NL-04-1267, Application for License Renewal - Supplemental Information and Future Action Commitment List
| ML042180163 | |
| Person / Time | |
|---|---|
| Site: | Farley  (NPF-002, NPF-008) |
| Issue date: | 07/27/2004 |
| From: | Stinson L Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-04-1267 | |
| Download: ML042180163 (21) | |
Text
L M. Stinson (Mike)
Southern Nuclear Vice President Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 SO1UTHERN N COMPANY Etergy to Serve Your World' July 27, 2004 Docket Nos.:
50-348 NL-04-1267
.50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Units I and 2 Application for License Renewal - Supplemental Information and Future Action Commitment List Ladies and Gentlemen:
By letter dated September 12, 2003, Southern Nuclear Operating Company (SNC) submitted an application for the renewal of the operating licenses of Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP). In accordance with 10 CFR 54.2 1(b), SNC has reported changes to the FNP current licensing basis (CLB) that materially affect the contents of the license renewal application by letter NL-04-1021 dated July 20, 2004.
There are also other changes at FNP that have occurred in the past year that SNC would like to document to support the NRC's preparation of the safety evaluation report. These have been verbally discussed with the NRC and are provided as supplemental information in Enclosure 1.
An updated License Renewal Future Action Commitment List is provided as Enclosure 2.
This list includes those future actions that SNC has identified through a review of the application and subsequent docketed correspondence between SNC and the NRC. This list has been prepared following the guidance of NEI letter to the NRC dated February 26, 2003 regarding the preparation of the consolidated list of commitments for license renewal.
In addition, miscellaneous responses to supplemental questions and a revised response to Request for Additional Information (RAI) 2.3.3.15-2 are included as Enclosure 3.
(Affirmation and signature are on the following page.)
U99
U. S. Nuclear Regulatory Commission NL-04-1267 Page 2 Mr. L. M. Stinson states he is a vice president of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.
If you have any questions, please contact Charles Pierce at 205-992-7872.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY L. M. Stinson S vorn to and subscribed before me this 27 day of H
a 2004.
Notary Public
- -cMy comnission expires: 6 -1c LMS/MAM/slb
Enclosures:
- 1. Joseph M. Farley Nuclear Plant Units 1 and 2 Application for License Renewal - Supplemental Information
- 2. License Renewal Future Action Commitment List
- 3. Joseph M. Farley Nuclear Plant Units 1 and 2 Application for License Renewal - Miscellaneous Responses cc:
Southern Nuclear Operating Company Mr. J. B. Beasley Jr., Executive Vice President Mr. D. E. Grissette, General Manager - Plant Farley Document Services RTYPE: CFA04.054; LC# 14084 U. S. Nuclear Rezulatorv Commission Ms. T. Y. Liu, License Renewal Project Manager Dr. W. D. Travers, Regional Administrator Mr. S. E. Peters, NRR Project Manager - Farley Mr. C. A. Patterson, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer
NL-04-1 267 ENCLOSURE I Joseph M. Farley Nuclear Plant Units I and 2 Application for License Renewal Supplemental Information El-1
Enclosure I NL-04-1267 Supplemental Information FAC - Resistant Materials Under the flow-accelerated corrosion (FAC) program, FNP has elected to replace some carbon steel piping and piping components with FAC-resistant chrome-molybdenum alloy steels. This is an on-going activity which recently resulted in alloy steel replacements in portions of FAC-susceptible systems that are in-scope for License Renewal. Although the alloy steel has increased resistance to flow-accelerated corrosion, the alloy steel components remain in the scope of the FAC program.
SNC's aging management review for alloy steels in FAC-susceptible systems concludes that the aging effects requiring management are conservatively bounded by those applicable to carbon steel components. The aging management programs applied to the carbon steel components are also applicable to the alloy steel components.
In lieu of adding new component/material combinations, the carbon steel material type in the aging management review summary tables for FAC-susceptible systems is defined to include FAC-resistant alloy steels.
EPRI Water Chemistry Guideline Revisions EPRI has issued 'EPRI PWR Primary Water Chemistry Guidelines Volume 1 and 2, Revision 5, 1002884, September 2003," and Revision 1 to TR-1 07396, 'Closed Cooling Water Chemistry Guideline." Section B.3.2.2 of the Farley LRA references Rev. 4 of EPRI TR-1 05714 of the Primary Water Chemistry Guidelines, and original issue of EPRI TR-1 07396 of the Closed Cooling Water Chemistry Guideline. The applicable FNP procedures and program documents are being revised to incorporate the latest EPRI guidance.
E1-2
NL-04-1 267 ENCLOSURE 2 Joseph M. Farley Nuclear Plant Units I and 2 Application for License Renewal Future Action Commitment List E2-1 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No.
Commitment i LRA App. A Implementation Source No.
ommimentLocation Schedule Suc The Flow Accelerated Corrosion (FAC) program will be enhanced prior to entering the A.2.8 Prior to entering the LRA extended period of operation by adding the auxiliary feedwater pump turbine exhaust period of extended App B, piping to the scope of the program.
operation Section B.4.1 2
The scope of the Diesel Fuel Oil Program and the need to enhance procedural guidance A.2.9 Prior to entering the LRA for maintaining and monitoring the diesel driven fire pump fuel oil system will be period of extended App. B, evaluated prior to the extended period of operation.
operation Section B.4.2 3
The Structural Monitoring Program will be enhanced prior to entering the period of A.2.10 Prior to entering the LRA extended operation to include portions of structures and components which are in scope period of extended App. B, for license renewal but are not currently monitored.
operation Section These additional structures and components include:
B.4.3 submerged portions of the Service Water Intake Structure (SWIS),
in-scope support features for ATWS, SBO, and fire protection safe shutdown equipment in the Turbine Building, structural portions of the Oil Static Pump House, I in-scope components in the Low Level Radwaste Building and Solidification/Dewatering Building (e.g., fire protection).
An enhancement will be made to the Structural Monitoring Program document to clarify the hangers and supports to be inspected in Category I buildings.
E2-2 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No.
Commitment LRA App. A Implementation ourc Location Schedule ioure 4
The scope of the Service Water Program will be expanded prior to the period of A.2.11 Prior to entering the LRA extended operation to include inspection of piping from the main service water header to period of extended App. B, the air compressor credited for Appendix R safe shutdown and inspection of the service operation Section water pump columns.
B.4.4 5
The FNP Fire Protection Program will be enhanced prior to entering the period of A.2.12 Prior to entering the LRA extended operation (with the exception of sprinkler head testing which will be period of extended App. B;-
implemented prior to 50 years of fire protection system service) as follows:
operation Section The fire protection sprinkler system piping will be subjected to wall thickness (with the exception of B.4.5 evaluations (e.g., non-intrusive volumetric testing and/or visual inspections during sprinkler head testing plant maintenance) prior to the period of extended operation and at specific intervals which will be thereafter. The plant-specific inspection interval will be established from the initial implemented prior to Inspection results and revised as appropriate for subsequent inspection results.
50 years of fire protection system A sample of sprinkler heads will be inspected by using the guidance of National Fire service)
Protection Association (NFPA) 25 (2002), Section 5.3.1.1.1, at or before 50 years service and every 10 years thereafter.
Diesel driven fire pump surveillance procedures will be upgraded to provide more detailed instructions related to inspection of the fuel oil supply piping.
The current practice of replacing CO2 hoses at 5 year intervals will be formalized in fire protection procedures.
E2-3 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No Commtment ILRA App. A IImplementation S uc No.
Commitment Location Schedule Source 6
The new FNP Reactor Vessel Internals (RVI) Program will be implemented prior to A.2.1 3 Program App. B, entering the period of extended operation to provide an integrated inspection program implementation: Prior Section that addresses the reactor internals. It will supplement the inspection requirements of to entering the period B.5.1 ASME Section Xi, IWB Category B-N-3 to ensure that aging effects do not result in a of extended loss of intended function of Internal components during the period of extended operation.
operation.
SNC will continue to participate in industry Initiatives intended to clarify the nature and Participation in Letter NL-extent of aging mechanisms potentially affecting the FNP reactor internals. SNC will industry initiatives:
pg E2-13 incorporate the results of these initiatives Into the RVI Program.
Ongoing Activity.
E2-135 FNP will submit an Inspection plan for the RVI Program for NRC review and approval at Submittal of least 24 months prior to entering the periods of extended operation for the FNP units.
inspection plan: At least 24 months prior The RVI Program will be consistent with the NUREG-1 801 Programs XI.M 3 and to entering the period XI.M16, except as described above.
of extended operation for the FNP unit.
7 The new Flux Detector Thimble Inspection Program will be implemented prior to entering A.2.14 Prior to entering the LRA the period of extended operation to formalize examinations already being performed. It period of extended App. B, will be used to identify loss of material due to fretting/wear in the detector thimbles during operation Section the extended period of operation.
B.5.2 The program will include flux detector thimbles for both units. It does not include the instrument guide tubes, which are covered under the ISI Program and the Reactor Vessel Internals Program.
E2-4 NL-04-1267 Farley Nuclear Plant - License Renewal Future Action Commitments No.Com itm ntLRA App. A Implementation S uc No.
Commitment Location Schedule Source 8
The new External Surfaces Monitoring Program will be implemented prior to entering the A.2.1 5 Prior to entering the LRA period of extended operation.
period of extended App. B, Th NP E t rn l S rf c s M o i o in
- operation Section The FNP Extemal Surfaces Monitoring Program will employ periodic visual inspections to B.5.3 manage accessible and insulated external surfaces susceptible to loss of material that require aging management for license renewal. Susceptible external surfaces include Letter NL-carbon steel and low alloy steels in inside and outside environments, and galvanized 04-0924 steel, cast Iron, copper alloys, and aluminum in an outside environment.
pg E2-4; The FNP External Surfaces Monitoring Program Is also credited for managing loss of Letter NL-material, cracking, and change In material properties in elastomers within the scope of 04-0678 the program.
pg E24 9
The new Burled Piping and Tank Inspection Program will be implemented prior to the A.2.16 Prior to entering the LRA period of extended operation. This program will be consistent with NUREG-1 801 period of extended App. B, Program XI.M34, with the exception that it also includes provisions for Inspection of operation Section buried stainless steel and copper alloy piping.
B.5.4 This program will be used to manage loss of material from the external surfaces of In-Letter NL-scope pressure-retaining buried carbon steel piping and tanks, as well as buried 04-0473 stainless steel and copper alloy piping components. Buried piping and tanks within the pg E1-23 scope of the program will be inspected when they are excavated for maintenance or when those components are exposed for any reason.
E2-5 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No.
Commitment I LRA App. A Implementation Source I
I Location I Schedule I
10 The new One-Time Inspection (OTI) Program will be Implemented prior to the period of extended operation. The One-Time Inspection Program will Include measures to verify the effectiveness of various other aging management programs and confirm the absence of aging effects. Insofar as practical with respect to scheduled outages, the inspections will be performed within a window of five years immediately preceding the period of extended operation. This program will be consistent with the aging management programs described in NUREG-1801 XI.M32 and XI.M33.
Specific Components Included in OTI Sample Population:
Pressurizer cast austenitic stainless steel spray heads and associated coupling/
lock bar.
Examination of Reactor coolant system small bore (<4-inch NPS) ASME Class 1 piping components, consistent with NUREG-1 801 Section Xl.M32 requirements, to address NRC concerns regarding cracking due to SCC or thermal cycling. This examination will also serve as an indicator of the potential for SCC in other stainless steel components exposed to a borated water environment.
An RCP thermal barrier CCW nozzle.
Cast iron, bronze, brass and other alloy components in any system requiring aging management that are exposed to environments that may lead to selective leaching.
A bounding CVCS letdown orifice or Charging / Si Pump mini-flow orifice (based on pressure drop).
Sample portion of the external surface of the service water piping in the Diesel Generator Building which is obscured by guard piping.
Alloy steel steam/fluid traps in a steam and treated water environment.
TDAFWP lube oil coolers (fouling of the tubes in a lube oil environment).
U-1 Condensate Storage Tank bottom (thickness measurement).
A.2.17 Inspections to be completed prior to entering the period of extended operation (as noted In the commitment)
LRA App. B.
Section B.5.5 Letter NL-04-0924 pg E2-11 NL 0318 pg E-17 & 19 E2-6 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments Comiten ILRA App. A implementation Suce No.
CommitmentLocation Schedule Sour 10 General LRA Systems In-scoDe:
(cont)
The OTI Program will select and inspect representative locations from the general LRA systems based on the applicable material/ environment/ aging effect combinations (as specified in the LRA and docketed correspondence) to confirm an aging effect does not require management and verify aging management program effectiveness.
11 The plant-specific NiCrFe Component Assessment Program will be implemented prior to A.2.18 Program LRA the period of extended operation. The program scope will include nickel base alloy implementation: Prior App. B, reactor coolant pressure boundary components with known or potential susceptibility to to entering the period Section primary water stress corrosion cracking (PWSCC), excluding steam generator tubes, of extended B.5.8 which are specifically addressed by the Steam Generator Program, and Reactor operation.
Internals which are addressed by the Reactor Vessel Internals Program.
Letter NL-Participation in 04-0715 SNC will continue to participate in industry initiatives (such as the Westinghouse Owners industry initiatives:
pg E1-6 Group and EPRI Materials Reliability Program). Susceptibility rankings and program Ongoing Activity.
Letter NL-inspection requirements will be consistent with the latest version of the EPRI Materials Reliability Program safety assessment regarding Alloy 82/ 182 pipe butt welds or its Submittal of 04-0967 successors.
inspection plan: At pg E-6 least 24 months prior FNP will submit an inspection plan for the NiCrFe Component Assessment Program for to entering the period NRC review and approval at least 24 months prior to entering the periods of extended of extended operation operation for the FNP units.
for the FNP unit.
E2-7 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No.
Commitment LRA App. A Implementation No omtetLocation Schedule Source 12 FNP will implement the new Non-EQ Cables and Connections Program. The Non-EQ A.2.19 Prior to entering the LRA Cables and Connections Program consists of two parts.
period of extended App. B, The first part addresses non-EQ electrical cables and connections used in circuits with operation Section sensitive, high voltage, low-level signals such as radiation monitoring and nuclear
.5.
instrumentation. An AMP designed specifically for these types of cables will be Letter NL-implemented as an alternate program to the XI.E2 program described in NUREG-1801.04-678 All in-scope instrument circuit cables with sensitive, high voltage, low-level signals which pg El-18 are installed in adverse localized environments will be tested.
The other part addresses non-EQ electrical cables and connections exposed to adverse localized environments caused by heat, radiation, or moisture and inaccessible medium voltage cables that are simultaneously exposed to significant moisture and voltage. This program section will be implemented consistent with NUREG-1 801 programs XI.E1 and XI.E3.
E2-8 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments No. lCommitment LRA App. A Implementation Source No.
ommimentLocation Schedule Suc 13 The FNP Fatigue Monitoring Program will be fully implemented consistent with NUREG-A.3.2 Prior to entering the LRA 1801 Program X.M1 prior to the period of extended operation. The Fatigue Monitoring period of extended App. B, program will be used to monitor fatigue conditions of the metal piping and components operation Section that form the reactor coolant pressure boundary (RCPB). Specifically included will be the B.5.7 pressurizer subcomponents, the RPV shell and head, RPV inlet and outlet nozzles, reactor coolant piping, charging nozzles, safety injection nozzles and the other Class 1 piping one-inch in diameter or larger. The other Class I components that have received a fatigue analysis will also be Included, since the cycles they were designed for are bounded by the cycle limits used by the program.
When fully implemented, the program will include:
Cycle Counting - Plant transients that are significant contributors to the fatigue cumulative usage factor will be monitored and counted; Thermal stratification monitoring - susceptible locations (NRC Bulletin 88-08) will be monitored; Stress-based fatigue'monitoring of the surge line and lower region of the pressurizer (includes thermal stratification and insurgeloutsurge effects).
14 The application of the appropriate environmental factors to the calculations for the A.4.2.1 Prior to entering the LRA following locations resulted in an environmentally-assisted fatigue adjusted value greater period of extended Section than 1.0. For the locations listed below, SNC will take corrective action prior to the operation 4.3.1 period of extended operation which might include a more refined analysis, replacement, inspection program approved by the NRC.
Charging nozzles and alternate charging nozzles RHR 6-inch RHR/SI nozzles to the Reactor Coolant System cold leg.
E2-9 NL-04-1 267 Farley Nuclear Plant - License Renewal Future Action Commitments LRA App. A Implementation No.
Commitment Location Schedule Source 15 Pursuant to 10 CFR Part 54.21 (1) (c) (ii), Southern Nuclear will update the RHR Relief A.4.7 Prior to entering the LRA Valve Flow Capacity analysis that utilizes P/T curves as an input to include the period of extended Section calculated 54 EFPY PIT Limit Curves prior to the period of extended operation.
operation 4.5.3 16 Prior to the period of extended operation FNP will collect data for transients (pressurizer N/A Prior to entering the LRA heat-up, small step increase/decrease in load of 10% full power per minute, and large period of extended Section step increase in load) not counted prior to the installation of the fatigue monitoring operation 4.3.1 software and use the data to develop a best estimate historical count and an expected 60-year count.
17 FNP will use the NUREG-1437 Supplemental Environmental Impact Statement for Farley Appendix D Prior to entering the LRA Nuclear Plant along with the original Farley Environmental Impact Statement as the basis period of extended App. D for any environmental reviews performed during the renewal term.
operation 18 Reactor Vessel Surveillance Program:
N/A Prior to operation LRA For each unit, FNP will install alternative dosimetry to monitor neutron fluence on the after removal of the App. B, reactor vessel after removal of the last surveillance capsule in that unit.
last surveillance Section After all surveillance capsules have been removed, the exposure conditions of the capsule in that unit B.3.4 reactor vessel will be monitored to ensure consistency with those used to project the (and effects of embrittlement to the end of the license period. Any plant operating restrictions consist-(on parameters such as inlet temperature, neutron flux, etc.) will be determined. The ency with program will include provisions that if reactor vessel exposure conditions are altered NUREG-such that analysis assumptions could be invalidated, appropriate actions will be 1801 performed (e.g., re-evaluation, re-instituting an active surveillance program, notifying the Section NRC, etc.) assure the adequacy of the projection to the end of the license period.
XI.M31)
E2-10 NL-04-1267 Farley Nuclear Plant - License Renewal Future Action Commitments No eLRA App. A Implementation ce No.
Commitment Location Schedule Sour 19 SNC will submit the following Information on the Unit 1 flux thimble tubes (after the N/A After the second Letter NL-second inspection of the new tube materials during Ul R20):
inspection (during 04-1096, The worst case cumulative wear from the Ul R20 flux thimble tube eddy current wear projection pg E-5 inspection.
analysis is The uncertainty applied to the actual measured wear data.
completed.
The thimble tube wall thickness.
The schedule for the next Unit I flux thimble tube inspection (inspection interval).
The projected wear value for the worst case wear location at the end of the next inspection interval.
- A discussion of the technical basis for establishing the Unit 1 inspection interval that will be implemented after performing the Ul R20 flux thimble tube eddy current inspection of the new tube materials. The discussion will address the use of the equation in Proprietary WCAP-1 2866 and the unit-specific wear data in projecting the wear to the next inspection outage. The curve coefficient (i.e., exponent on")
used in the projection of flux thimble tube wear will be provided.
20 The following Periodic Surveillance and Preventive Maintenance Activities credited for A.2.20 Prior to entering the Letter NL-license renewal are to be implemented prior to the period of extended operation:
(provided in period of extended 04-1 038 Boric Acid Tank Diaphragms Inspections NL-04-1038 operation pg E-8 to Reactor Makeup Water Storage Tank Diaphragms Inspections pg E-8)
E-1 I Condensate Storage Tank Diaphragms Inspections (new LRA Section
_5B 5.9 E2-11
NL-04-1 267 ENCLOSURE 3 Joseph M. Farley Nuclear Plant Units I and 2 Application for License Renewal Miscellaneous Responses NL-04-1 267 RAI 2.3.3.15-2 (Revised Response)
The staff is unable to determine how sight glasses, air distributors and vacuum manometers shown at many locations on the EDG boundary drawings (D-170800L, D-170801L, D-170804L, D-170805L, D-1 70806L, D-1 70807L, D-200209L, D-200211L, D-200212L) are addressed in the LRA. These components are shown as being within the scope of license renewal on the license renewal boundary drawings, however, they are not listed in LRA Table 2.3.3.7 for EDG components subject to an AMR. These components are passive and long-lived, and serve a pressure boundary function. Clarify whether the aforementioned components are included in the component types listed in LRA Table 2.3.3.7. If not, justify the exclusion of these components from being subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
Response
(The revised response clarifies the LRA Table references described below:)
SNC agrees that the sight glasses and air distributors are passive and long-lived, serve a pressure boundary function, and are subject to an aging management review (AMR).
Vacuum manometers are an active indicator and therefore are classified as an active component and not subject to an AMR (nor listed in LRA Table 2.3.3.15).
With respect to the air distributor assemblies, the depiction on the boundary drawing is confusing. SNC models the air distributors as valve bodies, the associated tubing as piping, and the filter casings as filter casings.
With respect to the sight glasses that serve a pressure boundary intended function in the liquid bearing subsystems associated with the diesel generator, SNC should have included a component type "Sight Glasses" in the LRA tables.
The component type "Sight Glasses" should have been included in LRA Table 2.3.3.15 as follows:
Component Type Intended Function Sight Glasses Pressure Boundary SNC has performed an AMR for the sight glasses and determined there is no aging effect requiring management. The components are made of glass, exposed on the interior to jacket water (closed cooling water) or lubricating oil, and on the exterior to an inside environment. There are no aging effects requiring management for these components.
E3-2 NL-04-1 267 The EDG System aging management review summary table 3.3.2-15 should have included the following:
Component Aging Effect NUREG-1801 Type Intended Requiring Aging Management Volume 2 Table I GALL Reference Function Material Environment Management Programs Item Item Notes Sight Glasses Pressure Glass Closed None None Required J
Boundary Cooling Water Lube Oil None None Required J
Inside None None Required J
E3-3 NL-04-1267 Supplemental Question:
Section 4.3.5 of the application identifies a TLAA associated with fatigue of the reactor coolant system supports. No corresponding discussion of the TLAA is included in Appendix A containing the FSAR Supplement.
Response
The wording in the application can be misread as implying that a TLAA exists with regard to the fatigue of Reactor Coolant System (RCS) supports at FNP. As a clarification, SNC determined that no such TLAA exists and therefore, no corresponding section was included in the FSAR Supplement (Appendix A).
Section 4.3.5 is included in the License Renewal Application to address a Renewal Applicant Action Item (RAAI) with regard to fatigue of RCS supports as discussed in the NRC Safety Evaluation ReportforWCAP 14422, Revision 2A. In RAAI 6, the NRC requires the applicant to confirm that the materials of construction for the supports are the same as those evaluated in the WCAP and to confirm that the American Institute of Steel Construction (AISC) Manual of Steel Construction used was not the 1963 version (Sixth Edition). If the AISC 1963 or earlier version of the code was used, then the design accounted for only 10,000 fatigue cycles, which the Safety Evaluation Report (SER) for WCAP 14422 implies is not sufficiently high enough to eliminate the concern for cracking of the supports due to fatigue loads during the period of extended operation.
The materials of construction for FNP's RCS supports are ASTM A36 which is one of the material types evaluated in the WCAP. The code of record for FNP is the AISC 1969 version (Seventh Edition), which incorporates 20,000 fatigue cycles. Westinghouse has indicated in the WCAP that "no fatigue calculations have been performed for the RCS supports as part of their design, since the number of cycles was much less than 20000."
The NRC, in their SER on this WCAP, agrees that, "fatigue is not an applicable aging mechanism because of the low number of cycles or fluctuating loads and the low cumulative usage factor," (see Section 2.4 and 3.3.1.7 of the SER).
Since no fatigue calculations exist for FNP's RCS supports, and there is no applicable aging mechanism to address, there is no TLAA on this issue and therefore no corresponding section in the FSAR supplement (Appendix A of the LRA).
E3-4 NL-04-1267 Supplemental Question to RAI 2.3.3.16-1 The applicant's response to RAI 2.3.3.16-1 stated in part the following:
'In this case, the filters, N1 PI1 F001 and N1PI1F003, are located in Auxiliary Building Rooms 186 and 342 and have a spatial interaction with safety-related SSCs in those rooms."
However, the Farley Inspection Report, dated June 22, 2004, stated in part the following:
"Filters F001, F003 and F004 were included in scope due to 54.4(a)(2). Filter F005 was not included in scope as it is in a line to an abandoned component and is therefore a dry line."
This is the inconsistency between the RAI response and the inspection report.
Therefore, the applicant is requested to add F004 and F005 to its RAI response to resolve this inconsistency.
Response
Boundary Drawing D175047L, Sheet 1 shows several filters that are in the Demineralized Water (DW) system. The response to RAI 2.3.3.16-1 (NL-04-0715, dated April 29, 2004) was provided prior to implementation of the new 10 CFR 54.4(a)(2) scoping methodology. The new methodology replaced the 20 foot spatial separation criterion with the spaces approach. In implementing the new methodology, SNC re-evaluated the location of each of the DW filters to determine if a spatial interaction with vulnerable SR SSCs (of any type) was possible.
Only filters N1PI1F001, NIP1IF003, N1PI1F004, and N1PI1F006 are located such that a spatial relationship exists that would bring the filters into scope for the 10CFR54.4(a)(2) criterion. During the Region II inspection at the plant site, the inspector concentrated on filters N1PI1F001 and NP1I1F003 and verified that N1P11FOO5 was not in scope. The inspection report also refers to N1 P11 F004 as being in the scope of the Rule.
N1PI1F004 is located in the same space as a vulnerable SR SSC (room 186), but it is separated from the SR equipment room 186 by a wall. Given the separation, SNC does not feel that a spray or leak from the filter would have any impact upon a SR function or a SR SSC. However, SNC has conservatively placed NIP11FO04 in the scope of the Rule.
NlPllFO05 is identified in the NRC Region Il Inspection Report as not being in scope; therefore, there is not an inconsistency with the RAI response.
In implementing the revised methodology (subsequent to the inspection and the initial RAI response), SNC determined that N1 P11 F006 is also within the scope of the Rule.
Filter F006 is located in the same room as the Component Cooling Water (CCW) Pumps (which are safety-related). Even though the filter is very far away from the pumps, SNC conservatively includes the filter within the scope of 1 OCFR54.4(a)(2).
E3-5 NL-04-1267 The following table illustrates the scoping of the filters that show up on boundary drawing D1750476L, Sheet 1.
Final Location Scoping Tag#
(Room)
Evaluation Comments In the same room as Spent Fuel Pool NIP11F001 342 In-Scope Cooling pumps.
In the same room as Boric Acid Transfer NlPllFO03 186 In-Scope Pumps Located in the same space as vulnerable SR'SSCs but separated from SR NlPIlFO04 180/186 In-Scope SSCs by a wall.
Normally dry and not in Not In-a room with vulnerable N1P11FOO5 154 Scope SR SSCs In same Room as SR N1PlFO06 185 In-Scope CCW Pumps Not In-These rooms have no NlPllFQ08 177/178 Scope SR SSCs in them.
E3-6