ML17334B815
| ML17334B815 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/08/1998 |
| From: | Sampson J INDIANA MICHIGAN POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:0969BK, AEP:NRC:969BK, NUDOCS 9807140309 | |
| Download: ML17334B815 (14) | |
Text
CATEGORY 1 REGULATO INFORMATION DISTRXBUTION ~ STEM (RIDS)
ACCESSION NBR:9807140309 DOC.DATE: 98/07/08 NOTARIZED: NO DOCKET FACIL:50-315iDonald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315
>50-916 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION SAMPSON,J.R. Indiana Michigan Power Co.
RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)
SUBJECT:
Provides info requested in 980114 RAI &. notes of withdrawal of relief request number seven, performing full-vee exam from vessel side in lieu of performing half-vee exam from both sides of pressurizer surge nozzle-to-vessel weld.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code - GL-89-04 E NOTES:
RECIPIENT COPIES " RECIPIENT COPIES 0 ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-3 LA 1 1 PD3-3 PD 1 1 STANG,J 1 1 XNTERNAL: AEOD/SPD/RAB 1 1 CENTEE~~O 1 1 NRR/DE/ECGB 1 1 NUDOCS -ABSTRACT 1 1 OGC/HDS3 1 0 RES/DET/EIB 1 1 RES/DET/EMMEB 1 1 EXTERNAL: LITCO ANDERSON 1 1 NOAC 1 1 NRC PDR 1 1 D 0
E NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 13 ENCL 12
Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 491071395 ENQIAi84 M CHITIN PQIPM July 8, 1998 AEP:NRC:0969BK Docket Nos.: 50-315 50"316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop 0-P1-17 Washington, D.C. 20555-0001 Gentlemen:
Donald C. Cook Nuclear Plant Units 1 and 2 ADDITIONAL ZNFORMATZON REGARDING THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN
References:
Letter AEP:NRC:0969AJ, "Donald C. Cook Nuclear Plant Units 1 and 2, RELIEF REQUESTS FOR THE THZRD 10-YEAR ZNSERVZCE INSPECTION PLAN", dated January 25, 1996.
- 2. Letter, John B. Hickman, NRC, to E. E. Fitzpatrick, Indiana Michigan Power'Company, "REQUEST FOR ADDITIONAL INFORMATION REGARDING THE D.C. COOK NUCLEAR PLANT, UNITS 1 AND 2 THZRD 10-YEAR INTERVAL INSERVZCE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF (TAC NOS. M94871 AND M94872)g dated January 14, 1998.
Cook Nuclear Plant's third 10-year inservice inspection plan, together with its associated relief request, was transmitted to the NRC in reference 1. Subsequent to this submittal, reference 2 transmitted a request for additional information regarding the test plan and the relief requests.
Attachment 1 to this letter provides the requested information and notes that we are withdrawing relief request number seven, performing a full-vee examination from the vessel side in lieu of performing half-vee examinations from both sides of the pressurizer surge nozzle-to-vessel weld. During a review of our program, we identified additional issues.to be addressed. These issues are identified in items B, E, and I of attachment 1.
Sincerely,
~24~
J. R. Sampson Site Vice President
/vlb Attachments 9807i4030'P 980708 PDR ADQCK 050003l5 8 PDR
U.S. Nuclear Regulatory Commission AEP: NRC: 1285A
,Page 2 c: Z. A. Abramson M. T. Anderson - INEEL MDEQ - DW & RPD NRC Resident Inspector C. J. Paperiello
ATTACHMENT 1 TO AEP:NRC:0969BK REQUEST FOR ADDITIONAL INFORMATION THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM AND ASSOCIATED CODE RELIEF REQUESTS
Attachment 1 to AEP:NRC:0969BK Page 1 The following presents the questions proposed by Idaho National Engineering and Environmental Laboratory's (INEEL) request for additional information (RAI), based on their review of Cook Nuclear Plant's third 10-year interval inservice inspection (ISI) program and associated co'de relief requests. Each question is followed by our response.
Item A "Request for Relief No. 2 proposes ultrasonic examination of the full examination volume of the recirculation loop nozzle-to-safe end welds in lieu of the Code-required surface examination. The request states, "Mockups of this area have been fabricated and the capability of the ultrasonic examination technique and procedure to detect outside surface indications has been demonstrated." Confirm that the ultrasonic examination technique has been demonstrated using outside surface connected cracks, and not. notches.
addition, provide documentation and/or details including when and In, where the demonstration was performed, who performed it, and who, if anyone, from the NRC witnessed the demonstration."
Res onse to Item A Southwest Research Institute (SwRI) personnel performed the demonstration in San Antonio, Texas, at the SwRI facilities on February 29, 1996, to permit Cook Nuclear Plant to use this examination alternative during the last unit 1 and 2 second 10-year interval outages. In attendance were two Cook Nuclear Plant employees and the authorized nuclear inspector (ANI) for the plant.
In accordance with IWA-2240 of the ASNE Section XI Code, 1983 Edition, 1983 Summer Addenda, this technique and procedure was demonstrated to the satisfaction of the ANZ. This demonstration was conducted on existing calibration blocks and employed notches to depict surface flaws. These notches were sized such that both the depth and length of the notches were smaller than the calculated planar flaws provided in tables IWB-3514-1 and -2. The notches and performance of the demonstration were in accordance with the requirements of ASME Section XI. A letter from the ANI approved this alternate examination based on the results of the demonstration in San Antonio.
No NRC personnel witnessed the demonstration.
Item B "Relief Request No. 4 states that limited access to the reactor vessel closure and lower head dollar plate welds prevent any examination of either weld. Considering that no volumetric examination will be performed on these welds, describe .how reasonable assurance of their structural integrity will be provided2 Are there other RPV circumferential head welds that are being examined2 What percentage of the intersecting meridional welds can be examined? In the licensee's proposed alternative, is stated that the accessible length of one closure head meridional it weld be examined. The Code requires examination of all head welds, is this a typographical error?"
Res onse to Item B Examination access to both the lower and closure head dollar plate welds is limited by the control rod drive penetrations and core to AEP:NRC:0969BK Page 2 shroud on the closure head and lower head instrument penetrations.
There are no other circumferential head welds in either of these heads. Assurance of the structural integrity of these welds is provided by the availability of the reactor coolant system leakage detection system and the VT-2 examination conducted during refueling outages at system pressure. Access to the intersecting meridional and dollar plate welds for the closure head is not possible due to the location of the core shroud and the control rod drive mechanism penetrations and the unacceptable radiation exposure associated with the removal and re-installation of this equipment to perform this examination. Access to the lower head dollar plate welds is not feasible due to the instrument penetrations that enter on both sides of the dollar plate weld. We have reviewed this examination with our ISI Vendor, who performed our first and second 10-year ZSZ reactor pressure vessel examinations, plate welds, and even it is their opinion that examination of dollar at the intersecting welds, is not feasible't is also their experience that this is true, not only for our plant vessels, but for the majority of pressurized water reactor units.
Code relief request no. 4 for both units did state that one of the meridional welds is volumetrically examined during the inspection interval as a means of ensuring structural integrity. However, the third 10-year interval ZSZ program indicates that the accessible lengths of all meridional welds are scheduled for examination during the third inspection interval in accordance with the cpde requirements. An amendment to AEP:NRC:0969AJ, code relief request no. 4 for units 1 and 2, will be submitted by August 31, 1998, to correct this mis-statement.
Xtem C "Requests for Relief Nos. 5 and 6 are for Class 2 welds within penetrations. However, relief cannot be granted on a generic basis. Identify each weld for which relief is requested."
Res onse to Xtem C Requests for relief nos. 5 and 6 requested approval to not perform examination of containment penetration welds that were inaccessible due to design. Each unit has four feedwater and eight main steam penetrations that are designed with the flued-head penetration that does not permit access to weldment. These are listed in table 1.
TABLE 1 WELD LOCATIONS Item Un t Summary Ident f cat on Descr ption 311030 1-FW-11-01S Pipe to Flued head 311480 1-FW-13-01S Pipe to Flue Head 311830 1-FW-16-01S Pipe to F ue Hea 312200 1-FW-18-01S Pipe to F ue Hea 313310 1-MS-1-10F Pipe to Flued Head 313320 1-MS-1-12F Pipe to Flued Hea 313790 1-MS-6-10F Pipe to F ue Hea 313810 1-MS-6-12F Pipe to F ue Hea 314260 1-MS-10-09F Pipe to Flue Hea 10 314280 1-MS"10-11F Pipe to Flued Head 314740 1-MS-14-09F Pipe to F ue Hea 12 314760 1-MS-14-11F Pipe to Flue Hea
Attachment 1 to AEP:NRC:0969BK Page 3 Item Un t Summary Ident f cat on Descr pt on 13 319165 2-FW-74-16S Pipe to F ued Head 14 319325 2-FW-75-16S Pipe to Flued Head 15 319486 2-FW-76" 16S Pipe to Flued Head 16 319626 2-FW-77-16S Pipe to F ue Hea 17 320610 2-MS-89-11F Pipe to F ue Hea 18 320630 2-MS-89-13S Pipe to Flue Head 19 321060 2-MS"91-09F Pipe to Flued Head 20 321080 2-MS-91-11F Pipe to Flue Hea 21 321510 2-MS-93-09F Pipe to F ue Hea 22 321530 2-MS-93-11S Pipe to Flued Head 23 321990 2-MS-95-09F Pipe to Flue Head 24 322010 2-MS-95-11S Pipe to F ue Hea Item D "Relief Request No. 7 seeks authorization to perform a full-vee examination from the vessel side in lieu of performing half-vee examinations from both sides of the pressurizer surge nozzle-to-vessel weld. Generally, full-vee path examinations are difficult to perform on cladded vessels. What is the level of confidence in the effectiveness of this examination and what portion of the Code-required volume can be examined2 Can the circumferential scans be performed to the extent required by the Code2 Provide coverage plots or a technical discussion summarizing volumetric coverage obtained for pressurizer nozzle-to-vessel Weld 2-RC-21."
Res onse to Ztem D This code relief request is withdrawn. During the last unit 2 refueling outage, we conducted a best effort examination and were able to obtain 92%. coverage. By invoking code case N-460, we are justified in withdrawing this code relief request because the code case allows coverages greater than 90%, but less than 100%, without requesting code relief. We attribute the change in coverages between the first and second 10-year intervals and the results obtained recently in 1997 to an improved examination technique.
Item B "The licensee has requested authorization to implement Code Case N-509, "Alternative Rules for the Selection and Bxamination of Class 1, 2, and 3 Zntegrally Welded Attachments". The NRC has allowed the use of Code Case N-509 provided the licensee's commit to examine a minimum of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among the Class 1, 2, and 3 systems. Confirm that this condition will be met."
Res onse to Item B We have reviewed this question with our contractor who prepared the third 10-year interval ISI long term plan and have determined that for class 1 and 2 integral attachments, we are in compliance with the code case and the provisions designated in Regulatory Guide 1.147. We have not yet determined whether class 3 integral attachment examination requirements have been adequately met.
Based on a review of our program plan, we have found that some component supports for class 3 have not been identified in the third 10-year interval and suspect that some integral attachments
Attachment 1 to AEP:NRC:0969BK Page 4 may also not be identified. We are in the process of completing this review and will either change our third 10-year interval ISZ program to reflect our program commitments, if or submit code relief requests that seek NRC approval for not in agreement, alternative examinations. We plan to complete this activity by August 31, 1998.
Item F "IWA-2441(a) requires that the inservice inspection plan identify the Code cases that will be used during the interval. Zn the licensee's September 10, 1996 response to the NRC's RAZ, Code Case N-481 is the only Code Case listed. Based on the difference between relief requests submitted with the second interval program plan and those submitted with the third interval program plan, appears that there are other Code Cases being implemented at it D. C. Cook (e.g., N-460) . Provide a list of all the Code Cases being used at Donald C. Cook, Units 1 and 2, that are included in Regulatory Guide 1.147, "Znservice Inspection Code Case Acceptability ASME Section XZ, Division 1"."
Res onse to F The following code cases will be used for the third 10-year ISI interval program plan: Code cases N-416-1, N-460, N-498-1, N-509, N491-1, N-522, N-521 and N-524.
We have previously requested code relief in our letter AEP:NRC:0969AA to use nuclear code cases N-491-1, N509, N521, and N524. N-460 is listed in Regulatory Guide 1.147, and the NRC has approved the use of N-416-1, N-498-1, and N-522.
As additional code cases are approved for use by the NRC during the third interval, we will review these for applicability to Cook Nuclear Plant and submit revisions to the plan, as required.
Item 0 "The licensee must state the specific paragraph of 10 CFR 50.55a under which the request is submitted and provide supporting justification as discussed below.
The Regulations allow a licensee to propose an alternative to CFR or ASME requirements in accordance with 10 CFR 50a(a)(3)(Z), or 10 CFR 50a(a)(3)(ii). Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative must be shown to provide an acceptable level of quality and safety, i.e., essentially be equivalent to the original requirement in terms of quality and safety. Pursuant to 10 CFR 50. 55a (a) (3) (ii), the licensee must show that compliance with the original requirement results in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components solely to provide access for examination, and development of sophisticated tooling that would result in only minimal increases in examination coverage.
A licensee may also submit a request for relief from ASME Code requirements, in accordance with 10 CFR 50.55a(g) (5) (iii),
licensee determines that conformance with certain Code requirements if a is impractical for its facility, the licensee shall notify the to AEP:NRC:0969BK Page 5 Commission and submit, as specified in 550.4, information to support that determination. When a licensee determines that an inservice inspection requirement is impractical, e.g., the system would have to be redesigned or a component would have to be replaced to enable inspection, the licensee should cite 10 CFR 50.55a(g)(5)(iii). The NRC may, giving due consideration to the burden placed on the licensee, impose an alternative examination requirement.
Provide the appropriate references to the Code of Federal Regulations for the following requests:
~ Requests 1 through 7, which were submitted without reference to a specific paragraph of the Code of Federal Regulations.
Requests provided in the letter dated November 7, 1996, which are referred to as alternatives to Code requirements."
Res onse to 0 The followi'ng table identifies the code of federal regulations part 50 reference for each of the code relief requests per your request.
AEP: NRC Code Description 10 CFR 50.55a Document No. Relief Reference Request No.
AEP:NRC:0969AJ Augment RPV shell 10 CFR 50.55a weld exam Ul/2 (g) (5) (iii)
AEP:NRC:0969AJ Class 1 nozzle to 10 CFR 50.55a safe end welds Ul/2 (a) (3) (i)
AEP: NRC: 096 9AJ RPV s e to ange 10 CFR 50.55a weld deferral U1/2 (a) (3) (i)
AEP:NRC:0969AJ 4 RPV closure an 10 CFR 50.55a lower dollar plate welds U1/2 (g) (5) (iii)
AEP:NRC:0969AJ C ass 2 pape to 10 CFR 50.55a flued head penetration welds (g) (5) (iii)
U1/2 AEP:NRC:0969J C ass 2 pape to 10 CFR 50.55a flued head welds on the whip restraint (g) (5) (iii) of MS penetrations U1/2 AEP:NRC:0969AJ Unxt 2 - pressurizer With rawn surge nozzle exam AEP: NRC: 0969AA Attac U1 2 Co e Case N 521 10 CFR 50.55a 1 (a) (3) (i)
AEP:NRC:0969AA Attach U1 2 Code Case N 10 CFR 50.55a 2 491-1 (a) (3) (i)
AEP:NRC:0969AA Attac Ul 2 Co e Case N 509 10 CFR 50 '5a 3 (a) (3) (i)
AEP: NRC: 0969AA Attach U1 2 Code Case N 524 10 CFR 50.55a 4 (a) (3) (i) to AEP:NRC:0969BK Page 6 Item H "Zn the November 7, 1996, submittal, the licensee stated that approval to use Code Case N-498-1 was provided by letter dated July 5, 1995. Was NRC authorization to use this code case for the third 10-year interval obtained2 Zf so, provide a copy of the NRC Safety Evaluation Report authorizing the use of Code Case N-498-1 for the third 10-year ZSZ interval. If not, a new request for authorization must be submitted for the third 10-year ISZ interval."
Res onse to Item H The letter and the safety evaluation report dated July 5, 1995, to E. E. Fitzpatrick from the NRC, authorizes the use of code case N-498-1 for Cook Nuclear Plant, until the code case is published in a future revision of Regulatory Guide 1.147. The approval provided in this document is independent of the ISZ interval, and therefore, we do not need to supply a code relief request for the third 10-year ZSZ interval. The approval cover letter is included as attachment 2.
Item I "ASME Section XZ, paragraph IWB-2420 states that the sequence of component examinations established during the first interval shall be repeated during each successive inspection interval to the extent practical. In the NRC RAZ, the licensee was requested to provide a detailed technical discussion regarding why the successive examination requirement could not be met. Zn the September 10, 1996, submittal, the licensee stated that the requirements of IWB-2420 were met to the extent practical and provided a list of welds that did not meet this requirement.
ma)ority of the welds listed were shifted from the first period to the third period during the third interval. This deviation from ZWB-2420 does not meet the intent of the Code and has not been adequately )ustified. To find this acceptable, a request for relief (or proposed alternative) must be submitted for staff review with the appropriate technical and regulatory basis."
Res onse to Item I We have reviewed this issue in our submittal AEP:NRC:0969AW, and with our third 10-year interval ZSZ program plan contractor, and offer the following. Zn all but one case, the rescheduling of class 1 and 2 piping was necessitated by the change in requirements specified in ZWB-2412 and IWC-2412 that requires that. examinations in each "examination category" be completed in accordance with tables ZWB-2412-1 and IWC-2412-1. The second 10-year interval code requirements specified that the required examinations (collectively all categories) meet table IWB-2412-1 for successive inspection intervals. We could not meet both code requirements (IWB and IWC-2420, "Successive Inspections for Class 1 and 2 Components", and IWB and ZWC 2412) simultaneously, and therefore, chose to redistribute examinations with the least impact for the third 10-year ZSI interval to agree with tables IWB and IWC-2412-1 in the 1989 edition of the Section XZ Code.
The one case where the examination was moved from the second to the third period was to correspond with other examinations that required common scaffolding, examination equipment, and personnel.
Attachment 1 to AEP:NRC:0969BK Page 7 The purpose of this change was to reduce radiation dose exposure by consolidating examinations. We agree that this change in schedule will require a code relief request. This change will be re-evaluated for impact. If we choose to perform the examination as specified in the period in the third 10-year interval, we will submit a code relief request. If not, the examination will be changed to the schedule specified in the first and second 10-year intervals, and the plan will be changed accordingly. This action will be completed by July 31, 1998.
Item J "Zn accordance with 10 CFR 50.55a(g) (6) (ii) (A), all licensees must implement once, as part of the inservice inspection interval in effect on September 8, 1992, an augmented volumetric examination of the reactor pressure vessel (RPV) welds specified in Item B1.10 of Examination Category B-A of the 1989 Edith. on of the ASME Code,Section XI. Examination Category B-A, Items B1. 11 and Bl. 12 require volumetric examination of essentially 100% of the RPV circumferential and longitudinal shell welds, as defined by Figures ZWB-2500>>1 and -2, respectively. Essentially 100't, defined by 10 CFR 50.55a(g)(6)(ii) (A)(2), is greater than 90't of the exauiLnation volume of each weld. Licensees unable to satisfy the requirements of 10 CFR 50.55a(g)(6)(ii)(A) must propose an alternative to the examination requirements, which may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. Based on the table provided by the licensee in Request for Relief No. 1, there are six welds that could not be examined to the extent required by the Regulations. Have the Regulations been satisf ied by the submission of a proposed alternative? If provide the dates of the submittals and the subsequent NRC Safety so, Evaluation Report. If not, a proposed alternative must be submitted and authorized before the limited examinations for the RPV shell welds can be evaluated for the third 10-year ZSZ interval. "
Res onse to Item J Our letter AEP:NRC:0969AO, dated April 30, 1996, provided additional information to our original relief request, letter AEP:NRC:0960AI, dated July 28, 1995, based on the results of a best effort examination on unit 1. A revised estimate on unit 2 was also provided, based on unit 1 results and similarity of RPV design. Our letter AEP:NRC:0969AP, dated May 6, 1996, officially revised the original code relief request for unit 1.
NRC correspondence (TAC No. M93613), dated July 26, 1996, to Mr. E. E. Fitzpatrick, concluded that the unit 1 augmented RPV examination conducted during the second 10-year ISI interval, maximized examination coverage to the extent practical, and that imposing additional examinations would have resulted in considerable hardship without a compensating increase in the level of quality and safety. The safety evaluation report for the unit 1, second 10-year ISI interval augmented vessel examination alternative states that, "Based on the information submitted, the INEEL staf f concludes that, pursuant to 10 CFR 50.55 (g) (6) (ii) (A),
the licensee' proposed alternative to the augmented RPV examination requirements, i.e., examination of the accessible volume from the inside diameter surface provides an acceptable level of quality and safety." Cook Nuclear Plant has, therefore, satisfied the regulations for unit 1.
Attachment 1 to AEP:NRC:0969BK Page 8 Our letter AEP:NRC:0969AR, dated July 29, 1996, revises our original code relief request for the unit 2 augmented RPV examination, based on the unit 2 examinations conducted during the second 10-year ISI interval. The examination coverages obtained for unit 2 were considerably higher than those obtained for unit 1, mainly attributable to the vessel manufacturer's design differences. Since we have satisfactory closure to the regulations for unit 1, and the examination coverages for the unit 2 augmented RPV examination are higher, we anticipate that the proposed alternative for unit 2 will also satisfy the regulations.
Item K "Request for Relief No. 1 regards limited examinations for six RPV shell welds and four RPV nozzle-to-vessel welds for Unit 1, and four shell welds and four nozzle-to-vessel welds in Unit 2. A note with the Unit 2 table states that coverage estimates are based on the Unit 1 10-year ISI examination results. Considering that Unit 2 is in the third 10-year interval, actual coverage information from previous examinations should be available and provided. In addition, the coverage obtained appears low for some of the welds and has not been adequately )ustified. Provide a detailed technical discussion describing the limitations and the coverage obtained."
Res onse to Item K A code relief was submitted for unit 2 in our letter AEP:NRC:0969AR, dated July 29, 1996. As reported in the previous answer (Item J), the coverage for unit 2 was considerably higher.
The detailed technical discussion describing the limitations and coverages are provided in this code relief request.
ATTACHMENT 2 TO AEP:NRC:0969BK APPROVAL LETTER FOR CODE CASE N-498"1
July 5, 1995 Mr. E. E. Fitzpatrick, Vice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza Columbus, Ohio 43215
SUBJECT:
D.C. COOK, UNITS 1 AND 2, REQUESTING APPROVAL OF CODE CASE N-498-1 AS AN ALTERNATIVE TO THE REQUIRED HYDROSTATIC PRESSURE TEST (TAC NOS. M91783 AND M91784)
Dear Hr. Fitzpatrick:
The staff reviewed and evaluated the information provided by Indiana Michigan Power Company (IHPCo) in its letter dated February 27, 1995, related to relief from IHPCo's Inservice Inspection program.
IHPCo requested approval for use of the alternative rules of ASHE Section XI Code Case N-498-1, dated May 11, 1994, "Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems" for 10-year hydrostatic testing on Class 1, 2, and 3 systems. Based on the information submitted, IMPCo's alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) as compliance with the specified requirements of thisSection XI would result in hardship
'uality andor unusual difficulty without a compensating increase in the level of safety. IHPCo's alternative, use of Code Case N-498-1, is authorized until such time as this code case is published in a future revision of Regulatory Guide 1.147. At that time if IHPCo intends to continue to implement this code case, it is to follow all provisions in Code Case N-498-1 with limitations issued in Regulatory Guide 1.147, if any.
The staff's evaluation and conclusions are contained in the enclosed safety evaluation. Should you have any questions please contact John B. Hickman at (301) 415-3017.
Sincerely, Cynthia A. Carpenter, Acting Director Project Directorate III-I Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation