems. The aging effect is managed by the One-Time Inspection (Appendix B2.1.20), Preventive Maintenance (Appendix B2.1.32), or Open-Cycle Cooling Water (Appendix B2.1.10) Program.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION For NMP2, the applicable system is the Hydrogen Recombiner System.
The aging effect is managed by the One-Time Inspection (Appendix B2.1.20) Program.
3.2.2.C.4 LOCAL LOSS OF MATERIAL DUE TO MICROBIOLOGICALLY INFLUENCED CORROSION In accordance with NUREG-1800, Section 3.2.2.2.4, this subsection discusses local loss of material due to microbiologically influenced corrosion (MIC) in containment isolation valves and associated piping.
NMPNS considers MIC to be an aging mechanism for systems with raw water as an environment. Neither NMP1 or NMP2 have a raw water environment for containment isolation valves or the associated piping.
Therefore, this issue is not applicable for NMP1 and NMP2.
3.2.2.C.5 CHANGES IN PROPERTIES DUE TO ELASTOMER DEGRADATION In accordance with NUREG-1800, Section 3.2.2.2.5, this subsection discusses change in material properties of seals in the standby gas treatment system.
For NMP1, the Reactor Building Ventilation System provides the equivalent function as a standby gas treatment system. For the internal surfaces of the system's seals (seals are grouped with blowers), the aging effects are managed by the Preventive Maintenance Program Appendix (82.1.32). For the external surfaces, the aging affects are managed by the Systems Walkdown Program (Appendix B2.1.33).
For NMP2, the Standby Gas Treatment System does not contain any seals. Therefore, this subsection is not applicable for NMP2.
3.2.2.C.6 LOCAL LOSS OF MATERIAL DUE TO EROSION In accordance with NUREG-1800, Section 3.2.2.2.6, this issue is applicable only to charging pumps in the Chemical and Volume Control Systems in PWRs. Therefore, it is not applicable to NMPNS.
3.2.2.C.7 BUILDUP OF DEPOSITS DUE TO CORROSION In accordance with NUREG-1 800, Section 3.2.2.2.7, this subsection addresses the plugging of components due to general corrosion in the spray nozzles and flow orifices of the drywell and suppression chamber spray system.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWALAPPLICATION TECHNICAL INFORMATION For NMP1, the Containment Spray System contains the subject spray nozzles and flow orifices. The plugging of spray nozzles due to general corrosion is not an applicable aging effect since these components are stainless steel and not susceptible to general corrosion. The plugging of flow orifices due to general corrosion is not an applicable aging effect since the lines containing these components are completely drained following each system operation in which they are wetted. The draining ensures no corrosion products accumulate in the flow orifices. The flow orifices are located in the Containment Spray heat exchanger drain lines such that, should plugging occur, the intended safety function would not be adversely impacted.
For NMP2, the Containment Spray Cooling mode of the Residual Heat Removal System contains the subject spray nozzles and flow orifices. The plugging of spray nozzles due to general corrosion is not an applicable aging effect since these components are stainless steel and not susceptible to general corrosion. The plugging of flow orifices due to general corrosion is not an applicable aging effect since the lines containing these components are flushed during quarterly testing, which precludes the buildup of deposits.
3.2.2.C.8 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section B13.3 for further discussion of this topic.
AGING MANAGEMENT REVIEW
Section 3.3.2
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.3.2.C FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1 801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the LRA. For the Auxiliary Systems, those programs are addressed in the following subsections.
3.3.2.C.1 LOSS OF MATERIAL DUE TO GENERAL, PITTING, AND CREVICE CORROSION In accordance with NUREG-1800, Section 3.3.2.2.1, Item 1, this subsection discusses the loss of material due to general, pitting, and crevice corrosion for components in the spent fuel pool cooling and cleanup system.
For NMP1 and NMP2, components in the Spent Fuel Pool Cooling Systems are managed by the combination of the Water Chemistry Control (Appendix 62.1.2) and One-Time Inspection (Appendix 62.1.20) Programs.
In accordance with NUREG-1800, Section 3.3.2.2.1, Item 2, this subsection discusses the loss of material due to pitting and crevice corrosion of components in the spent fuel cooling and cleanup system and the shutdown cooling system of older BWRs.
For NMP1, the applicable systems are the Reactor Water Cleanup and Shutdown Cooling Systems. The aging effect is managed by the combination of the Water Chemistry Control and One-Time Inspection Programs.
For NMP2, it is not an older BWR and does not have a shutdown cooling system. Therefore, this item is not applicable.
3.3.2.C.2 HARDENING AND CRACKING OR LOSS OF STRENGTH DUE TO ELASTOMER DEGRADATION OR LOSS OF MATERIAL DUE TO WEAR In accordance with NUREG-1800, Section 3.3.2.2.2, this subsection discusses aging effects that could occur for the elastomer lining of some components exposed to the treated water environment of the spent fuel pool cooling system and elastomer seals and collars in the ductwork of certain ventilation systems exposed to a range of atmospheric conditions.
Elastomers are not used in the lining of spent fuel pool system components within the scope of license renewal at NMPNS.
For NMP1 and NMP2 ventilation systems, the aging effects for seals and collars are managed by the Preventive Maintenance Program (Appendix 62.1.32).
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.3.2.C.3 CUMULATIVE FATIGUE DAMAGE In accordance with NUREG-1800, Section 3.3.2.2.3, fatigue is a TLAA as defined in 10 CFR 54.3. TLAAs are required to be evaluated in accordance with 10 CFR 54.21(c). The evaluation of this TLAA is addressed in Section 4.3.
3.3.2.C.4 CRACK INITIATION AND GROWTH DUE TO CRACKING OR STRESS CORROSION CRACKING In accordance with NUREG-1800, Section 3.3.2.2.4, this subsection discusses cracking due to SCC for the stainless steel Reactor Water Cleanup System regenerative and non-regenerative heat exchangers.
For NMP1, this aging effect for the Reactor Water Cleanup System regenative and non-regenative heat exchangers is managed by the combination of the Water Chemistry Control (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs.
For NMP2, this aging effect is not applicable to the Reactor Water Cleanup System regenative and non-regenative heat exchangers since only the carbon steel shells are within scope and subject to AMR. Cracking is not an applicable aging effect for this material in the system environment.
3.3.2.C.5 LOSS OF MATERIAL DUE TO GENERAL, MICROBIOLOGICALLY INFLUENCED, PITTING, AND CREVICE CORROSION In accordance with NUREG-1800, Section 3.3.2.2.5, this subsection discusses the loss of material from corrosion that could occur on internal and external surfaces of components exposed to a range of atmospheric conditions. Specifically included in the subsection are the ventilation systems, the diesel generator systems' fuel oil, starting air, and combustion air intake and exhaust subsystems, and auxiliary systems' external carbon steel surfaces WSLR.
For NMP1, this aging effect is managed by the Closed-Cycle Cooling Water (Appendix B2.1.11), Fire Water System (Appendix B2.1.17), One-Time Inspection (Appendix B2.1.20), 10 CRF 50 Appendix J (Appendix B2.1.26),
Preventive Maintenance (Appendix B2.1.32), and Systems Walkdown (Appendix B2.1.33) Programs for the applicable systems and components.
For NMP2, this aging effect is managed by the Fire Water System (Appendix B2.1.17), One-Time Inspection (Appendix B2.1.20), Preventive Maintenance (Appendix B2.1.32), Systems Walkdown (Appendix B2.1.33),
and Bolting Integrity (Appendix B2.1.36) Programs for the applicable systems and components AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.3.2.C.6 LOSS OF MATERIAL DUE TO GENERAL, GALVANIC, PITTING, AND CREVICE CORROSION In accordance with NUREG-1800, Section 3.3.2.2.6, this subsection discusses loss of material due to general, galvanic, pitting, and crevice corrosion in the reactor recirculation pumps oil collection system in fire protection. This item is not applicable since NMPNS does not have oil collection systems for its reactor recirculation pumps.
3.3.2.C.7 LOSS OF MATERIAL DUE TO GENERAL, PITTING, CREVICE, AND MICROBIOLOGICALLY INFLUENCED CORROSION AND BIOFOULING In accordance with NUREG-1800, Section 3.3.2.2.7, this subsection discusses loss of material due to general, pitting, crevice, and microbiologically influenced corrosion and biofouling for the internal surfaces of components in the diesel fuel oil system.
For NMP1 and NMP2, this aging effect is managed by the combination of the Fuel Oil Chemistry (Appendix 62.1.18) and One-Time Inspection (Appendix B2.1.20) Programs.
3.3.2.C.8 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section 61.3 of this application for further discussion.
3.3.2.C.9 CRACK INITIATION AND GROWTH DUE TO STRESS CORROSION CRACKING AND CYCLIC LOADING In accordance with NUREG-1800, Section 3.3.2.2.9, this discussion is applicable to PWR systems only and is therefore not applicable to NMPNS.
3.3.2.C.10 REDUCTION OF NEUTRON-ABSORBING CAPACITY AND LOSS OF MATERIAL DUE TO GENERAL CORROSION In accordance with NUREG-1 800, Section 3.3.2.2.10, this subsection discusses reduction of neutron-absorbing capacity and loss of material due to general corrosion in the neutron-absorbing (Boral or boron steel) sheets of the spent fuel storage racks. This item is not applicable since NMPNS identified no aging effects for these components.
3.3.2.C.11 LOSS OF MATERIAL DUE TO GENERAL, PITTING, CREVICE, AND MICROBIOLOGICALLY INFLUENCED CORROSION In accordance with NUREG-1 800, Section 3.3.2.2.11, this subsection discusses the loss of material due to general, pitting, crevice, and microbiologically influenced corrosion for buried piping and fittings.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION For NMPNS, this aging effect is managed by the Buried Piping and Tanks Inspection Program (Appendix B2.1.22) for the NMP1 Diesel Generator Systems and the NMP2 Fire Detection and Protection System.
AGING MANAGEMENT REVIEW
Section 3.4.2
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.4.2.C FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the LRA. For the Steam and Power Conversion Systems, those programs are addressed in the following subsections.
3.4.2.C.1 CUMULATIVE FATIGUE DAMAGE In accordance with NUREG-1800 Section 3.4.2.2.1, fatigue is a TLAA as defined in 10 CFR 54.3. TLAAs are required to be evaluated in accordance with 10 CFR 54.21 (c)(1). The evaluation of this TLAA is addressed separately in Section 4.3.
3.4.2.C.2 LOSS OF MATERIAL DUE TO GENERAL, PITTING, AND CREVICE CORROSION In accordance with NUREG-1800 Section 3.4.2.2.2, this subsection addresses the management of loss of material due to general, pitting, and crevice corrosion for various carbon steel components.
For NMP1 and NMP2, this aging effect is managed by the combination of the Water Chemistry Control (Appendix B2.1.2) and One-Time Inspection (Appendix B2.1.20) Programs for the applicable systems and components.
3.4.2.C.3 LOSS OF MATERIAL DUE TO GENERAL, PITTING, AND CREVICE CORROSION, MICROBIOLOGICALLY INFLUENCED CORROSION AND BIOFOULING In accordance with NUREG-1800, Section 3.4.2.2.3, this discussion is applicable to PWR systems only and is therefore not applicable to NMPNS.
3.4.2.C.4 GENERAL CORROSION In accordance with NUREG-1800 Section 3.4.2.2.4, this subsection applies to loss of material due to general corrosion on the external surfaces of all carbon steel structures and components, including closure bolting, exposed to operating temperatures less than 2120F.
For NMP1 and NMP2, this aging effect is managed by the Systems Walkdown Program (App B2.1.33)
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.4.2.C.5 LOSS OF MATERIAL DUE TO GENERAL, PITTING, CREVICE, AND MICROBIOLOGICALLY INFLUENCED CORROSION In accordance with NUREG-1800, Section 3.4.2.2.5, Item 1, this discussion is applicable to PWR systems only and is therefore not applicable to NMPNS.
In accordance with NUREG-1800, Section 3.4.2.2.5, Item 2, this discussion is applicable to PWR systems only and is therefore not applicable to NMPNS.
3.4.2.C.6 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section B13.3 for further discussion of this topic AGING MANAGEMENT REVIEW
Section 3.5.2
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.5.2.D FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the license renewal application. For the Structures and Component Supports, those programs are addressed in the following subsections.
3.5.2.D.1 PWR AND BWR CONTAINMENTS 3.5.2.D.1.1 AGING OF INACCESSIBLE CONCRETE AREAS In accordance with NUREG-1800, Section 3.5.2.2.1.1, this subsection discusses aging of inaccessible concrete areas in BWR Containments.
For NMP1, this subsection is not applicable since NMP1 is a BWR with a Mark I Containment.
For NMP2, the aging of inaccessible concrete areas due to leaching of calcium hydroxide, aggressive chemical attack, and corrosion of embedded steel are not significant for concrete components of the Primary Containment Structure. The concrete was designed in accordance with ACI 318-71 and ACI 318-77, and constructed in accordance with ACI 301, which meets the intent of ACI 201.2R-77. This ensures a durable concrete that is dense, well-cured, has low permeability, and cracking is well controlled.
Additionally, NMP2 is not exposed to aggressive ground water. As part of the Structures Monitoring Program (Appendix B2.1.28), a regularly scheduled ground water monitoring will be implemented to ensure that a benign environment is maintained.
Although evaluated as being not significant, NMP2 credits the ASME Section Xl Inservice Inspection (IWL) Program (B2.1.24) to monitor for aging of inaccessible concrete areas. Inaccessible concrete areas are compared against accessible concrete areas with similar environments. If warranted, additional inspections are performed.
3.5.2.D.1.2 CRACKING, DISTORTION AND INCREASE IN COMPONENT STRESS LEVEL DUE TO SETTLEMENT; REDUCTIONOF FOUNDATION STRENGHT DUE TO EROSION OF POROUS CONCRETE SUBFOUNDATIONSIF NOT COVERED BY STRUCTURES MONITORING PROGRAM In accordance with NUREG-1800, Section 3.5.2.2.1.2, this subsectionr discusses cracking, distortion, and increase in component stress level due to settlement; and reduction of foundation strength due to erosion of porous concrete subfoundations in BWR Containments.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION For NMP1, this subsection is not applicable since NMP1 is a BWR with a Mark I Containment.
For NMP2, cracking, distortion, and an increase in component stress level due to settlement is not significant. The Primary Containment Structure is founded on impervious rock. Although evaluated as being not significant, NMP2 credits the Structures Monitoring Program (Appendix B2.1.28) to monitor for settlement. NMP2 does not utilize a dewatering system.
For NMP2, reduction of foundation strength due to erosion of porous concrete subfoundation is not applicable. Porous concrete is not utilized in the construction of the Primary Containment Structure.
3.5.2.D.1.3 REDUCTION OF STRENGTH AND MODULUS OF CONCRETE STRUCTURES DUE TO ELEVATED TEMPERATURE In accordance with NUREG-1800, Section 3.5.2.2.1.3, this subsection discusses reduction of strength and modulus of concrete structures due to elevated temperature in BWR Containments.
For NMP1, this subsection is not applicable since NMP1 is a BWR with a Mark I Containment.
For NMP2, reduction of strength and modulus of concrete structures due to elevated temperature is not significant. In the Primary Containment Structure, general area temperatures do not exceed 150 deg F and local area temperatures do not exceed 200 deg F. These temperatures are not sufficient to result in this aging effect for the applicable components.
3.5.2.D.1.4 LOSS OF MATERIAL DUE TO CORROSION IN INACCESSIBLE AREAS OF STEEL CONTAINMENT SHELL OR LINER PLATE In accordance with NUREG-1800, Section 3.5.2.2.1.4, this subsection discusses loss of material due to corrosion in inaccessible areas of steel containment shell or liner plate in BWR Containments.
For NMP1 and NMP2, the ASME Section XI Inservice Inspection (IWE)
Program (Appendix B2.1.23) is credited for managing aging effects due to corrosion of accessible Primary Containment Structure carbon steel components comprising the containment pressure boundary. Inaccessible areas are compared against accessible areas with similar environments. If warranted, additional inspections are performed.
NMP1 also credits the Water Chemistry Control Program (Appendix B2.1.2) and the Torus Corrosion Monitoring Program (Appendix B3.3) to manage AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION aging effects due to corrosion of Primary Containment Structure carbon steel components in demineralized untreated water.
3.5.2.D.1.5 LOSS OF PRESTRESS DUE TO RELAXATION, SHRINKAGE, CREEP AND ELEVATED TEMPERATURE In accordance with NUREG-1800, Section 3.5.2.2.1.5, this subsection discusses loss of prestress due to relaxation, shrinkage, creep and elevated temperature in BWR Containments.
For NMP1, this subsection is not applicable since NMP1 is a BWR with a Mark I Containment.
For NMP2, this subsection is not applicable since prestressed tendons are not utilized in the construction of the Primary Containment Structure.
3.5.2.D.1.6 CUMULATIVE FATIGUE DAMAGE In accordance with NUREG-1800, Section 3.5.2.2.1.6, this subsection discusses cumulative fatigue damage in BWR Containments.
For NMP1 and NMP2, cumulative fatigue damage of Primary Containment Structure steel components is not significant. The maximum operating temperatures of these components is insufficient to cause the magnitude of thermal cycles necessary for fatigue.
See section 4.6 for further discussion of the Torus Shell and Vent System Fatigue Analysis (NMP1 only), Torus Attached Piping Analysis (NMP1 only),
Torus Wall Thickness (NMPI only), Containment Liner Analysis (NMP2 only), and Fatigue of Primary Containment Penetrations.
3.5.2.D.1.7 CRACKING DUE TO CYCLIC LOADING AND SCC In accordance with NUREG-1800, Section 3.5.2.2.1.7, this subsection discusses cracking due to cyclic loading and SCC in BWR Containments.
For NMP1 and NMP2, the ASME Section Xi Inservice Inspection (Subsection IWE) Program (Appendix B2.1.23) and the 10 CFR 50 Appendix J Program (Appendix B2.1.26) are credited for managing cracking due to cyclic loading and SCC of Primary Containment Structure steel components. In addition, an augmented VT-1 visual examination will be performed on containment bellows using enhanced techniques qualified for detecting SCC.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.5.2.D.2 CLASS I STRUCTURES 3.5.2.D.2.1 AGING OF STRUCTURES NOT COVERED BY STRUCTURES MONITORING PROGRAM In accordance with NUREG-1 800, Section 3.5.2.2.2.1, this subsection discusses aging of Class I structures not covered by the Structures Monitoring Program.
There are no Group 6 structures (water control structures) at NMP1 or NMP2.
Aging management of components in accessible areas of Class I structures will be performed through general visual inspections of the Structures Monitoring Program (Appendix B2.1.28). Aging management is performed for the following aging mechenisms: freeze-thaw, leaching of calcium hydroxide, aggressive chemical attack, reaction with aggregates, corrosion of embedded steel, and corrosion of structural steel.
For NMP1 and NMP2, cracking, distortion, and an increase in component stress level due to settlement for Group 1-3, 5, 7-9 structures is not significant. Class I structures are founded on impervious rock. Although evaluated as being not significant, NMP1 and NMP2 credit the Structures Monitoring Program to monitor for settlement. Neither NMP1 nor NMP2 utilizes a dewatering system.
For NMP1, reduction of foundation strength due to erosion of porous concrete subfoundation for Group 1-3, 5, 7-9 structures is not applicable.
Porous concrete is not utilized in the construction of Class I structures.
For NMP2, reduction of foundation strength due to erosion of porous concrete subfoundation is not applicable since the Class 1 structures were designed and analyzed to ACI 318-71 and ACI 318-77. Nonetheless, NMP2 manages the aging of these components with the Structures Monitoring Program (Appendix B2.1.28).
For NMP1 and NMP2, loss of material due to corrosion of structural steel components for Group 1-5, 7-8 structures is managed by the Structures Monitoring Program. Although the NMP1 and NMP2 Vent Stack steel components and NMP2 Reactor cavity Plug Liners are not identified in NUREG-1 801, they are also managed with the Structures Monitoring Program (Appendix 82.1.28). Additionally, NMP1 credits the ASME Section Xi Inservice Inspection (IWE) Program (Appendix B2.1.23) in lieu of the Structures Monitoring Program to manage loss of material due to corrosion of high strength structural fasteners in demineralized untreated water.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION For NMP1 and NMP2, loss of strength and modulus of concrete structures due to elevated temperatures for Group 1-5 structures is not significant. In Class I structures, general area temperatures do not exceed 150 deg F and local area temperatures do not exceed 200 deg F. These temperatures are not sufficient to result in this aging effect for the applicable components.
For NMP1 and NMP2, crack initiation and growth due to SCC and loss of material due to crevice corrosion of stainless steel liners for Group 7 and 8 structures is not applicable. No tank liners were identified as being subject to aging management review.
3.5.2.D.2.2 AGING MANAGEMENT OF INACCESSIBLE AREAS In accordance with NUREG-1800, Section 3.5.2.2.2.2, this subsection discusses aging management of inaccessible areas of Class I structures.
For NMP1 and NMP2, cracking, spalling, and increases in porosity and permeability due to aggressive chemical attack; and cracking, spalling, loss of bond, and loss of material due to corrosion of embedded steel are not significant. Ground water tests confirm that a below-grade aggressive environment does not exist. Although evaluated as being not significant, NMP1 and NMP2 credit the Structures Monitoring Program (Appendix B.1.28) to monitor for aggressive chemical attack and corrosion of embedded steel. A regularly scheduled ground water monitoring will be implemented to ensure that a benign environment is maintained.
3.5.2.D.3 COMPONENT SUPPORTS 3.5.2.D.3.1 AGING OF SUPPORTS NOT COVERED BY STRUCTURES MONITORING PROGRAM In accordance with NUREG-1800, Section 3.5.2.2.3.1, this subsection discusses aging of component supports not covered by the Structures Monitoring Program.
Aging management of component supports will be performed through general visual inspections of the Structures Monitoring Program (Appendix B2.1.28). Aging management is performed for the following aging effect/mechanism combinations: reduction in concrete anchor capacity due to degradation of the surrounding concrete, loss of material due to environmental corrosion, and reduction/loss of isolation function due to degradation of vibration isolation elements.
AGING MANAGEMENT REVIEW
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.5.2.D.3.2 CUMULATIVE FATIGUE DAMAGE DUE TO CYCLIC LOADING In accordance with NUREG-1800, Section 3.5.2.2.3.2, this subsection discusses cumulative fatigue damage due to cyclic loading of component supports.
For NMP1 and NMP2, cumulative fatigue damage of component supports is not significant. The maximum operating temperatures of these components is insufficient to cause the magnitude of thermal cycles necessary for fatigue.
See section 4.6 for further discussion of the Torus Shell and Vent System Fatigue Analysis (NMP1 only), Torus Attached Piping Analysis (NMP1 only),
Torus Wall Thickness (NMPI only), Containment Liner Analysis (NMP2 only), and Fatigue of Primary Containment Penetrations.
3.5.2.D.4 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section B1I.3 for further discussion of this topic.
AGING MANAGEMENT REVIEW
Section 3.6.2
NINE MILE POINT NUCLEAR STATION LICENSE RENEWAL APPLICATION TECHNICAL INFORMATION 3.6.2.2 FURTHER EVALUATION OF AGING MANAGEMENT AS RECOMMENDED BY THE GALL REPORT NUREG-1 801 provides the basis for identifying those programs that warrant further evaluation by the reviewer in the license renewal application. For the Electrical and Instrumentation and Control systems, those programs are addressed in the following subsections.
3.6.2.2.1 ELECTRICAL EQUIPMENT SUBJECT TO ENVIRONMENTAL QUALIFICATION In accordance with NUREG-1800, Section 3.6.2.2.1, Environmental Qualification is a TLAA as defined in 10 CFR 54.3. TLAAs are required to be evaluated in accordance with 10 CRF 54.21(c)(1). The evaluation of this TLAA is addressed separately in Section 4.4.
3.6.2.2.2 QUALITY ASSURANCE FOR AGING MANAGEMENT OF NONSAFETY-RELATED COMPONENTS See Section B13.3 for further discussion of this topic.
AGING MANAGEMENT REVIEW