JPN-98-042, Provides Response to Request for Addl Info Re Inservice Insp Program Relief Requests.Commitments Made by Util in Ltr Listed on Attachment II

Provides Response to Request for Addl Info Re Inservice Insp Program Relief Requests.Commitments Made by Util in Ltr Listed on Attachment II

ML20153D827
Person / Time
Site:	FitzPatrick
Issue date:	09/21/1998
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JPN-98-042, JPN-98-42, TAC-MA0711, TAC-MA711, NUDOCS 9809250201
Download: ML20153D827 (13)

Text

, ,

123 Main Street White PWns. New York 10301 ,

914 681.6950 l 914 287.3309 (Fa4 4 NewYorkPower a - xa 8 i

& Authority s"aT *J*#1e",""

l September 21,1998 l JPN-98-042 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137  !

Washington, DC 20555  !

l

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Request for Additional Information Regarding Inservice insoection Proaram Relief Reauests (TAC No. MA0711) l

References:

1. JPN-98-002, NYPA Letter to NRC, " Third Inservice Inspection Interval I Inspection Program Plan", dated January 26,1998.

2. NRC Letter to NYPA, " Additional Information Regarding inservice Inspection Program Relief Requests (TAC No. MA0711)", dated August 20,1998.

3. JPN-97-031, NYPA Letter to NRC, ' Request for Relief from 1992 ASME Section XI Article IWA-2000 " Examination and Inspection"', dated October 6, 1997.

I

Dear Sir:

On January 26,1998 the New York Power Authority submitted the third ten year interval inspection program and relief requests for the James A. FitzPatrick Nuclear Power Plant (Reference 1). The NRC staff subsequently determined that additionalinformation was required to complete the review of the relief requests (Reference 2). Attachment I to this letter provides the requested information.

\

The information provided in the response to Item F, Request for Relief No. 9, supersedes Reference 3 in it's entirety.

g0 9809250201 980921 PDR ADOCK 05000333 G PDR

Commitments,made by the Authority in this letter are listed on Attachment 11. If you have any questions, please contact Ms. C. D. Faison.

Very truly yo a, 3 #

J A.

J.K ubel Senior Vice President &

Chief Nuclear Officer

. Attachments: as stated cc: Regional Adniinistrator U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident inspector's Office James A. FitzPatrick Nuclear Power Plant U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 130931 Mr. Joseph Williams, Project Manager Project Directorate I-1 Division of Reactor Projects - 1/11 U.S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, DC 20555 Mr. Michael T. Anderson INEEL Research Center 2151 North Boulevard P.O. Box 1625 Idaho Falls, Idaho 83415-2209

Attachment I to JPN-98-042 Response to Request for Additional Information Regarding inservice inspection Program Relief Requests (TAC No. MA0711)

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 l

l .

e 4

Attachment I to JPN-98-042 Page 1 of 9 i

I l Response to Request for Additional

information Regarding inservice inspection i Program Relief Requests (TAC No. MA0711) i I

4 Additional Information Reauired 4

A. Relief Request No.1 - The licensee has requested authorization to implement Code Case N-416-1, Attemative Pressure Test Requirement for Welded Repairs orInstallation of Replacement Items by Welding, Class 1, 2, and 3. The licensee has stated in the request that i

"the Authority does not intend to perform a surface examination on the root pass layer of butt and socket welds of the pressure retaining boundary of Class 3." The NRC staff has found the use of Code Case N-416-1 acceptable only if the licensee commits to performing a surface examination of the root pass weld layer for welded repairs and replacements in Class 3 systems, if surface examination is required by ASME Section Ill for final acceptance. Confirm that the condition above will be met.

Response -The Authority confirms and commits that a surface examination of the root pass weld layer for welded repairs and replacements in Class 3 systems, if required by ASME Section ill for final acceptance, wil! be performed. The requirement to perform root pass weld layer inspections has been incorporated in Administrative Procedure AP-19.02, " Post-Work Pressure Testing and Visual Inspection Requirements".

B. Relief Request No. 4 - The licensee has requested authorizatin to implem3nt Code Case N- l 509, Altemative Rules for Selection and Examination of Class 1, 2, and 3 Integrally Welded l

Attachments. The NRC has allowed the use of Code Case N-509 provided the licensee's i commit to supplement the examination requirements of N-509 to include a minimum of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among all Class 1,2, and 3 systems. Confirm that these conditicns will be met.

Response - The Authority confirms and commits to include a minimum of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among all Class 1,2, and 3 systems.

C. Requests for Relief Nos. 5-10 and 13 - The licensee must state the specific paragraph of the Regulations under which each proposed request for relief or alternative is submitted. A licensee may propose an alternative to CFR or Code requirements in accordance with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3)(ii). When submitting a proposed alternative the licensee must specify the appropriate regulatory basis. Under 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. Under 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 compensatory increase in the level of quality and t

safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components soley to provide access for examinations, and

Attachment I to JPN-98-042

. Page 2 of 9 development of sophisticated tooling that would result in only minimal increases in examination coverage. Financial factors should not be included as a hardship.

In accordance with 10 CFR 50.55a(g)(5)(iii), a licensee may submit a request for relief from ASME Code requirements. If a licensee determines that conformance with certain ASME Code requirements are impractical for its facility, the licensee shall notify the Commission and submit, as specified in $50.4, 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 replaced to enable inspection, the licensee should cite this part of CFR to support the criteria for evaluation. The NRC may, giving due consideration to the burden placed on the licensee, impose an alternative examination requirement.

Revise the subject requests for relief to include the required reference (s) and regulatory basis to ensure that the requests for relief are evaluated in accordance with the approporiate criteria.

Response - Request for Relief No. 5 - relief is requested pursuant to 10CFR50.55a(a)(3)(ii)

Request for Relief No. 6 - relief is requested pursuant to 10CR50.55a(a)(3)(l)

Request for Relief No. 7 - relief is requested pursuant to 10CFR50.55a(a)(3)(ii) ,

Roquest for Relief No. 8 - relief is requested pursuant to 10CFR50.55a(a)(3)(l)

Request for Relief No. 9 - relief is requested pursuant to 10CFR50.55a(a)(3)(l) and 10CFR50.55a(a)(3)(ii) e i

Request for Relief No.10 - reliefis requested pursuant to 10CFR50.55a(a)(3)(ii)

Request for Relief No.13 - relief is requested pursuant to 10 CFR 50.55a(a)(3)(l).

The subject relief requests will be revised to include the required reference (s) and regulatory I basis. I D. Request for Relief No. 7 - The licensee is seeking relief from the ultrasonic examination testing frequency requiremer"s specified in NUREG-0619, Table 2, and requests to eliminate the routine liquid penetrant examination testing, and the FW Nozzle Leakage Monitoring System (LMS), as required by NUREG-0619.

The Boiling Water Reactor Owners Group (BWROG), by letter dated October 30,1995, submitted report GE-NE-523-A71-0594, " Alternate BWR Feedwater Nozzle Inspection Requirements." This report proposed alternatives to the recommendations set forth in NUREG-0619, "BWR Feedwater Nozzle and Control Rod Return Drain Line Nozzle Cracking."

The staff performed an evaluation on the BWROG submittal and the results of the evaluation are contained in the Safety Evaluation (SE) dated June 5,1998. The staff found the proposed inspection program submitted by the BWROG acceptable subject to the conditions listed in Section 5.0 of the SE. However, as stated in Section 2.0 LICENSING REQUIREMENTS, of the

Attachment I to JPN-98-042 Page 3 of 9 SE, "the NRC has received commitments from licensees to follow the examination schedule for UT, visual testing (VT), and PT contained in Paragraph 4.3.2 of NUREG-0619." " Licensees incorporated these commitments into their ISI programs through technical specification changes, FSAR changes, or letters of commitment. Therefore, any changes to licensees' commitments are predicated on how the commitment were incorporated in the licensees' documents." It is unclear how the licensee committed to the NUREG-0619 examination requirements. Submittal of a request for relief may or may not be the correct means of changing the licensee's commitment. Provide additional information explaining how the original commitment was made to NUREG-0619 examination requirements.

Also, provide verification, that the proposed alternative is in compliance with the requirements identified in the " Safety Evaluation By The Office Of Nuclear Reactor Regulation BWR Owners' Group Report GE-NE-523-A71-0594 Alternate BWR Feedwater Nonle Inspection Requirements."

Response - The Authority withdraws this relief request and will proceed with a commitment change, using the guidance provided in the Nuclear Energy institute's " Guideline for Managing NRC Commitments", Revision 2, December 19,1995.

While not a required ASME Section XI inspection, the Feedwater nonle inspections required by NUREG-0619 will be maintained in the ISI Program as an augmented examination for the third ten-year interval for administrative purposes only. Elimination of the Leskage Monitoring System (LMS) will be evaluated in an Engineering Report in accordance with the Authority's Design Control Process and will address compliance with the NRC SER on the BWROG Topical Report GE-NE-523-A71-0594.

Ultrasonic exami.1ations were performed by automated methods during the 1990 refuel outage and manual methods during the 1996 refuel outage. Both inspection processes included a qualification process similar to that described in the NRC SER. In a letter to the NRC, "NUREG-0619 - Feedwater Nonle Inspections", dated March 25,1988, the Authority requested that the inspection schedule for the Feedwater nonles as required by NUREG-0619 be changed to every third refueling outage. The NRC in a letter to the Authority, " Relief From Augmented Inspection of Feedwater Nonle/Sparger", dated September 9,1988, approved this change. The next inspection is currently scheduled for 2002. In addition, the Authority will maintain the current visual inspection requirements and frequency required by NUREG-0619.

The fatigue evaluation required by the BWROG Topical Report GE-NE-523-A71-0594 / NRC SER is currently under review for future revision. A copy of the evaluation was provided to the NRC in the March 25,1988 letter.

E. Request for Relief No. 8 -The licensee has submitted this request for relief relating to the Control Rod Drive System. However, it is unclear as to the specific Code requirement (s) the licensee is seeking relief from. The licenses references Category C-H. However, Category C-H has multiple examination requirements. The licensee states In the Examination Requirements portion of the request that:

. . relief should be from the 10 year inservice / hydro test and Code Case N-498-1 that the

Attachment I to JPN-98-042 Page 4 of 9 inservice test will be done during a normally scheduled shutdown from plant operation.

In the Relief Requested portion of the Request for Relief the licensee states that:

Relief is requested from the requirements specificJ in IWC-2000, Table IWC-2500-1, Code Category C-H, All Pressure Retaining Components of the ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition for inservice pressure testing of the C.D System each inspection period. Also Code Case N-498-1 for the third period test.

It is unclear whether the licensee is seeking relief from the interval requirement, period requirement, or both. Identity the specific Code requirement (s) that will not be met (include category and item number (s)). Provide a specific description of the portion of the CRD system requiring relief. Include boundary / isometric drawings identifying the portion of the CRD system requiring relief.

Also, provide a basis for relief, explaining / justifying why the Code requirements cannot be met (see question C). The licensee has provided an alternative examination, explain as applicable:

1. how the alternative examination is equivalent to the Code-required examination and how it provides an acceptable level of quality and safety (10 CFR 50.55a(a)(3)(l)) or,

2. how the alternative examination provides reasonable assurance of operational readiness, and why compliance with the Code requirements would result in a hardship without a compensating increase in safety (10 CFR 50.55a(,a)(3)(ii)) or,

3. how the alternative examination provides reasonable assurancu of operational i

readiness, and why compliance' with the Code requirements is considered to be impractical (10 CFR 50.55a(g)(6)(l)).

, Response -The Authority,has determined that this request for relief will not be needed for the upcoming outage. However, the Authority will require this relief for future activities. A written response will be provided to this request for additional information within 90 days from the date of this letter.

. F. Request for Relief No. 9 - The licensee has requested relief from the requirements of ASME Section XI,1992 Edition, Article IWA-2000, as invoked by 10 CFR 50.55a(b)(2)(vi). The licensee has proposed as an alternative, to meet the requirements specified in ASME Section XI,1989 Edition without Addenda, Article IWA-2000.

Article IWA-2000 contains multiple requirements concerning examination and inspection. In order to perform an evaluation of this request for relief, it is necessary to know the specific requirements that will not be met. Provide the specific Code requirements in ASME Section XI, 1992 Edition, Article IWA-2000 that will not be met. Also, explain as applicable:

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1. how the alternative examination is equivalent to the Code-required examination and 1

1 .

Attachment I to JPN-98-042 Page 5 of 9 how it provides an acceptable level of quality and safety (10 CFR 50.55a(a)(3)(l)) or,

2. how the alternative examination provides reasonable assurance of operational readiness, and why compliance with the Code requirements would result in a hardship without a compensating increase in safety (10 CFR 50.55a(a)(3)(ii)) or,  ;

3. how the alternative examination provides reasonable assurance of operational readiness, and why compliance with the Code requirements is considered to be I impractical (10 CFR 50.55a(g)(6)(l)). I Response -This response supersedes a previous request for relief transmitted to the NRC under NYPA letter JPN-97-031, " Request for Relief from 1992 ASME Section XI Article IWA-l 2000 " Examination iand Inspection", dated October 6,1997. I Relief Reauested Relief is requested from the use of ASME Section XI,1992 Edition paragraphs IWA-2210

" Visual Examinations", IWA-2211 "VT-1 Examination", IWA-2213 "VT-3 Examination", and i Subsection IWA-2300, " Qualifications of Nondestructive Examination Personnel". This reFef is requested under 10CFR50.55a(a)(3)(l) and 10CFR50.55a(a)(3)(ii). The proposed alternative inspection described below provides an acceptable level of quality and safety and is essentially equivalent with regards to VT-1 and VT-3 examination results. Compliance with the 1992 ASME Code requirements for which relief is requested would result in a burden without a compensatory increase in quality or safety.

l i

Basis for Relief The JAF ISI and NDE programs for Class 1, 2 and 3 components and their supports are based on the 1989 Edition, No Addenda. These programs define the VT-1, VT-3 and supplemental

, examinations (surface and volumetric per IWE) including all factors necessary to qualify the examiners and assure the examinations are performed at the highest quality levels.

Conformance to the requirements of the 1992 edition would necessitate a separate NDE administrative program or the existing program would need to be revised to include a dual set of requirements. This would be necessary because the 1992 edition is not currently accepted by the NRC for general use, other than IWE / IWL. Procedurally the 1992 IWE requirements would not only encompass the personnel certification and VT-1, VT-3 procedures, but also will require any supplemental examinations performed (i.e. PT, MT, UT and RT) to meet the 1992 requirements when used. These administrative requirements would also need to be identified within the Site Administrative Procedures (AP's) and current ISI contras with vendors performing ISI activities, as applicable.

As stated above, the imposition of the 1992 Code requirements for Examination Methods (IWA-2200) and Qualifications of Nondestructive Examination Personnel (IWA-2300) would necessitate a separate (or complete update) to the NDE administration program including procedures, certifications and record keeping requirements. Separate programs or updates to meet the requirements of both the 1989 and 1992 edition of the code would merely parallel the

__._._. __ _._ _ _._ _._ _._._ _ _._ _ _ _ _____._ _~

Attachment I to JPN-98-042 Page 6 of 9 existing program and provide significant challenges to the Authority in ensuring compliance to the dual criteria with no substantial benefit to quality or safety. The existing NDE administration program meets the requirements of ASME Section XI 1989 Edition, No addenda, and is currently used on Class 1,2, and 3 safety related components including the reactor to provide an acceptable level of quality and safety. Maintaining two separate programs (or separate code requirements) for ISI NDE examinations and the associated personnel certifications is a significant burden that has no benefit to quality or safety.

Updating the personnel certification requirements to the 1992 Edition involves the use of CP-189 instead of the current SNT-TC-1A documents. Training is required to be provided by an NDE Instructor in all disciplines and the standard applies different administrative requirements than those identified within SNT-TC-1 A. These requirements would also need to be identified within Administrative Procedures (AP's) and current ISI contract requirements concerning vendor programs as applicable.

The application of separate programs for IWE examinations will complicate the work performed and challenge the site to ensure the separate set of requirements are met. The challenge of applying two sets of requirements will be compounded when vendors providing programs and services are also considered.

Prooosed Alternative The ASME Section XI 1989 Edition (no addenda) VT-1, VT-3 and Supplemental examination (surface and volumetric per IWE) requirements, for " Examination Methods" (lWA-2200) and

" Qualifications of Nondestructive Examination Personnel" (IWA-2300) will be applied to IWE required examinations. These are the same requirements that apply to all other safety related components in the ISI program and will be supplemented with indoctrination overview training by FitzPatrick's site Level lli (who has attended an EPRI IWE training course) covering the IWE examination requirements prior to personnel performing IWE VT-1 and VT-3 examinations. The l r' esults of examinations performed under this proposed alternative are considered equivalent to )

the results that would be obtained under the 1992 Edition of ASME Section IX. I Recently an EPRI training course was given to NYPA Engineering /ISI personnel and NYPA QC Level ll and lli examiners. This two-day course provided an overview and guidance for containment inspections, coating inspections and engineering evaluations. This ensures that  ;

QC and Engineering personnel have received specific training on containment inspections. '

This alternative will apply until the beginning of Refuel Outage 14 (RO14) currently scheduled to begin in October 2000.  ;

Justification for Grantina of Relief Requiring redundant administrative programs for IWE Visual and Supplemental examinations (surface and volumetric per IWE) in parallel with the existing programs for all other ISI components is a significant burden that has no benefit to quality and safety. 'nspections performed to the present ISI Program and Code of record are essentially equivalent to that required by the 1992 Edition with 1992 Addenda of the A$ME XI Code. Specific training has been provided to Authority Engineering /Isl personnel and QC personnel who oversee contract

1 Attachment I to JPN-98-042 Page 7 of 9

. inspectors and provide review of inspection data sheets as required for all ISI examinations.

9 G. Request for Relief No.10 -The licensee has requested relief from the requirements of IWA-5250(a)(2), concerning leakage at bolted connections. The licensee has proposed to include in the ISI program the requirement to perform VT-3 examination in-place on bolted connections 1

when leakage occurs. Evidence of degradation to the bolted connection or bolting shall require the bolt closest to the source of the leakage to be removed, visually examined VT-3 and evaluated to IWA-3100. When the removed bolt has evidence of degradation, all remaining

, bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with 1

IWA-3100.

Other licensees have proposed acceptable alternatives to the requirements of IWA-5250(a)(2)

, that have included a detailed and well-defined engineering evaluation of the bolting and the bolted connection when leakage is detected. The evaluation should, at a minimum, consider a the following factors: bolting materials, corrosiveness of process fluid leaking, leakage location, leakage history at connection or other system components, visual evidence of corrosion at connection (while connection is assembled), and service age of the bolting materials.

In order for the licensee's proposed alternative to be found acceptable, a specific leabe j evaluation procedure with the appropriate corrective actions to be taken if the evalu#.. is inconclusive or identifies bolting degradation is necessary. Discuss the intended action regarding this Request for Relief.

Response - An engineering evaluation will be performed to identify the most prudent course of action to determine the condition of the bolting and / or the root cause of the leak when leakage is detected at a bolted connection.

(

l A situation may be encountered that involves a leaking joint following the complete replacement I of bolting materials. The root cause of.the leak may be thermal expansion of the piping and

bolting matenic .iue to system heatup. In such a case, re-torquing the joint belting usually

] stops the leak. i<emcval of the joint bolting to evaluatl a V corrosion would be a hardship and

+

unwarranted if the bolting meterial is new or underws.iy ~ ^ namination prior to installation. l ASME Section XI Interpretation XI-1-92-01 recognizes this uuation as one in which the l requirements of lWA-5250(a)(2) do not apply. Other situations which could also result in

hardship include

a bolted connections which require a specific torque pattern, a

primary containment isolation valves which require stroke time testing and local leak

, rate testing following reassembly, t

components which would require extension disassembly to comply with the i requirements of IWA-5250(a)(2) due to equipment clearances.

i

. IWA-5250(a)(2) does not include other factors which may indicate the condition of the bolted i connection. Other factors that will be considered in the Authority's evaluation of leakage at i bolted connections will include, but not necessarily be limited to:

l

- , - , , . . . ,m. .- _. ,. - . - ,, _ . , -_ , .,,-

i Attachment I to JPN-98-042

. Page 8 of 9

. botting materials

. corrosiveness of the process fluid ,

a leakage location j .. history of leakage at the connection or other system components '

visual evidence of corrosion at the connection (while connection is assembled) I service age of the bolting materials.

l l Training on the use of the above leakage evaluation criteria, as well as the appropriate i

)

corrective actions to be taken if the evaluation is inconclusive or identifies bolting degradation, j will be provided to appropriate ISI, Operation Review Group, and Quality Assurance personnel '

I prior to the use of the proposed alternative. The leakage evaluation criteria and appropriate j corrective actions will be formally proceduralized within 6 months following completion of the ,

upcoming refuel outage. '

i l

H. Request for Rollef No.12 - The licensee has proposed the use of Code Case N-522, Pressure l Testing of Containment Penetration Piping,Section XI, Division 1. The licensee stated that the i leakage testing requirement results in unusual difficulties without a compensating increase in ,

the leve! of quality and safety.10 CFR 50.55a(a)(3)(ii) requires that the licensee demonstrate '

3 that comp!iance with the specified requirements would result in hardship or unusual difficulty

without a compensating increase in the level of quality and safety.

i j Provide a discussion describing / demonstrating the hardship and/or unusual difficulty associated i with the requirements, and why compliance with the examination requirements would result in j hardship or unusual difficulty without a compensating increase in the level of quality and safety.

i l Response -The use of Code Case N-522 is required for those systems that have been classified as ISI Class 2 but that cannot be pressure tested under ASME Code Case N-416-1 (see Relief Roquest No.1).

Many of these systems are air systems or are open to the atmosphere and the pressure test requirements of Code Case N-416-1 as stated in Relief Request No.1 do not apply. Two examples are provided below:

1. Portions of the Drywell, Inerting, CAD and Purge and Containment Differential Pressure System (system 27), which forms part of the containment, are classified as ISI Class 2 in accordance with ASME XI IWA-1320 of the ASME XI 1992 Edition through 1992 Addenda, if a piping replacement were to be completed between the containment and an AOV (e.g., AOV-112 and line number 24"-N-152A-8), use of Code Case N-416-1 would require augmented NDE (ASME Ill) and pressure testing per ASME XI 1992 Edition. A system leakage test per paragraph IWA-5000 would be required. This section of piping is normally open to containment atmosphere and the required system leakage test could not be performed.

In addition, the exemptions per IWC-5222 would require demonstration of an open flow path, which is not feasible for this air system. This would require having an air pressure source or flooding of containment to test this piping and also might require modification of the system.

l J

Attachment I to JPN-98-042 Page 9 of 9 in this case, performance of the augmented NDE per ASME Code Case N-416-1 and the 10 CFR 50 Appendix J test required by Code Case N-522 is equivalent and ensures an acceptable level of quality and safety.

2. The RHR containment spray piping (line number 10"-W20-302-12A/B) is an open ended system in containment and is classified as ISI Class 2. If a piping replacement were to be completed between the containment and one of the containment isolation valves (e.g.,10MOV-3-A), use of Code Case N-416-1 would require augmented NDE (ASME lil) and pressure testing per ASME XI 1992 Edition. A system leakage test per paragraph IWA-5000 would be required. This section of piping is normally open to containment atmosphere and the required system leakage test could not be performed without wetting the containment and electrical equipment inside the containment.

In addition, the exemptions per IWC-5222 would require demonstration of an open flow path. This can be done but requires additional manpower and radiation exposure to I line up the RHR system with an air test source. Personnel would have to verify flow by I checking the containment spray nozzles in containment in a radiation area thereby contributing to the increases in radiation exposure.

In this case, performance of the augmented NDE per ASME Code Case N-416-1 and the 10 CFR 50 Appendix J test required by Code Case N-522 is equivalent and ensure an acceptable level of quality and safety while reducing radiation exposure to plant staff.

Note that in all cases the augmented NDE under ASME Code Case N-416-1 shall be performed prior to the Code Case N-522 test if the code case is implemented during a modification, repair or replacement. This provides additional assurance of safety due to the imposition of ASME Ill NDE requirements for ISI Class 2 systems.

Thus the imposition of pressure testing under the Authority's Relief Request No.1 and ASME Code Case N-416-1 will result in unnecessary radiation exposure and potential modification of components to perform the pressure tot if used during the course of a plant modification, repair or replacement. The use of the augm i NDE under Code Case N-416-1, when required, and ASME Code Case N-522 for the p ;te tests provide an acceptable level of quality and safety.

Note that pressure test exemptions of ASME XI 1909 Edition, IWB-5000, lWC-5000 and IWC-5000 would still apply.

. = . . - - _ - _ _ ,

1

, . 1 e 1 Attachment 11 to JPN-98-042 '

Commitment List Commitment Commitment Due Date Number JPN-98-042-1 The subject relief requests will be revised to include the required reference (s) and regulatory basis. 6/30/99 l JPN-98-042-2 Provide a written response to request for additional 90 days from information on RR#8 in approximately 90 days from the date the date of this l of this letter. letter j l JPN-98-042-3 Proceduralize evaluations of leakage at bolted connections I and appropriate corrective action. Evaluations shallinclude 1 (but not necessarily limited to) bolting materials, 6 months  !

corrosiveness of the process fluid, leakage location and I U *I"9 leakage history, visual evidence of corrosion at the completion of next RO connection (while the connection is assembled) and service age of the bolting materials.

JPN-98-042-4 Provide training on the use of the above leakage evaluation criteria, as well as the appropriate corrective actions to be Prior to use of l taken if the evaluation is inconclusive or identifies bolting the proposed l degradation, to appropriate ISI, Operation Review Group, alternative i

and Quality Assurance personne. .

! I

, I 1

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59