NRC-15-0010, Response to NRC Request for Additional Information for the Review of License Renewal Application - Set 16
| ML15037A531 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/05/2015 |
| From: | Kaminskas V DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-15-0010 | |
| Download: ML15037A531 (57) | |
Text
{{#Wiki_filter:Vito A. Kaminskas Site Vice President DTE Energy Company 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.6515 Fax: 734.586.4172 Email: kaminskasv@dteenergy.com Proprietary Information - Withhold Under 10 CFR 2.390 10 CFR 54 February 5, 2015 NRC-15-0010 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
References:
- 1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
- 2) DTE Electric Company Letter to NRC, "Fermi 2 License Renewal Application," NRC-14-0028, dated April 24, 2014 (ML14121A554)
- 3) NRC Letter, "Requests for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 16 (TAC No.
MF4222)," dated December 23, 2014 (ML14350B365)
Subject:
Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 16 In Reference 2, DTE Electric Company (DTE) submitted the License Renewal Application (LRA) for Fermi 2. In Reference 3, NRC staff requested additional information regarding the Fermi 2 LRA. Enclosures 1, 2, and 5 to this letter provide the response to the request for additional information (RAI). Enclosure 1 provides the responses to all RAIs except for RAI 4.7.4-1 and portions of RAI 4.1-4. provides the response to RAI 4.7.4-1 and Enclosure 5 provides the remaining response to RAI 4.1-4. Enclosures 2 and 5 contain proprietary information as defined by 10 CFR 2.390. General Electric - Hitachi (GEH), as the owner of the proprietary information, has executed the affidavits in Enclosures 4 and 7, which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customarily Enclosures 2 and 5 contain Proprietary Information - Withhold Under 10 CFR 2.390. When separated from Enclosures 2 and 5, this document is decontrolled.
USNRC NRC-15-0010 Page 2 held in confidence, and has been withheld from public disclosure. The proprietary information was provided to DTE in a GEH transmittal that is referenced by the affidavit. The proprietary information has been faithfully reproduced in the enclosed documentation such that the affidavit remains applicable. GEH herein requests as set forth in the enclosed affidavits of Lisa K. Schichlein that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. Non-proprietary versions of the documentation in Enclosures 2 and 5 are provided in Enclosures 3 and 6, respectively. No new commitments are being made in this submittal. However, revisions have been made to commitments previously identified in the LRA. The revised commitments are in Item 7, BWR Vessel Internals, in LRA Table A.4 as indicated in the response to RAI 4.1-2. Should you have any questions or require additional information, please contact Lynne Goodman at 734-586-1205. I declare under penalty of perjury that the foregoing is true and correct. Executed on Feb uary 5, 20? I Vito A. Kaminslias Site Vice President Nuclear Generation
Enclosures:
- 1) DTE Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 16
- 2) Enclosure 1 to GEH Letter 318178-11, "Response to RAI 4.7.4-1" -
PROPRIETARY
- 3) Enclosure 2 to GEH Letter 318178-11, "Response to RAI 4.7.4-1" -
NON-PROPRIETARY
- 4) GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318178-11
- 5) Enclosure 1 to GEH Letter 318178-7, "Response to RAI 4.1-4(b)" -
PROPRIETARY
- 6) Enclosure 2 to GEH Letter 318178-7, "Response to RAI 4.1-4(b)" -
NON-PROPRIETARY
- 7) GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318178-7
USNRC NRC-15-0010 Page 3 cc w/ all
Enclosures:
NRC Project Manager NRC License Renewal Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III cc w/o Enclosures 2 and 5: Michigan Public Service Commission, Regulated Energy Division (kindschl@michigan.gov)
Enclosure 1 to NRC-15-0010 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 DIE Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 16
Enclosure 1 to NRC-15-0010 Page 1 RAI 3.1.2.3.2-2
Background
License Renewal Application (LRA) Section 4.7.3 discusses the applicant'splant-specific time-limited aging analyses (TLAA) for evaluatingloss of preloaddue to irradiation-assisted stress relaxationor creep in the jet pump auxiliary spring wedge assembly. LRA Section 4.7.4 discusses the applicant'splant-specific TLAA that evaluated relaxation of thejet pump slipjoint repair clamps. The applicantdispositionedboth of these TLAAs in accordancewith Title 10 of the Code of FederalRegulations (10 CFR) 54.21 (c) (1)(ii). Issue LRA Table 3.1.2-2 does not include any applicableaging management review (AMR) items for managingloss ofpreloaddue to irradiation-assisted stress relaxationor creep in the jet pump spring wedge assemblies andjet pump slip joint repairclamps that are associatedwith the applicable TLAAs. Request Provide the basis why LRA Table 3.1.2-2 does not include any applicableAMR items to manage loss ofpreloaddue to irradiation-assisted stress relaxationor creep in the jet pump auxiliary spring wedge assembly andjet pump slip joint repairclamps that are associatedwith the applicableplant-specific TLAAs in LRA Sections 4.7.3 and 4.7.4.
Response
The TLAA for loss of preload due to neutron fluence of the jet pump auxiliary spring wedge and slip joint clamp was evaluated. The evaluation considered the projected neutron irradiation through the period of extended operation and determined that the TLAA conclusion remained valid. That conclusion was that the loss of preload due to neutron irradiation was not an aging effect requiring management. Therefore, LRA Table 3.1.2-2, "Reactor Vessel Internals" did not include a line for loss of preload due to neutron irradiation. For clarification, LRA Table 3.1.2-2 will be revised to provide a link to the evaluation of this TLAA. LRA Revisions: LRA Table 3.1.2-2 is revised as shown on the following page. Additions are shown in underline and deletions are shown in strike-through. to NRC-15-0010 Page 2 Table 3.1.-2 Reactor Vessel Internals Summary of Aging Management Evaluation Table 3.1.2-2: Reactor Vessel Internals Aging Effect Aging Intended Requiring Management NUREG- Table 1 Component Type Function Material Environment Management Program 1801 Item Item Notes _ __ __. 3",. Sjaljtcigrepf __ _ v32 ,e _ ig ° 5Los nTI4exTi. .. w__ ... Jet pump assembly Floodable Nickel Treated water Cracking BWR Vessel IV.B1.R- 3.1.1-103 A Adapter ring volume alloy (ext) Internals 100 (lower) Structural support Water Chemistry Control - BWR Jet pump assembly Floodable Nickel Treated water Loss of material Water IV.B1.RP- 3.1.1-43 E, 101
- Adapter ring volume alloy (ext) Chemistry 26 (lower) Structural Control -
support BWR rja *agei d s a semblies °.. .u X fl ence
- 11 cp adustab bol to NRC-15-0010 Page 3 RAI3.1.2.3.2-3
Background
LRA Table 3.1.2-2 states that the jet pump assembly: slip joint clamp adjustable bolt andratchet lock springwill be managedby the BWR [boiling water reactor] Vessel InternalsProgramfor cracking and reduction offracture toughness. The LRA states that the BWR Vessel Internals Program,when enhanced, will be consistent with the program element criteriain "Generic Aging Lessons Learned (GALL) Report" (GALL Report) XiM9, "BWR Vessel Internals." GALL Report aging managementprogram (AMP) XL M9 recommends that the jet pump assembly be managed in accordancewith the recommendedcriteriain BWRVIP [boilingwater reactor vessel and internalsproject] Technical Report No. BWRVIP-41. Issue The staff is unclear if the jet pump assembly: slipjoint clamp adjustable bolt and ratchet lock spring is within the inspection strategy of BWRVIP-41. The staff could not confirm which location in BWRVIP-41, Table 3.3-1, "Matrix of Inspection Options, " recommends specific inspection of these components. Request
- 1) Clarify whether the jet pump assembly: slip joint clamp adjustable bolt and ratchetlock spring components are within the scope of BWRVIP-41 and whether the criteriain Table 3.3-1 of BWRVIP-41 recommends specific inspection of these components. If so, identify the inspection methods andfrequencies that will be appliedto these components.
- 2) If the components are not within the scope of any inspection methods recommended in the BWR VIP-41 report, clarify andprovide the basis on how cracking and reduction offracture toughness will be managedin the components such that the intendedfunction(s) of the components will be maintainedduring the periodof extended operation.
Response
- 1. Specific inspection of slip joint clamp adjustable bolt and ratchet lock spring are not recommended in BWRVIP-41.
- 2. Recommended inspection instructions were provided as part of the engineering design change that installed the hardware. Periodic inspections of the slip joint clamp are scheduled and performed in accordance with plant procedures. This practice is in accordance with BWRVIP-51-A (Jet Pump Repair Design Criteria).
to NRC-15-0010 Page 4 LRA Revisions: None.
Enclosure 1 to NRC-15-0010 Page 5 RAI 3.2.2.3.1-1
Background
For certainAMR items dealing with carbon steel pipingexposed to treatedwater in LRA Tables 3.2.2-1 through 3.2.2-5, DTE indicates that it will use the One-Time Inspection Program to manage loss of material. DTE assignedgeneric note G andplant-specific note 203 to these items. Plant-specificnote 203 explains that the environment may alternatebetween wet and dry for the piping thatpasses through the waterline region of the suppressionpool andstates that the One-Time Inspection Programwill inspect this piping "to manage the potential acceleratedloss of material." LRA Section B.1.33, "One-Time Inspection," includes a table that describes activities to confirm the insignificanceof aging effects and identifies the piping segments that pass through the waterline region of the suppressionpoolfor the correspondingsystems in LRA Tables 3.2.2-1 through 3.2.2-5. For the five entries in the LRA One-Time Inspection Programtable, the activity description states that the one-time inspection "will confirm that loss of material is not occurringor is occurringso slowly that the aging effect will not affect the component intendedfunction." Issue There appears to be a disparitybetween the two statements in the LRA regardingthe purpose of the inspection - either to manage the potentialacceleratedloss of material,or to confirm that loss of materialis not occurringor occurringso slowly that the aging effect will not affect the component intendedfunction. Request Clarmfy the intent of the use of the One-Time Inspection Programfor the portions of carbon steel pipingpassingthrough the waterline region of the suppressionpool. If you determine that loss of material can be accelerated,explain why a one-time inspection is sufficient to manage the effects of aging.
Response
Significant loss of material is not expected due to the control of torus water quality in accordance with the Water Chemistry Control - BWR Program and the normally inert gas environment in the piping splash zone. In addition, this potential aging effect has not been observed at Fermi 2. As noted in License Renewal Application (LRA) Section B.1.33 "One-Time Inspection", any indication or relevant condition of degradation will be evaluated and the need for follow-up examinations will be evaluated based on inspection results. Plant-specific note 203 used in the notes for LRA Tables 3.2.2-1 through 3.2.2-5 will be revised to be consistent with LRA Sections A.1.33 and B..1.33. The environment for items that reference to NRC-15-0010 Page 6 note 203 in LRA Tables 3.2.2-1 through 3.2.2-5 will be revised to include gas in addition to treated water. LRA Revisions: LRA Tables 3.2.2-1 through 3.2.2-5 and associated plant-specific note 203 are revised as shown on the following pages. Additions are shown in underline and deletions are shown in strike-through. to NRC-15-0010 Page 7 Notes for Table 3.2.2-1 through 3.2.2-8-6 Plant-Specific Notes 203. This piping passes through the waterline region of gie suppression pool. The environment for the internal and external surfaces of the piping inthis region may alternate between wet and dry.O ieLrne z r i n yijcorriaos S~ t~f ?Oki Au ) If tnrjjsn; o caccugingu 1 JI 10 2Tifl"AflI S nol 9-_ ord . is cuina so sloly tha the01 aan effect wil not afettecmoetintended _
-rftun Cd=uurii- f< ;rnsed-ogeuor.
,ggrjad 2 - e-rerspe--en37-ees 4semIsr4t4w- 4 t4 4he t9E4e4qe-npe4 -rs-pi4rg;4-manag+4as-pet enda to NRC-15-0010 Page 8 Table 3.2.2-1: Nuclear Pressure Relief System Aging Effect Aging Component Intended Requiring Management NUREG- Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (ext) Ga Inspection 203 Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel int)_/as _ _ _ Inspection 203 Table 3.2.2-2: Residual Heat Removal System Aging Effect Aging Component Intended Requiring Management NUREG- Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Piping Pressure Carbon Treated water Loss of material One-Time -- -- G boundary steel (ext)/Gias Inspection 203 Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (int)IGa Inspection 203 Table 3.2.2-3: Core Spray System Aging Effect Aging Component Intended Requiring Management NUREG- Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (ext) Ga, Inspection 203 Piping Pressure Carbon Treated water Loss of material One-Time -- -- G,
_ boundary steel (int) / Ga s Inspection _ 203 to NRC-15-0010 Page 9 Table 3.2.2-4: High Pressure Coolant Injection System Aging Effect Aging Component Intended Requiring Management NUREG- Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (ext)./sis Inspection 203 Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (int)IG Inspection 203 Table 3.2.2-5: Reactor Core Isolation Cooling System Aging Effect Aging Component Intended Requiring Management NUREG- Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Piping Pressure Carbon Treated water Loss of material One-Time -- -- G boundary steel (ext) Cali Inspection 203 Piping Pressure Carbon Treated water Loss of material One-Time -- -- G, boundary steel (int) / Gs Inspection l 203
Enclosure 1 to NRC-15-0010 Page 10 RAI 3.2.2.2-1
Background
LRA Table 3.2.2-2, "Residual Heat Removal, " includesAMR itemsfor nozzles, which are being managed by the Water Chemistry Control - BWR Program. The LRA shows these components with intendedfunctions that include 'flow control," but only lists "loss of material" as the aging effect requiringmanagement. LRA Table 2.0-1 defines 'flow control" as "provide control of flow rate or establish a pattern of spray." The only pipinglisted in LRA Table 3.2.2-2for this system is carbon steel, which is also being managedfor loss of material. The staff noted thatpiping exposed to Air - Indoor, both internally and externally, is being managed by the External Surfaces MonitoringProgram. The associatedAMR item (3.2.1-44) notes thatfor the components where the internalcarbon steel surfaces are exposed to the same environment as the external surfaces, external surface conditions will be representative of internalsurfaces. DrawingsLRA-M-2083 and LRA-M-2084, "ResidualHeat Removal," apparentlyshow the spray headers in the drywell and above the suppressionpool. The staff noted that the portions of the residual heat removal system associatedwith the spray headers inside the drywell are classified as nonsafety-related. The further evaluation included in NUREG-1800, Revision 2, "StandardReview Planfor Review of License Renewal Applicationsfor Nuclear Power Plants" (SRP-LR), Section 3.2.2.2.5, states thatplugging of the spray nozzles can occur in the spray systems for the drywell and suppression chamber, and that this aging effect will apply even though the system is mostly in a standby mode, because components in the system are occasionally wetted. The staff noted that the surveillance requirementfor Technical Specification 3.6.2.4, "ResidualHeat Removal SuppressionPool Spray," requiresat least 500 gallons per minute flow through the suppression pool spray spargers,which periodicallywets the piping upstream of the spray nozzles. The SRP-LR states that wetting and drying can acceleratecorrosion andfouling, and that the GALL Report recommendsfurther evaluation of a plant-specificAMP to ensure the aging effect is adequately managed. For this issue, the LRA states that the associatedAMR item (3.2.1-6) was not used because the spray nozzles are not steel, but insteadare copper alloy, which is not subject to general corrosion in an indoor air environment. Issue Although the nozzles in LRA Table 3.2.2-2 have a "flow control" intendedfunction, and are intended to "establisha pattern of spray," the LRA only addresses loss of materialfor these components and does not appearto considerflow blockage that could resultfrom corrosion product accumulation in the upstream carbon steel piping. As discussed in SRP-LR Section 3.2.2.2.5, the occasionalwetting and drying of the upstream carbon steel components can accelerate corrosion, which would result in an accumulation of corrosionproducts leading
Enclosure 1 to NRC-15-0010 Page 11 to flow blockage of the spray nozzles due to fouling. Flow blockage is not precluded in the spray nozzles simply because they are constructedof a material that is not susceptible to general corrosion in an indoor air environment. In addition, loss of materialon the external surfaces of the spray headerpiping is being credited for managing loss of materialon the internalsurfaces, basedon the assumption that the environment on the inside of the piping is the same environment as on the outside of the piping. Since the inside of the piping is periodicallywetted andany generated corrosionproducts would tend to accumulate at the bottom of thispiping, it is not clear that the environments inside and outside the suppression chamber spray headerpipingwill be sufficiently similarto justify this assumption. Request
- 1) For the drywell spray nozzles verify that the portionsof the system inside the drywell are nonsafety-related,as shown on drawings LRA-M-2083, andM-2084, such that the intended function of 'flow control" does not need to be consideredas partof the aging management review.
- If "flow control" should be included as partof the agingmanagement reviewfor these components, discuss whetherflow blockage could potentially occur in the spray nozzles.
Include information on whether the carbon steel piping,downstream of isolation valves El150-FO21A & B, has been wetted since plant operationbegan, and whether the environmentfor the interiorof the associatedcarbon steel piping is conducive to an accumulationof corrosionproducts (e.g., low points that do not drain well).
- If 'flow control" is an intendedfunction of the drywell spray nozzles andflow blockage could potentially occur, state how flow blockage due to fouling will be managed during the period of extended operation.
- 2) For the suppression chamber spray nozzles, provide the bases thatperiodic wetting and drying of the upstream carbon steel piping does not result in corrosionproduct accumulation,which couldpotentially causeflow blockage. Include the results of previous technicalspecificationsurveillances to show that operatingexperience supports not managing this aging effect.
- 3) For the AMR items in LRA Table 3.2.2-2, which credit the ExternalSurfaces Monitoring Programto manage internalsurfaces of carbon steel components, provide the bases to show that the periodic wetting of the internalsurfaces and the potential accumulationof corrosion products on the inside bottom of the components does not cause conditions that result in the need to consider the internalenvironment as dfferent from the external environment.
to NRC-15-0010 Page 12
Response
- 1) The drywell spray headers inside the drywell are not safety-related and there is no flow control function.
- 2) The suppression chamber spray nozzles are located on the bottom of the spray header. The spray header can be exposed to a wetting and drying cycle. However, the suppression pool is inerted to less than 4% oxygen during power operation which reduces the potential for corrosion. Surveillance testing is performed every 92 days to verify that flow rates of at least 500 gallons per minute (gpm) are achieved. Flow rates greater than the minimum required for suppression chamber spray were achieved during every surveillance run for the last four years. If significant blockage were to develop, the flow rates through the spray header would indicate a decreasing trend. If the minimum flow rate is not achieved in the future, the cause of the condition, such as flow blockage, would be corrected through the Fermi 2 Corrective Action Program.
The suppression chamber spray header is identified in LRA Table 3.2.2-2 as carbon steel piping with an internal surface exposed to treated water and an external surface exposed to air-indoor. The AMPs credited for this material/internal environment combination are the Water Chemistry Control - BWR Program and the One-Time Inspection Program. The One-Time Inspection Program is intended to confirm the effectiveness of the Water Chemistry Control - BWR Program. The One-Time Inspection "detection of aging effects" in NUREG-1801 describes the sample inspection that will be based on assessment of materials of fabrication, environment, plausible aging effects, and operating experience; and identification of the inspection locations in the system or component based on the potential for the aging effect to occur.
- 3) The piping and valve bodies listed in LRA Table 3.2.2-2, that credit the External Surfaces Monitoring Program to manage the effects of aging for internal surfaces of carbon steel components, are associated with the drywell spray piping and the reactor pressure vessel (RPV) head spray piping. The drywell spray header inside the drywell is normally drained and has not been wetted intentionally or unintentionally. The isolation valves are routinely leak checked and operating procedures ensure that the piping between the two isolation valves is drained. All the drywell spray header low points can be drained. Once every 5 years, air is injected into the header to verify that the nozzles are unobstructed per a surveillance test. The results of these tests have been satisfactory. As shown on draw7ing LRA-M-2083 (location G-3), the RPV head spray line was disconnected as part of a modification; it is blanked and no longer connected to RHR piping such that it does not get wetted. Since the internal environment does not get wetted, there is no periodic wetting and drying of the internal surface of the drywell spray piping and nozzles or the disconnected RPV head spray line.
The carbon steel RHR piping with an internal air environment that is periodically wetted and drained (managed as carbon steel piping with an internal surface exposed to treated water) to NRC-15-0010 Page 13 listed in LRA Table 3.2.2-2 is associated with suppression chamber spray and cooling. The AMPs credited for this material/environment combination are the Water Chemistry Control - BWR Program and the One-Time Inspection Program (i.e. not External Surfaces Monitoring). In addition, the quarterly technical specification surveillance testing described above for the suppression chamber spray header also monitors flow through the suppression pool cooling line which must achieve a flow of at least 9,800 gpm. The minimum required flow rates for suppression chamber cooling were achieved during every surveillance run for the last four years. If the minimum flow rate is not achieved in the future, the cause of the condition would be corrected through the Fermi 2 Corrective Action Program. Therefore, there are no items in LRA Table 3.2.2-2 for which the External Surfaces Monitoring Program is used to manage the aging effects for internal surfaces that are periodically wetted. LRA Revisions: None.
Enclosure I to NRC-15-0010 Page 14 RAI 4.7.4-1
Background
LRA Sections 4.4.7 andA.2.5.4 describe the slipjoint repairclamps as being connected to the diffuser and the mixer (throat)in thejet pump assembly. The LRA states that the clamps were installedwith a preloadthat may decrease due to neutronfluence and thermal exposure. The LRA also states that the analysis that evaluated the decrease of the installationpreloadfor the slipjoint repairclamp is a TLAA that has beenprojected to the end of the periodof extended operationin accordancewith 10 CFR 54.21(c)(1)(ii). The LRA further states that after 52 effective full-power years (EFPY)ofplant operation the expectedfluence at the location of the re/airclamps is 3.07E+18 n/cm 2 (E > 1 Me V), which is below a level necessary (1.0E+19 n/cm) to cause stress relaxation in stainless steel. Issue The stafflacks sufficient information to evaluate thejet pump slipjoint repairclamp TLAA (LRA Sections 4.4.7)for the period of extended operation and determine if the UpdatedFinalSafety Analysis Report (UFSAR) supplement, LRA Section A.2.5.4, adequately summarizes the TLAA in accordance with 10 CFR 54.21(d). The LRA does not include thefollowing information that the staff needs for its determination: (a) the intendedfunction of the jet pump slipjoint repair clamps, (b) how the loss of preloadaffects the capability of the clamps to perform their intended function, (c) a physical description of the slip joint repairclamp, and (d) the specific methodology and details of the methodology that was used to assess loss ofpreload in the jet pump slip joint repairclamps during the periodof extended operation. Request
- 1) State the intendedfunction of the jet pump slipjoint repairclamps and how the loss of preloadaffects the capabilityof the clamps to perform their intendedfunction.
- 2) Provide a physical description or drawing of the slipjoint repairclamps. The level of detail in the descriptionshouldprovidefor an understandingof the style of clamp construction (e.g., bolted,pressed,pined, keyway) and how the clamps are retainedin the jet pump assembly.
- 3) Provide summaries of the stress andfluence analysis, as applicable,used to evaluate thejet pump slip joint repairclampsfor the periodof extended operation. The summaries should include the:
e methodology andpertinentdetails of the methodology used in the analysis e calculations (mathematicalmodeling, includingpertinent safety assumptionsor coefficients used in the modeling) used to evaluate the decrease in preload to NRC-15-0010 Page 15 e key variablesused to evaluate the decrease in preload,such as the design basispreload and minimum value of the preloadrequiredfor the clamps to perform their intended function o basis used to establish thefluence threshold of 1.0E+19 n/cm2 to cause stress relaxation in stainless steel e basis used to determine that the loss ofpreload that has occurredprior to entering the periodof extended operation is acceptable duringthe periodof extended operation Additionally, if the analysis used to evaluate thejet pump slip joint repairclamps utilized any industrialtopical reports or methodologies reviewed and approvedby the NRC, provide the referencesfor these documents and the dates of the staff's safety evaluation (SE) approvingthe reports.
Response
The response contains proprietary information and is provided in Enclosure 2 of this letter. LRA Revisions: None. to NRC-I 5-0010 Page 16 RAI 4.1-1
Background
LRA Table 4.1-2 states that the current licensing basis (CLB) does not include any flow-induced vibration analyses for the Fermi 2 reactor vessel internal (RVI) components that would need to be identified as TLAAs. The LRA states that the flow-induced vibration analyses for the R VI components are not based on time-dependent assumptions defined by the life of the plant and, therefore, they do not conform to the definition of a TLAA in 10 CFR 54.3. Issue UFSAR Section 1.5.2.3 states thatflow-induced vibrations of the RVI components were qualified by prototypical testing performed in accordance with General Electric (GE) Report No. NEDO-24057-P "Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants," dated November 1977, and this report is the design basis for demonstrating conformance with NRC Regulatory Guide (RG) 1.20, "Comprehensive Vibration Assessment Programfor Reactor Internals During Preoperational and Initial Startup Testing." However, the UFSAR does not indicate whether the methodology in GE Report No. NEDO-24057-Pincludes a time-dependent analysis for qualifying the structural integrity of the RVI components against the consequences offlow-induced vibrations. Request Clarfy whether the methodology in GE Report No. NEDO-24057-P includes a time-dependent analysis and whether the analysis is relied upon to qualify the structural integrity of the RVI components against the consequences offlow-induced vibrations. If the analysis is time-dependent, provide justificationas to why it would not need to be identified as a TLAA when compared to the six criteria in 10 CFR 54.3(a).
Response
The methodology in GE Report No. NEDO-24057-P does not include a time-dependent analysis as long as the flow-induced vibration stress is less than 10,000 psi, 0-p. Based on startup vibration measurements at the prototype plant, the maximum peak stress amplitude due to flow induced vibrations is less than 10,000 psi, 0-p. A value of less than 10,000 psi, 0-p implies that no fatigue usage is accumulated by the component due to flow-induced vibration (ASME Section III, Division 1, Appendix I, Figure I-9.2.2, Design Fatigue Curve for Austenitic Steels). In such a case, operating time will have no effect on the RVI component evaluation. LRA Revisions: None.
Enclosure 1 to NRC-15-0010 Page 17 RAI 4.1-2
Background
Section 2.4 of the NRC's December 7, 2000, safety evaluation (ML003775989) on Electric Power Research Institute (EPRI)Technical Report BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines," states that the analysisfor loss of preloaddue to stress relaxationfor the core plate rim holddown bolts is a generic TLAA that was demonstratedto be acceptable in accordancewith 10 CFR 54.21(c)(1)(ii) (i.e., the generic analysis was projected to the end of a postulatedperiodof extended operation). As a result, the staff's safety evaluation includes Applicant Action Item (AAI) No. 4 on the BWR VIP-25 methodology and recommends that BWR applicantsfor license renewal identify and evaluate whether the analysis of stress relaxationin coreplate rim hold down bolts is a TLAA. The applicantprovided its response to AAI No. 4 in LRA Appendix C. The LRA states that the core plate design relies on pre-tensionedrim holddown bolts to maintainposition duringnormal and transientoperationsandpostulateddesign-basis and seismic events. To address AAI No. 4, the LRA also states that the applicantwill enhance the BWR VIP to perform one of thefollowing two options:
- 1) install wedges in the core plate designprior to entering the periodof extended operation
- 2) complete a plant-specific analysis to determine acceptance criteriafor continued inspection of the coreplate rim holddown bolts in accordancewith BWR VIP-25 and submit the inspection plan, along with the acceptance criteriaandjustificationfor the inspectionplan, to the NRC two years priorto enteringthe period of extended operation The applicant included these enhancements in LRA Section A.1.10 and LRA Table A.4, Commitment No. 7.
Issue Commitment No. 7 needs to be clarified,particularlyif the second option is selected as the basis for managing aging of the core plate rim holddown bolts. (a) Option 2 in LRA Table A.4, Commitment No. 7, does not address whether the analysis will evaluate loss ofpreloaddue to stress relaxationin the core plate rim holddown bolts and whether the analysis will quantifj the loss of preload/stressrelaxationthat will occur in these bolts during the periodof extended operation. (b) Presumingthat the analysis in response to Issue (a) of this RAI will be a loss of preload/stressrelaxationanalysis, Option 2 of Commitment No. 7 does not identify whether the analysis will be based on the generic loss ofpreload/stressrelaxation analysis in BWRVIP-25, which was approved in the NRC safety evaluation of December
Enclosure l to NRC-15-0010 Page 18 7, 2000, or a plant-specific loss ofpreload/stressrelaxationanalysis applicableto the Fermi 2 core plate rim holddown bolts. (c) Option 2 of LRA Commitment No. 7 does not require submittal of the applicableanalysis for NRC approval (i.e., if not already approvedby the NRC). Request (a) Clarify whether the specific analysis in Option 2 of LRA Commitment No. 7 will address loss of preloaddue to stress relaxation in the core plate rim holddown bolts, and if so, whether the analysis will quantify the loss ofpreload/stressrelaxationthat will occur in these bolts during the periodof extended operation. If not, justify why the analysis would not quantify the amount of preloadloss/stress relaxationthat would occur in the core plate rim holddown bolts at the end of the periodof extended operation. (b) Clarify whether the analysis referredto in this commitment will be a plant-specific loss of preload/stressrelaxationanalysisfor the core plate rim holddown bolts or the generic analysis loss of preload/stressrelaxationanalysis that was evaluated in BWR VIP-25 and approvedin the NRC SE of December 7, 2000. If the analysis will be the generic analysis in BWR VIP-25, provide your basis why the analysis has not been identifiedas a TLAA for the LRA and evaluated (withjustification)in accordancewith one of the TLAA acceptance requirementsin 10 CFR 54.21(c)(1) (1), (ii), or (iii), andjustify why the generic coreplate rim holddown analysis is consideredto be bounding and acceptablefor the design and loadings of the core plate assembly at Fermi 2. (c) Explain why Option 2 of Commitment No. 7 does not require the loss of preload/stress relaxationanalysis to be submittedfor NRC approval (i.e., if the analysis has not already been demonstratedto be applicable to the bolt design at Fermi 2 and approved by the staff).
Response
(a) The specific analysis referred to in commitment No. 7 will address loss of preload due to stress relaxation in the core plate rim holddown bolts and quantify the loss of preload/stress relaxation that will occur in these bolts during the period of extended operation. (b) The BWRVIP is developing justification for the elimination of core plate bolt inspections for plants without installed core plate wedges. This new evaluation is different than the generic analysis evaluated in BWRVIP-25 and approved in the NRC SE of December 7, 2000. While this new evaluation will also be generic (applicable to multiple plants), each site will have to meet criteria dependent on plant-specific load combinations and peak fluence at the core plate bolt locations. Like the analysis evaluated in BWRVIP-25, the new core plate bolt evaluation will address loss of preload/stress relaxation through the period of extended operation. to NRC-15-0010 Page 19 (c) The new evaluation described above will be performed when instructed by the BWRVIP through incorporation into an existing BWRVIP Inspection & Evaluation Guideline. The evaluation will be performed under the provisions of BWRVIP implementation guidelines in BWRVIP-94, Revision 2, which also provides for NRC notification and submittal as appropriate. Therefore, notification and submittal to the NRC will be performed in accordance with BWRVIP-94, Revision 2. If Option 2 of Commitment No. 7 is performed and the analysis results in acceptance criteria for continued inspection, then the inspection plan, along with the acceptance criteria and justification for the inspection plan, will be submitted to the NRC two years prior to the period of extended operation. The License Renewal Application (LRA) will be revised to reflect this response. LRA Revisions: LRA Sections A.1.10, A.4, B.l.10, and C are revised as shown on the following pages. Additions are shown in underline and deletions are shown in strike-through. to NRC-15-0010 Page 20 A.1.10 BWR Vessel Internals Program The BWR Vessel Internals Program will be enhanced as follows. BWR Vessel Internals Program procedures will be revised as follows. In accordance with an applicant action item for BWRVIP-25 safety evaluation: (a) install core plate wedges prior to the period of extended operation, or (b) complete a plant-specific analysis that justifies no inspections are required or to determine acceptance criteria for continued inspection of core plate hold-down bolts in accordance with BWRVIP-25.1 b.olts- durna twhe perioa d of, xtendd operatonr
,s i-sas-the, ri the analvsi reu.t in
'e l .g?fi a _ ojd ns tt. inspection plan, along with the acceptance criteria and justification for the inspection plan, y U sjAmiei_ to the NRC two years prior to the period of extended operation r .djJo-ai r ' lR /jjbe E i
'i r47it' d¢ 4 . i ris f m d 4 m ee krCY5fb -. rs ia to NRC-15-0010 Page 21 A.4 LICENSE RENEWAL COMMITMENT LIST No. Program or Activity Commitment mee n Source 7 BWR Vessel Internals Enhance BWR Vessel Internals Program as follows: Perform initial .1.10 inspection either
- c. BWR Vessel Internals Program procedures will be prior to March 20, revised as follows. In accordance with an applicant action 2025 or before item for BWRVIP-25 safety evaluation: (a) install core plate March 20, 2030.
wedges prior to the period of extended operation, or (b) Submit inspection complete a plant-specific analysis that justifies no plan to NRC prior inspections are required or to determine acceptance criteria to March 20, for continued inspection of core plate hold-down bolts in 2023. accordance with BWRVIP-25 _,a3j Va vij ad s Remaining SsOf; r e to se i jxa ntio ri activities: hc jt and quanjt ___oss Prior to g i lC uEinLte gdg rjje eno d o September 20, xgb °M s _y w rt, des j ,j 2024, or the end n rejs in inspection of the last plan, along with the acceptance criteria and justification for refueling outage the inspection plan, wjjj e tdinjttedto the NRC two years prior to March 20, prior to the period of extended operation ._,djtj n _e 2025, whichever R Rjr3A baevjseo ad res is later. to NRC-15-0010 Page 22 B.1.10 BWR Vessel Internals Enhancements Element Affected Enhancement
- 4. Detection of Aging Effects BWR Vessel Internals Program procedures will be revised as follows. In accordance with an applicant action item for BWRVIP-25 safety evaluation: (a) install core plate wedges prior to the period of extended operation, or (b) complete a plant-specific analysis that justifies no inspections are required or to determine acceptance criteria for continued inspection of core plate hold-down bolts in accordance with BWRVIP-25_h ajs yji
'd d s jgss 9- ooig rs. t 5 regs q i r tjaj
'lat i m h oi dow bo ts n d G u i y thee_ ..
th ceo r ; rjfrj r J M the inspection plan, along with the acceptance criteria and justification for the inspection plan, vj ibe ujpmiWu d;to the NRC two years prior to the period of extended operation. __a__A_____tjs___tw____ars__tor__ _t ps;at_ 4o ; extnde opertio.
Enclosure 1 to NRC-15-0010 Page 23 Appendix C Response to BWRVIP Applicant Action Items Action Item Description Response BWRVIP-25 (4) Due to susceptibility of the rim hold-down For BWRs that do not have core plate wedges, bolts to stress relaxation, applicants BWRVIP-25 recommends evaluation of two referencing the BWRVIP-25 report for aging effects on the core support plate hold-down license renewal should identify and evaluate bolts: loss of preload and cracking. Fermi 2 is a the projected stress relaxation as a potential BWR 4 without core plate wedges, so these TLAA issue. aging effects apply and are evaluated as follows. Prior to the period of extended operation, Fermi 2 will enhance the BWR Vessel Internals Program (refer to Appendix B, Section B.1.10) to perform one of the following.
- Install core plate wedges prior to the period of extended operation, or
- Complete a plant-specific analysis th-ajustifies no rns eoona regtujred o to determine acceptance criteria for continued inspection of core plate hold-down bolts in accordance with BWRVIP-25 j-a n j j = dre so H
Freloa d due to str ess rel'aaton ini the Core SfiPecor.einspection plan, along with the acceptance criteria and justification for the inspection plan, wiHt um to the NRC two years prior to the period of extended operation.. ;jj9o a'j ~~aArd
=he v;ili be revied adres o te aaly operation.'~
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Enclosure 1 to NRC-15-0010 Page 24 RAI 4.1-3
Background
In the staff's December 7, 2000, safety evaluation (ML003776110) on EPRI Technical Report BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines, " the staff includedAAI No. 4 for identificationof anyplant-specific TLAAs that may be applicable to the evaluation of BWR top guide components. Specifically, AAI No. 4 states that BWR applicantsfor license renewal should identify and evaluate the impact of accumulatedneutronfluence on the potential to initiate irradiation-assistedstress corrosion cracking (IASCC) in BWR top guide components and evaluate whether such an evaluation is a TLAA. In response to AAI No. 4, LRA Appendix C states that the 60-yearprojectedfluence exceeds the thresholdfor the initiationof IASCC in the Fermi 2 top guide and its subcomponents. However, the LRA also states that the methodology in BWRVIP-26-A does not include any analyses that would constitute a TLAA for Fermi 2 because this reportwas not used to make any safety determinationor tojusti a reduction to the number of inspectionsfor these components. The LRA further states that, since the applicanthas implemented the inspection requirements of BWRVIP-26-A and BWRVIP-183, the BWR Vessel InternalsProgramwill adequately manage the effects of aging on the top guide assemblyfor the period of extended operation. Issue Appendix B of BWRVIP-26-A includes a genericflaw analysisfor postulated cracks in BWR top guide grid beam components. This flaw analysis uses a proprietaryupper boundfluence value as the basisfor the criticalstress intensity value. Therefore, it is not evident as to why the neutronfluence-dependent IASCC analysisfor the top guide grid beam locationswould not need to be identified as a TLAA, particularlybecause the applicantis relying on theflaw evaluation in BWRVIP-26-A to justify the conservatisms and validity of the augmented inspection methods and frequenciesfor the top guide grid beam locations at Fermi 2. Request Clarify whether theflaw evaluationfor BWR top guide gridbeam locations in BWRVIP-26-A, Appendix B, is relied upon to justfy the conservatisms and validity of the augmented inspection methods andfrequenciesfor the top guide grid beam locationsat Fermi 2. If so, provide justification as to:
- 1) Why the flaw evaluationfor verification of the inspection andflaw evaluation methods in BWRVIP-26-A is not partof the safety basis decision or determinationfor implementing the BWRVIP-26-A guidelines as partof the BWR Vessel Internals Program,and
- 2) Why the genericflaw evaluationfor the BWR top guide grid beam locations in BWRVIP-26-A, Appendix B, has not been identified as a TLAA when comparedto the six criteria for TLAAs in 10 CFR 54.3(a).
to NRC-15-0010 Page 25
Response
Appendix B of BWRVIP-26-A describes an example flaw evaluation of selected locations of BWR top guide grid beams. Inspection guidelines are not tied to the results of this example evaluation. The example evaluation results are not relied upon to justify the conservatisms and validity of augmented inspection methods and frequencies for the top guide grid beam locations at Fermi 2. BWRVIP-183, Top Guide Grid Beam Inspection & Flaw Guidelines, establishes the inspection strategy. Fermi 2 implements the inspections in accordance with the plant procedure that implements the BWR Vessel Internals Program. License Renewal Application (LRA) Table 3.1.2-2 identifies the aging effect of cracking for the top guide and credits the BWR Vessel Internals Program and Water Chemistry Control - BWR Program for managing the aging effect. The fluence for the top guide does not need to be treated as a TLAA because the inspection strategy established in BWRVIP-183 assumes that the fluence threshold for irradiation assisted stress corrosion cracking has been exceeded for the top guide. The inspections conducted at Fermi 2 as part of the BWR Vessels Internals Program to implement the inspection strategy in BWRVIP-183 manage the postulated cracking as an aging effect. LRA Revisions: None. to NRC-15-0010 Page 26 RAI 4.1-4
Background
The staff's December 20, 1999, safety evaluation (ML993630179 and ML993630186) on EPRI Technical Report BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines," includesAAI No. 4 for addressingplant-specific TLAAs that may be applicable to the evaluation of BWR standby liquid control (SLC) and core AP nozzle components. A AINo. 4 states that BWR applicantswho reference BWRVIP-27-A for license renewal should identify and evaluate the projectedfatigue cumulative usagefactors as a potential TLAA for their SLC and core AP lines. In response to AAI No. 4, LRA Appendix C states that the BWRVIP-27-A fatigue analysisfor the SLC and core AP line for 60 years of operation is a TLAA. The LRA also states that, at Fermi 2, the SLC and core AP lines inside the reactorpressure vessel (RP V) are not subject to an AMR. Issue (a) UFSAR Section 4.5.2.4.3 states that the SLC system is needed to remain operable in order to comply with the requirementsof 10 CFR 50.62, "Requirementsfor Reduction of Riskfrom AnticipatedTransients without Scram (ATWS) Eventsfor Light-Water-CooledNuclear Power Plants." UFSAR Section 4.5.1.2.11 states thatportions of the SLC and core AP lines internalto the RPV are needed to facilitate good mixing and dispersionof boron into the RPV when the SLC system is activated. The UFSAR also states that the portions of the SLC and core AP lines internalto the RPV also reduce thermal shock to the SLC and core AP nozzle. Given this information, it is not clear why the portions of the SLC and core AP lines internal to the RPV are not identified as within the scope of license renewal in accordance with: e 10 CFR 54.4(a)(2), because the failure of the SLC line inside the RPV would result in its becoming incapable of mitigating a thermal shock to the RPV's SLC and core AP nozzle (a safety-related component that is partof the reactor coolantpressure boundary component), and e 10 CFR 54.4(a)(3), because the SLC line inside the RPV is relied upon to properly mix and disperse boron-10 inside the reactorfollowing an ATWS event. (b) The applicant'sresponse to AAI No. 4 does not sufficiently demonstrate that the LRA does not need to include a metalfatigue analysis (i.e., CUFanalysis) or other type of cycle loading TLAA for those portions of the SLC and core AP line that are internalto the RPV. Request (a) Justi why those portions of the SLC and core AP line internal to the RPV have not been identifiedas within the scope of license renewal. In the response, indicate whether these components are in-scope under 10 CFR 54.4(a)(1), 10 CFR 54.4(a)(2), or 10 CFR
Enclosure I to NRC-15-0010 Page 27 54.4(a)(3). If these components are within the scope of license renewal,provide the basisfor why they are not subject to aging management review, as requiredby 10 CFR 54.21(a)(1) for passive, long-lived structures,systems and components. Amend the LRA accordingly if it is determined that these components are subject to an AMR. (b) Identifj> the design code or design analyses of record usedfor the design of those portions of the SLC and core AP lines that are internal to the RPV (i.e., not inclusive of the SLC and core AP nozzle adjoined to the RPV). Clarify whether the design code or design analyses of record include a metalfatigue analysis or other type of cyclical loading analysis (e.g., cycle-based expansion stress or maximum allowable stress range reduction analysis or afatigue waiver analysis)for thoseportions of the SLC and core AP line that are internal to the RPV If so, explain why the analysis would not need to be identifiedas a TLAA when compared to the six criteriafor TLAAs in 10 CFR 54.3(a).
Response
(a) As stated in BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines," Appendix B License Renewal Appendix, "The SLC piping is designed to ensure the capability to shut down the reactor and maintain it in a safe-shut down condition (54.4(a)(1)(ii)) and prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to 10 CFR 100 guidelines (54.4(a)(1)(iii)). The core plate AP line provides information on core flow performance for diagnostic purposes. This function is not needed to accomplish any of the requirements in 54.4(a). Therefore, the intended function for license renewal is limited to the function of the SLC piping. That is, to shut down the reactor from full power by injecting a neutron absorber (referred to as boron in the report) into the reactor core when the normal method of controlling core reactivity with control rods cannot be accomplished." Updated Final Safety Analysis Report (UFSAR) Section 4.5.1.2.11 states that portions of the SLC and core AP lines internal to the RPV are needed to facilitate good mixing and dispersion of boron into the RPV when the SLC system is activated. BWRVIP-27-A concludes that the boron need only reach the bottom head and there is sufficient natural circulation to distribute boron uniformly through the core. Therefore, the distribution of liquid by the SLC and core AP line is not essential and the portions of the SLC and core AP lines internal to the RPV do not perform a license renewal intended function. (b) The response to (b) contains proprietary information and is provided in Enclosure 5 of this letter. 1That is systems, structures, or components that are not active or subject to moving parts, or that are not subject to replacement based on a specified time period or qualified life. to NRC-15-0010 Page 28 LRA Revisions: None.
Enclosure 1 to NRC-15-0010 Page 29 RAI 4.1-5
Background
In its response to AAI No. 6 on EPRI Technical Report BWRVIP-76-A, "BWR Core Shroud Inspection andFlaw Evaluation Guidelines," LRA Appendix C states that the core shroudis fabricatedfrom Type 304L stainlesssteel and that its aging effects are loss of materialand cumulative fatigue damage. The LRA states that the BWR Vessel InternalsProgramand Water Chemistry Control - BWR Programwill manage loss of material due to pitting and crevice corrosionthat may occur in the core shroudduringthe periodof extended operation. The LRA also states that no cracking of vertical (axial) or horizontal (circumferential)weld seams in the core shroud have been detected; therefore, no repairdesign modificationsfor the core shroud have been implemented. The LRA further states that the metalfotigue TLAAs for the RVI components are evaluated in LRA Section 4.3.1.4. Issue LRA Appendix C states that the evaluation of the metalfatigue TLAAs for the RVI components are in LRA Section 4.3.1.4; however LRA Section 4.3.1.4 does not indicate which design code or specificationwas usedfor the design andfabricationof the core shroud, nor does it indicate whether the design code or specification requireda metalfatigue analysis or other type of cyclical loading analysisfor the core shroudand its subcomponents. Request Identify the design code or design specificationof record that was usedfor the design and fabricationof the core shroud. Clarify whether the design code or design specification of record requireda metalfatigue analysisor other type of cyclical loading analysis (e.g., cycle-based expansion stress or maximum allowablestress range reductionanalysis or a fatigue waiver analysis)for the design of the core shroud. If so, providejustificationas to why the analysis would not need to be identifiedas a TLAA when comparedto the six criteriafor TLAAs in 10 CFR 54.3(a).
Response
For BWR-4 plants such as Fermi 2, the core shroud is not an ASME Code component, but was designed using the ASME Code as a guideline. There was not a specific design code or design specification for the shroud. Fabrication of the shroud was performed using welders and techniques qualified in accordance with ASME Section IX. As discussed in License Renewal Application (LRA) Sections 3.1.2.2.1 and 4.3.1.4, the RPV internals and core support structure including the shroud, are not part of the reactor coolant pressure boundary. Since the shroud is not an ASME Code component, fatigue and cyclical load evaluation was not a Code requirement. Fatigue was not a significant design consideration and a fatigue analysis was not performed for the shroud. This was verified with the original equipment supplier. Since no analysis was to NRC-15-0010 Page 30 performed, no analysis was identified as a TLAA for the shroud. However, fatigue analyses were performed for selected internal components where fatigue is a design consideration as identified in LRA Table 4.3-3. LRA Revisions: None.
Enclosure 1 to NRC-15-0010 Page 31 RI 41-6
Background
LRA Section 4.1.2 states that the applicant performed a search to find any exemptions that were grantedfor the Fermi 2 CLB in accordancewith the exemption approval criteria in 10 CFR 50.12 and based on a TLAA. The LRA states that this search was based on a review of relevant licensing basis or design basis information in the UFSAR, American Society of Mechanical Engineers (ASME) Code, Section XI, program documentation, fire protection documentation, operating license, Technical Specifications,and docketed correspondence. The LRA states that the applicant did notfind any exemptions that are based on a TLAA and that will remain in effect for the periodof extended operation. Issue Pursuant to 10 CFR 54.21(c)(2), the applicant is required to identify a particular exemption as partof the LRA if the exemption was grantedin accordance with the requirements of 10 CFR 50.12 and the exemption is based on a TLAA. For exemptions that meet these criteria, the requirements in 10 CFR 54.21(c)(2) apply regardlessof whether the exemptions will remain in effect for the period of extended operation. Therefore, LRA Section 4.1.2 may have omitted exemptions that were granted in accordance with 10 CFR 50.12 and are based on a TLAA, but will not remain in effect for the period of extended operation. Any such omissions would not be in compliance with the requirements of10 CFR 54.21(c)(2). Request Identify all exemptions that were granted in accordance with 10 CFR 50.12 and are based on a TLAA but will not remain in effect for the periodof extended operation.
Response
There are no exemptions that were granted in accordance with 10 CFR 50.12 that are based on a TLAA that are currently in effect, but will not remain in effect for the period of extended operation. LRA Revisions: None.
Enclosure 1 to NRC-15-0010 Page 32 RAI4.1-7 Backeround LRA Section 4.1.2 states that the applicantperformed a search to find any exemptions that were grantedfor the Fermi 2 CLB in accordancewith the exemption approvalcriteriain 10 CFR 50.12 and based on a TLAA. The LRA states that this search was based on a review of relevant licensing basis or design basis information in the UFSAR, ASME Code, Section XI, program documentation,fire protection documentation, operatinglicense, Technical Specifications, and docketed correspondence. The LRA states that the applicantdid not find any exemptions that are based on a TLAA and that will remain in effect for the periodof extended operation. Issue UFSAR Section 6.2 states that the NRC granteda number of exemptionsfrom meeting the requirements of 10 CFR Part50, Appendix J, Option B, for the containment leak rate testing program. The UFSAR does not describe what these exemptions involve or whether the alternative testing requirements or exceptions authorizedby the exemptions are basedon or supportedby a time-dependent analysis. Therefore, additionalinformation is needed to determine whether these exemptions are based on a TLAA. Request Describe each exemption from the 10 CFR Part50, Appendix J, leak rate testing requirements and explain whether the alternative testing requirements or exceptions authorizedby each exemption are basedon or supportedby a time-dependent analysis, calculation,or evaluation that conforms to the six criteriafor TLAAs in 10 CFR 54.3. If it is determined that a specific 10 CFR Part50, Appendix J, leak rate testing exemption was granted under 10 CFR 50.12 and is based on a TLAA, amend the LRA, as appropriate,to identify and evaluate the exemption in accordancewith the requirements in 10 CFR 54.21(c)(2).
Response
Exemptions have been granted to the 10 CFR Part 50, Appendix J, leak rate testing requirements over the life of the plant as follows: a) Main steam isolation valves (MSIVs) are tested at a lower pressure than peak calculated containment pressure associated with the design basis accident and their leakage is not included in the combined Type B and Type C local leak rate test leakage totals, because of the main steam leakage control system (MSLCS). Reduced pressure testing between the valves, and not including the leakage in the Type B and Type C leakage totals, was approved in Fermi 2 Safety Evaluation Report, Supplement 5 (SSER #5). License Amendments 160 and 169 documented the removal of the MSLCS and replaced the justification of excluding leakage from the Type B and Type C total based on Alternate Source Term, and increased the allowable MSIV leakage to NRC-1 5-0010 Page 33 further, respectively. The bases for SR 3.6.1.3.12 states that "...This leakage test is performed in lieu of 10 CFR 50, Appendix J, Type C test requirements, based on an exemption to 10 CFR 50, Appendix J. MSIVs have separate leakage limits, and the dose consequence of this leakage path is evaluated separately and added to those calculated from Primary containment La leakage. As such, this leakage is not combined with the Type B and C leakage rate totals. The frequency is required per the Primary Containment Leakage Rate Testing Program." This exemption is not based on a time-limited analysis. b) An exemption allowed Fermi 2 to perform alternate testing of the air lock, which consisted of testing the seals of the inner and outer doors at Pa rather than testing the entire air lock, after the opening of the doors. The deletion of this exemption was approved in License Amendment 108 based on implementation of 10 CFR 50, Appendix J, Option B based on the guidance in NEI 94-01 as endorsed by Regulatory Guide 1.163. This exemption was not based on a time-limited analysis and it is no longer in effect. c) Reclassification of the inboard containment low pressure coolant injection (LPCI) valve configuration as other than containment isolation valves and no longer subject to Type C testing, by using an alternate testing method for these valves was approved by License Amendment 98 and Exemption to the Facility Operation License No. NPF-43 Fermi 2. The LPCI penetrations X-13A & B were designed and built to General Design Criteria 55, requiring both an inboard and outboard isolation capability. License Amendment 98 took exemption to 10 CFR 50, Appendix J testing based on identifying two alternative barriers, which included the residual heat removal system being a closed loop outside of primary containment and that penetrations (X-13A & B) will remain water sealed during a loss-of-coolant accident. This exemption is not based on a time-limited analysis. d) Alternate techniques to the mass plot method for analyzing Type A test data had been approved. The need for this exemption was eliminated when the Primary Containment Leakage Testing Program was implemented per 10 CFR 50, Appendix J, Option B based on NEI 94-01 and Regulatory Guide 1.163 as approved in License Amendment 108. This exemption was not based on a time-limited analysis and it is no longer in effect. e) and f) Two one-time schedule exemptions had been issued and expired. These schedule exemptions were not based on a plant lifetime related time-limited analysis. Therefore, none of the above exemptions are based on or supported by a time-dependent analysis, calculation, or evaluation that conforms to the six criteria for TLAAs in 10 CFR 54.3. LRA Revisions: None.
Enclosure 3 to NRC-15-0010 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 to GEH Letter 318178-11, "Response to RAI 4.7.41" NON-PROPRIETARY
ENCLOSURE 2 318178-11 Response to RAI 4.7.4-1 Non-Proprietary Information - Class I (Public) INFORMATION NOTICE This is a non-proprietary version of Enclosure 1 of 318178-11, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside open and closed bracket as shown here [[ I ].
Enclosure 2 Non-Proprietary Information - Class I (Public) 318178-11 Page 1 of 4 RAI 4.7.4-1
Background:
LRA Sections 4.4.7 and A.2.5.4 describe the slip joint repair clamps as being connected to the diffuser and the mixer (throat) in the jet pump assembly. The LRA states that the clamps were installed with a preload that may decrease due to neutron fluence and thermal exposure. The LRA also states that the analysis that evaluated the decrease of the installation preload for the slip joint repair clamp is a TLAA that has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii). The LRA further states that after 52 effective full-power years (EFPY) of plant operation the expected fluence at the location of the repair clamps is 3.07E+18 n/cm2 (E > 1 MeV), which is below a level necessary (1.0E+19 n/cm 2) to cause stress relaxation in stainless steel. Issue: The staff lacks sufficient information to evaluate the jet pump slip joint repair clamp TLAA (LRA Sections 4.4.7) for the period of extended operation and determine if the Updated Final Safety Analysis Report (UFSAR) supplement, LRA Section A.2.5.4, adequately summarizes the TLAA in accordance with 10 CFR 54.21(d). The LRA does not include the following information that the staff needs for its determination: (a) the intended function of the jet pump slip joint repair clamps, (b) how the loss of preload affects the capability of the clamps to perform their intended function, (c) a physical description of the slip joint repair clamp, and (d) the specific methodology and details of the methodology that was used to assess loss of preload in the jet pump slip joint repair clamps during the period of extended operation. Request:
- 1) State the intended function of the jet pump slip joint repair clamps and how the loss of preload affects the capability of the clamps to perform their intended function.
- 2) Provide a physical description or drawing of the slip joint repair clamps. The level of detail in the description should provide for an understanding of the style of clamp construction (e.g., bolted, pressed, pined, keyway) and how the clamps are retained in the jet pump assembly.
- 3) Provide summaries of the stress and fluence analysis, as applicable, used to evaluate the jet pump slip joint repair clamps for the period of extended operation. The summaries should include the:
e methodology and pertinent details of the methodology used in the analysis e calculations (mathematical modeling, including pertinent safety assumptions or coefficients used in the modeling) used to evaluate the decrease in preload e key variables used to evaluate the decrease in preload, such as the design basis preload and minimum value of the preload required for the clamps to perform their intended function o basis used to establish the fluence threshold of 1.0E+19 n/cm2 to cause stress relaxation in stainless steel basis used to determine that the loss of preload that has occurred prior to entering the period of extended operation is acceptable during the period of extended operation Additionally, if the analysis used to evaluate the jet pump slip joint repair clamps utilized any industrial topical reports or methodologies reviewed and approved by the NRC, provide the references for these documents and the dates of the staff's safety evaluation (SE) approving the reports. Non-Proprietary Information - Class I (Public) 318178-11 Page 2 af 4
Response
The following RAI response contains GEH proprietary information:
- 1. The intended function of the slip joint clamp is to act as a lateral preload restraint by forcing the jet pump inlet mixer against the inner surface of its diffuser collar. This action suppresses abnormal jet pump vibration that is caused by excessive leakage through the slip joint. A loss of preload would preclude the slip joint clamp from its ability to perform this function.
- 2. Drawings of the slip joint clamp are provided. The drawings are taken from Fermi 2 site procedures in support of visual examination of the clamp.
SLIP JOINT CLAMP INLET4MER BOT-RA CHET LOKI NG GUDE VANES SPRING (x DIFFUJSER CATING Non-Proprietary Information - Class I (Public) 318178-11 Page 3 of 4 NOTE #1 DIFFUSER
$NLETMER O, \ SO SL~pORr STRUT (aX)
NOTE#5 CLAMP BOLT - CLAMP BODY INSPECTION POINTS!NOTES RATCHET LOCKING SPR9NO LOCATED IN THIS SURFACE i GUIDE VANE SEATED INCAMrP CROOVE. SEE FIGURE ON PREVIOUS PAGE, 2, CiMP MIDDLE SECTION IS INCONTACT WITH NOTE DFFUSER GUIDE VANE.
- a. CLAAW BOLT IS INCONTACT WITH ThE 00 OF THE R4LETXER SHELL, 4 RATCHET LOCKING SPRING IS ENGAGED WITH NO PHYSICAL DEGRADATION.
5 NTACT WITh UPPER SURFACE OF THE DiPFUSER COLLAR ISNOT REQUIRED. LOOK FOR EVDENCE OF WEAR AT STRUT/DIFFUSER CONTACT POINT
- 3. [[
]] The basis used to determine that the loss of preload occurring prior to entering the period of extended operation is acceptable was the performance of a slip joint clamp relaxation evaluation. The evaluation was to determine if the original structural evaluation (considering a design life of 40 years) would remain applicable for 52 effective full power years (EFPY).
The flux synthesis method employed in the evaluation is described in section 2.1 of NEDC-32983P-A:
"Licensing Topical Report General Electric Methodologyfor Reactor Pressure Vessel Fast Neutron Flux Evaluations," Revision 2, January 2006. This methodology, [[
Non-Proprietary Information - Class I (Public) 318178-11 Page 4 of 4
]], has been approved by the NRC via MFN-05-143 and MFN-01-050 and is adherent to Regulatory Guide 1.190.
The flux analysis models the [[
]]
Using the flux synthesis method, the flux for the slip joint clamp isdetermined. Several conservative assumptions were made in support of the flux analysis:
~ Flux at the peak azimuthal location was used to bound all azimuths. - [ ]]
because the jet pump assemblies are located within the annulus.
- Since the slip joint clamp elevation is outside the beltline, [[
]], was applied to bound the results.
- For the purpose of the fluence calculation, the jet pump slip joint clamp was conservatively assumed to have been installed at beginning of plant life.
The jet pump slip joint clamp fluence was calculated to be 3.07E18 n/cm2 (E > 1 MeV) over the service life of 52 EFPY. However, at fluence levels less than or equal to 1E19 n/cm2, neutron irradiation does not impact the amount of expected relaxation as considered in the original slip joint clamp stress analysis. The basis used to establish the fluence threshold of 1.0E+19 n/cm 2 is from a curve contained in a proprietary document developed and verified by the GEH Materials Engineering Group. The description of the data used and the utilization of the data in establishing the curve contained in this proprietary document was reviewed by the NRC as part of the RAI response located in ML051940291. It is concluded that the relaxation value from the original analysis as well as the results of the slip joint clamp stress report remain applicable for the period of extended operation.
Enclosure 4 to NRC-15-0010 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318178-11
ENCLOSURE 3 318178-11 Affidavit for Enclosure 1
GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Lisa K. Schichlein, state as follows: (1) I am a Senior Project Manager, NPP/Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in Enclosure 1 of GEH letter 318178-11, "Proprietary Review of Fermi 2 License Renewal Application RAI 4.7.4-1," dated January 22, 2015. The GEH proprietary information in Enclosure 1, which is entitled "Response to RAI 4.7.4-1," is identified by a dotted underline inside double square brackets. This sentence is an examnple. 3 In each case, the superscript notation 3 refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C. Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enemy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983). (4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;
- b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
- d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.
(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my Affidavit for 318178-11 Page 1 of 3
GE-Hitachi Nuclear Energy Americas LLC knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7). (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements. (8) The information identified in paragraph (2), above, is classified as proprietary because it contains details on the GEH fluence methodology for boiling water reactors (BWRs). Development of these methods, techniques, and information and their application for the design, modification, and analyses methodologies and processes was achieved at a significant cost to GEH. The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their Affidavit for 318178-11 Page 2 of 3
GE-Hitachi Nuclear Energy Americas LLC own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 22nd day of January 2015. Lisa K. Schichlein Senior Project Manager, NPP/Services Licensing Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Road, M/C A-65 Wilmington, NC 28401 Lisa.Schichlein@ge.com Affidavit for 318178-11 Page 3 of 3
Enclosure 6 to NRC-15-0010 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 to GEH Letter 318178-7, "Response to I 4.1-4(b)" NON-PROPRIETARY
ENCLOSURE 2 318178-7 Response to RAI 4.1-4(b) Non-Proprietary Information - Class I (Public) INFORMATION NOTICE This is a non-proprietary version of Enclosure 1 of 318178-7, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside open and closed bracket as shown here [[ ]].
Enclosure 2 Non-Proprietary Information - Class I (Public) 318178-7 Page 1 of 2 RAI 4.1-4(b)
Background:
The staff's December 20, 1999, safety evaluation (ML993630179 and ML993630186) on EPRI Technical Report BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines," includes AAI No. 4 for addressing plant-specific TLAAs that may be applicable to the evaluation of BWR standby liquid control (SLC) and core AP nozzle components. AAI No. 4 states that BWR applicants who reference BWRVIP-27-A for license renewal should identify and evaluate the projected fatigue cumulative usage factors as a potential TLAA for their SLC and core AP lines. In response to AAI No. 4, LRA Appendix Cstates that the BWRVIP-27-A fatigue analysis for the SLC and core AP line for 60 years of operation is a TLAA. The LRA also states that, at Fermi 2, the SLC and core AP lines inside the reactor pressure vessel (RPV) are not subject to an AMR. Issue: (b) The applicant's response to AAI No. 4 does not sufficiently demonstrate that the LRA does not need to include a metalfatigue analysis (i.e., CUF analysis) or other type of cycle loading TLAA for those portions of the SLC and core AP line that are internal to the RPV. Request: (b) Identify the design code or design analyses of record used for the design of those portions of the SLC and core AP lines that are internal to the RPV (i.e., not inclusive of the SLC and core AP nozzle adjoined to the RPV). Clarify whether the design code or design analyses of record include a metal fatigue analysis or other type of cyclical loading analysis (e.g., cycle-based expansion stress or maximum allowable stress range reduction analysis or a fatigue waiver analysis)for those portions of the SLC and core AP line that are internal to the RPV. If so, explain why the analysis would not need to be identified as a TLAA when compared to the six criteria for TLAAs in 10 CFR 54.3(a). GEH Response (b) The design analysis of record used for the design of those portions of the SLC and core AP lines that are internal to the RPV is Reference 4.1-4(b)-1, which is an internal proprietary GEH document. Both fatigue analysis and flow induced vibration analysis were included in the design. The fatigue analysis was performed in accordance with ASME Code NB 3650 and resulted in an alternating stress of [[ ]] psi, which is less than 10,000 psi. For the flow induced vibration analysis, the natural frequency of the SLC and core AP line was well above the excitation frequencies due to flow. The natural frequency was more than [[ ]] times the excitation frequency during normal operation and more than [[ ]] times the excitation Non-Proprietary Information - Class I (Public) 318178-7 Page 2 of2 frequency due to single loop operation. Hence the flow induced alternating stresses will be small, much less than the 10,000 psi 0-peak stress limit. The GEH criteria of 10,000 psi peak stress intensity is much more conservative than the ASME allowable peak stress intensity of 13,600 psi. There is no change in structural natural frequency or excitation frequency due to a 60 year life extension; hence, the design analysis of record is still valid. This is additionally confirmed by an in-reactor test performed at a BWR/5-251 prototype plant. The test results at that plant showed that the maximum measured stress under any plant operating condition was [[ ]] of the acceptance criteria or [[ ]] psi, which is well below the 10,000 psi 0-peak. A value of less than 10,000 psi, 0-peak implies that no fatigue usage is accumulated by the component (Figure 1-9.2.2, Design Fatigue Curve for Austenitic Steels, in Reference 4.1-4(b)-2). In such a case, a 60 year life will have no effect on component evaluation. References 4.1-4(b)-1 General Electric Company, "Differential Pressure and Liquid Control Line Stress Report," 22A5557, October 31, 1977. (Proprietary Internal Document) 4.1-4(b)-2 ASME Section III, Division 1, Appendix I.
Enclosure 7 to NRC-15-0010 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 GE-Hitachi Nuclear Energy Americas LLC Affidavit for Enclosure 1 of 318178-7
ENCLOSURE 3 318178-7 Affidavit for Enclosure 1
GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Lisa K. Schichlein, state as follows: (1) I am a Senior Project Manager, NPP/Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in Enclosure 1 of GEH letter 318178-7, "GEH Response to Fermi 2 License Renewal Application RAI 4.1-4(b)," dated January 14, 2015. The GEH proprietary information in Enclosure 1, which is entitled "Response to RAI 4.1-4(b)," is identified by a dotted underline inside double square brackets. [[This. sentence is an example 13 11 In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of InformationAct ("FOIA"), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C. Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983). (4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;
- b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
- d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.
(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my Affidavit for 318178-7 Page 1 of 3
GE-Hitachi Nuclear Energy Americas LLC knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7). (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH, (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements. (8) The information identified in paragraph (2), above, is classified as proprietary because it contains details on the GEH methodology for the design of the standby liquid control lines and core pressure differential lines internal to the reactor pressure vessel for boiling water reactors (BWRs). Development of these methods, techniques, and information and their application for the design, modification, and analyses methodologies and processes was achieved at a significant cost to GEH. The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their Affidavit for 318178-7 Page 2 of 3
GE-Hitachi Nuclear Energy Americas LLC own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to the public. Making such infornation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 14th day of January 2015. Lisa K. Schichlein Senior Project Manager, NPP/Services Licensing Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Road Wilmington, NC 28401 Lisa.Schichlein@ge.com Affidavit for 318178-7 Page 3 of 3}}