JPN-99-026, Forwards Relief Requests 18 & 19 to Jaf ISI Program Re Proposed Alternatives IAW 10CFR50.55a(a)(3)(i) & Relief from ASME Section XI Code Re Insp of RPV Vertical Shell & Shell to Flange Welds

Forwards Relief Requests 18 & 19 to Jaf ISI Program Re Proposed Alternatives IAW 10CFR50.55a(a)(3)(i) & Relief from ASME Section XI Code Re Insp of RPV Vertical Shell & Shell to Flange Welds

ML20210M802
Person / Time
Site:	FitzPatrick
Issue date:	08/05/1999
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JPN-99-026, NUDOCS 9908100183
Download: ML20210M802 (42)

Text

123 Main Street White Plains, New York 10601 914 681.6950 914 287 3309 (Fad NmW&%er a.-.. xa sei 4# Authority !7;7%'af0""

August 5,1999 JPN-99-026 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, DC 20555

Subject:

James A. FitzPatrick Nuclear Power Plant (Relief Requests #18 and #19)

Docket 50-333 Proposed Alternatives in Accordance with 10CFR50.55a(a)(3)(1) and Relief From ASME Section XI Code Regarding inspection of RPV Vertical Steell and Shell to Flance Welds

Reference:

1. JPN-99-025, NYPA Letter to NRC, " Proposed Altematives in Accordance with 10CFR50.55a(a)(3)(i) for Reactor Pressure Vessel Circumferential Shell Weld Examinations," dated August 5,1999.

2. NRC letter, S. Richards to O. Kingsley, " Evaluation of Second 10-Year Interval inservice Inspection Requests for Relief for Byron Nuclear Power Station, Units 1 and 2 (TAC No.s MA3982 and MA3933), dated March 26, 1999.

Dear Sir:

This letter transmits Relief Requests 18 and 19 to the James A. FitzPatrick's inservice inspection Program. 10CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for Reactor Pressure Vessel (RPV) shell welds specified in item B1.10 of Examination Category B-A,

" Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). Additionally, 10CFR50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement to submit information to the U.S. Nuclear Regulatory Commission to support the determination, and propose an attemative to the examination requirements that would provide an acceptable level of quality and safety. The Authority is unable to obtain essentially 100% of each vertical weld without disassembly or ,

removal of intemal interference, removal of permanently installed bio-shield, or modification of the inspection equipment. The Authority's intention is to review and evaluate methods to allow I 0i]

accessibility to greater than 90% of the vertical RPV shell welds in the belt-line region. The attemative plan (Relief Request 18) would allow time for review and evaluation of alternatives that could provide greater vertical weld examination coverage and ensure an acceptable level of safety and quality. The attemative plan, however, would exceed the time provisions, for 9908100183 990005 PDR ADOCK 05000333 p PDR

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completion of the augmented exams, specified in 50.55a(g)(6)(ii)(A)(2) and (3). Relief is requested for the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric

- examination of reactor pressure vessel vertical shell welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item B1.12, Longitudinal (Vertical) Shell Welds).

Permanent deferral of the circumferential shell welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, item B1.11, Circumferential Shell Welds), in accordance with Generic Letter 98-05, was requested in Reference 1.

In addition, pursuant to the provisions specified in 10 CFR 50.55a(g)(5)(iii) and in accordance with 10 CFR 50.55a(a)(3)(ii), this submittal requests relief (Relief Request 19) from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Subsection lWB,1989 Edition, for the volumetric examination requirement of the shell to flange weld during the first inspection period, as required by the ASME Code, Section XI,1989 Edition, IWB-2500, Table IWB-2500-1 for Examination Category B-A, item Number B1.30. Also, reliefis requested from IWB-2420(a) to allow James A.

FitzPatrick to defer the entire vessel shell to flange weld inspection to the end of the inspection interval. A similar relief request for the vessel shell to flange weld was approved by the NRC staff for the Byron Station (Reference 2). This relief request is submitted in conjunction with the <

relief request for the augmented inspection of the RPV vertical shell welds (Relief Request 18).

Attachment I contains the basis for Relief Request 18 and Attachment ll contains the basis for Relief Request 19. The Authority would like to use these reliefs in the upcoming refueling outage (RO 14) at James A. FitzPatrick, and therefore requests disposition of these relief requests prior to December 15,1999.

This letter contains no new commitments. If you have any questions, please contact Ms.

C. D. Faison.

1 Very truly urs,

. nubel Senior Vice President and Chief Nuclear Officer Attachments: As stated cc: See next page I

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cc: Regional Administrator U.S. Nuclear Regulatory Commission 475 Allendale Road j King of Prussia, PA 19406 Office of the Resident inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Guy Vissing, Project Manager l Project Directorate i Division of Licensing Project Management U.S. Nuclear Regulatory Commission Mail Stop 8C2

! Washington, DC 20555 l

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Attachment I to JPN-99-026

' Relief Request 18 Relief Request Regarding Augmented Inspection of Reactor Pressure Vessel Vertical Shell Wolds l

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NEW YORK POWER AUTHORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 L i

AttachmInt i JPN-99-026 Page 1 of 8 Relief Request 18 Relief Request from ASME Section XI Code Regarding Reactor Pressure Vessel Vertical Shell Wolds

Background:

10CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implernenting the examination requirements for Reactor Pressure Vessel (RPV) shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). As stated in 10CFR50.55a(g)(6)(ii)(A)(2) for the purposes of this augmented examination, essentially 100 percent as used in Table IWB-2500-1 means more than 90 percent of the examination volume for each weld. Additionally, 10CFR50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement to submit information to the U.S. Nuclear Regulatory Commission to support the determination, and propose an altemative to the examination requirements that would provide an acceptable level of quality and safety. The Authority is unable to obtain essentially 100% of each vertical weld without disassembly or removal of intemal interference, removal of permanently installed bio-shield, or modification of the inspection equipment. The Authority's intention is to review and evaluate methods to allow accessibility to greater than 90% of the vertical RPV shell welds in the belt-line region. The altemative plan would allow time for review and evaluation of altematives that could provide greater vertical weld examination coverage and ensure an acceptable level of safety and quality. The attemative plan, however, would exceed the time provisions, for completion of the augmented exams, specified in 50.55a(g)(6)(ii)(A)(2) and (3).

The purpose of this letter is to request approval, pursuant to provisions contained in 10CFR50.55a(a)(3)(i), of an attemative plan for performing the reactor pressure vessel (RPV) augmented examination requirements of 10CFR55a(g)(ii)(A)(2) for the James A. FitzPatrick Nuclear Power Plant. The Authority's attemative plan would defer the augmented exams to no later than refueling outage 16 (currently scheduled for 4* quarter 2004). The deferred inspection plari would include performance of RPV vertical weld examinations to the maximum extent possible, close to or exceed 90% coverage of the Vertical welds in the belt-line region and incidental coverage of 2-3 percent of the intersecting circumferential welds.

A. Component identification:

ISI Class 1, Code Category B-A, " Pressure Retaining Welds in Reactor Vessel", item B1.12,

" Longitudinal Shell Welds".

B. . Examination Requirements:

10CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for Reactor Pressure Vessel (RPV) shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). As stated in 10CFR50.55a(g)(6)(ii)(A)(2) for the purposes of this augmented examination, essentially 100 percent as used in Table IWB-2500-1 means more than 90 percent of the examination volume for each weld. Additionally,

Attichmtnt 1 JPN-99-026 Page 2 of 8 10CFR50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the

' augmented RPV shell weld examination requirement to submit information to the U.S. Nuclear Regulatory Commission to support the determination, and propose an attemative to the examination requirements that would provide an acceptable level of quality and safety.

C. Alternative To The Examination Requirements:

The attemative' plan would defer the augmented exams of vertical welds to be completed no later than on refueling outage 16 (3*-10 year ISI interval 9/97-9/06). Refueling outage 16 is currently scheduled for 4* quarter 2004. Unusual and a large number of RPV intemal obstructions / interference prevents achieving the " essentially 100%" coverage requirements of 10CFR50.55a(g)(6)(ii)(A)" Augmented Examination of Reactor Vessel". Calculated exam coverage obtained by three vendors using present tooling and technology is limited to a range of no more than 51% to 64% for all vertical welds and 33% to 52% for belt-line region welds only. While the low end range values can be achieved by the conventional, well experienced tooling, it would take a newer less proven tooling to presently achieve the higher end values but nevertheless, still lower than code requirement. Therefore, the Authority has recently pursued and encouraged as many as available domestic and foreign vendors to research and develop with "new generation" scanner tooling to take advantage of best technology that when developed, would increase coverage of the belt-line region vertical welds at the JAFNPP Reactor Vessel to close to or exceed 90%, including incidental coverage of 2% to 3% of the intersecting circumferential welds.

Four of the contacted vendors showed interest in the challenge. These vendors have already started the tooling conceptualization process. The newer scanner tooling will be smaller sized, thinner, and lighter weight equipment, some with flexible delivery systems, using phased array ultrasonic technique, shear wave (full vee coverage) methods to maximize scanning coverage, and for specific applications, using tooling that has been successfully used in the aeronautics industry. Some of the proposed new tooling (but not limited to these) is listed below:

• Standard UT transducers and/or phase array, with a flexible delivery mast, magnetic wheels and a telescopic arm for greater side reach.

• Unique lightweight / thinner scanner which can be carried by one person, remotely

- controlled, but only requiring a delivery connector for specific applications.

• A low profile flexible scanner used for aircraft fuselage and wing inspections (for OD use)

. A low profile phased array probe wand that could be used in the access panel region from the vessel 00, to increase belt-line coverage beyond areas previously not able to be accessed by present tooling methods used for RPV inspections (clearance less than 1" between vessel OD and insulation).

. . A remotely-operated manipulator presently used in PWR reactor vessels.

• A retrofitted suction cup scanner presently used for core shroud weld inspections.

UT scanning coverage for the belt-line vertical welds is estimated at equal to or greater than 80% for current conceptualized tooling. Enclosure 1 (4 drawings) provides a weld inspection coverage mapping and relevant notes for each of the four vendors.

The Authority's present implementation plan is listed below. This plan takes into consideration the development, fabncation mock-up testing, and qualification of the new tooling (s), it is estimated that the plan will take a minimum of 12-18 months after vendor selection. Also, the EPRI NDE Center is currently conducting experimental trials to determine the effectiveness of phased array techniques for the examination of BWR and PWR reactor pressure vessels.

Attachment i JPN-99-026 Page 3 of 8 Results of demonstrations from the OD surface indicate that improved coverage and detection capability can be achieved using sector scans with the ultrasonic beam focused at the clad to base metal interface. The beam is swept over a range of angles producing acceptable detection sensitivity over 10 to 12 inches of the clad to base-metal interface from a single probe position.

Demonstrations from the intemal surface are scheduled to start within the next two months and be completed by mid-2000. The demonstration would evaluate the effectiveness of " full Vee path" techniques as well as other limited access approaches. Commercialization of the intemal approach could potentially be initiated by the end of the year 2000 or 2001.

Imolementation Plan:

1. RO14 - Measure as-built gap between vessel OD and reflective insulation from the access openings to recirculation inlet / outlet nozzles. These measurements will help determine  ;

extent of tooling accessibility.

2. Cycle 14/15 - Select vendor consistent with most weld coverage and tooling reliability.

3. RO15 - Tentative start of wold examination during RO15 with the newly developed tooling. ,

4. RO16 - Start or complete the weld inspections.

D. Basis For Alternative Plan: j The Authority is unable to meet the greater than 90% coverage requirement for each weld due to intemal interference of the JAF reactor vessel components and the examination tooling equipment lower scan limitations. The attemative proposed in Reference 1 (and authorized by NRC in Reference 2) was to perform an augmented examination of the RPV welds in refueling outage 14 (currently scheduled for 4* quarter of 2000), and to evaluate methods for performing the examinations to the maximum extent possible from the inner diameter (10) and provide greater than 90% coverage of the vertical welds in the belt-line region. Accessibility studies by three different vendors would allow a maximum of 64% of the cumulative length of all vertical shell welds. This would have only allowed coverage of approximately 52% of the cumulative length of the belt-line region vertical welds. Further examination from the ID is not possible without disassembly of vessel intemal components. The attemative to defer the inspection until no later than RO16 will allow for development, testing and qualifying of a newer technology and smaller size tooling. The proposed plan will enable scanning of welds in confined areas not accessed by present tooling.

The industry basis document, BWRVIP-05, considered several issues related to BWR RPV integrity to provide a basis for eliminating the requirement to perform circumferential welds and the performance of only 50 % of the vertical RPV shell welds. These issues included fabrication practices, in-service inspection data, operational issues, degradation mechanics, and probabilistic fracture mechanics analysis results. As stated in the report "Results of the evaluation performed in this report clearly demonstrate the inherent safety and integrity of BWR reactor pressure vessels." The following basis for deferral uses a similar approach but utilizes  ;

more plant specific data to justify deferral of the required vertical examinations to RO-16.

Previous ShellWeld Examinations:

During the fabrication process of the RPV, oli of the shell welds were thoroughly examined j using several examination methods as required by the original construction code. Additionally, all of the shell welds received volumetric examination prior to initial plant operations, as prescribed by ASME Section XI pre-service inspection requirements.

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JPN-99-026 Page 4 of 8 A search of original construction " weld travelers" records identified among others, a F'eport of Ultrasonic Testing for Vessel Assembly dated 4/10/71, stating "UT of Pressure Boundary Welds.

No indications Reportable"; and a Shop Quality Control, inspection and Document Record document (by Stone and Webster), with a listing of performed and checked tests, dated 9/16/70.

All shell weld original radiographs have been digitized per latest EPRI guidelines. The digitized radiographs, for the vertical welds in the belt-line region, were reviewed by a JAFNPP QA Level ill inspector. The review identified minor inclusions / slag / porosity randomly oriented throughout the welds. These indications are considered minor with no safety significance. These radiographs were accepted during original vessel fabrication.-

Selected shell welds have received outer diameter (OD) volumetric examinations during the first and second interval in accordance with ASME Section XI in-service inspection requirements.

The OD examination totaled 28% of total vertical length of shell welds with 12% at belt-line vertical welds. Most of the intersecting welds,10 of 15, were inspected. Some welds only ,

received partial coverage (i.e., one sided examination coverage only). The OD examinations I resulted in only four recorded spot indications, with no measurable length or width. These indications were found acceptable for operation. A sketch of the previous OD exams, locations and results was included in transmittal to NRC via NYPA Letter (JAFP-98-0292), from Michael J. Colomb, to NRC, dated September 10,1998. ,

1 Industry Results of Past Examinations:

l Survey data compiled by W EPRI for the BWR fleet indicate that a total of 5,257 feet (63,084 inches) of vessel shell weld length was examined, or 36% of the total possible weld length of 24 units, resulting in only 16 indications exceeding the acceptance criteria of ASME Section XI, lWB-3500. All 16 indications were subsurface flaws shown to be acceptable by meeting the criteria of IWB-3600. The total length of the indications was 29.9 inches, which is just 0.05% of the tutal weld length examined.

Recent data provided by General Electric indicate that the reportable 16 indications (15 on  ;

circumferential wolds and one on vertical welds) have only been recorded on non-CE l' (Combustion Engineering) Reactor Pressure Vessels; JAF is a CE plant. All indications were determined to be construction related, and were evaluated and accepted for operation. No service related flaws were present.

Separate data by General Electric (Enclosure 2) show segregated ID (Inside Diameter) coverage data for vertical welds and are summarized below as follows:

BWR Fleet VerticalWelds: BWR-4 Vertical Welds:

Total Welds = 17,050" Total Welds = 8,350" Total Examined = 11,600" Total Examined = 6,000" Total Belt-line = 8,500" Total Belt-line = 4,100" Total Examined = 5,450" Total Examined = 2,700" It is clear from these survey results that a substantial amount of examinations have been performed which verify the integnty of BWR vessels, and that only a negligible number of construction related indications have been detected as a result of tiiese inspections with no

. service related defects r

Attichm:nt 1 JPN-99-026 Page 5 of 8 Fatiaue Life. Radiation Embrittlement and Stress Corrosion Crackina:

l l Some known degradation mechanisms may influence crack initiation and growth in the RPV low

alloy steel (A 533B1); fatigue life, radiation embrittlement and stress corrosion cracking (SCC).

W Fatigue life can be correlated with the water chemistry. Test results show a significant improvement in the number of cycles (load and unload) it takes to develop a 0.1 inch deep crack with conductivity approaching 0.06 S/cm when compared with 0.4-0.5 S/cm. A 0.1-inch crack is often considered to be the sizs of a crack of engineering significance. JAF conductivity values for the reactor coolant water have decreased substantially over the past years from a high of 0.31 S/cm in 1990 to a range of 0.066-0.067pS/cm for 1997 and 1998 respectively (Enclosure 3). For JAF, the number of startup and shutdown transient cycles is within overall limits of the Technical Specifications.

W Radiation embrittlement can be correlated with neutron fluence, which is highest at the belt-  !

line region. JAF has lower fluence values than the limiting plants analyzed by the NRC's evaluation of the BWRVIP-05 report. Below is a comparison between FitzPatrick versus the l limiting plants contained in the BWRVIP report for the projected exposure at 32 EFPY.

WJAF - peak fluence values at each belt-line weld (Enclosure 4):

1.06 E + 18 (welds W-3A and W-3C) 8.20 E + 17 (welds W-3B and W-4B) 1.10 E + 18 (weld W-4A) 7.40 E + 17 (weld W-4C)

W Plant 1 - peak fluence at belt-line weld:

! 6.76 E + 18 WPlant 2 - peak fluence at belt-line weld 1.50 E + 18 WThere are no cases of RPV damage in BWR plants that indicate susceptibility of the low-alloy steel base material to SCC during normal reactor operation. This is attributed to good BWR water chemistry. JAF conductivity, chloride and sulfate values have significantly improved over the last ten years and have consistently been within EPRI limits with average conductivity in 1998 being best in the GE BWR fleet. JAF has used hydrogen water chemistry and zine addition since 1989 and is planning to initiate noble metal chemical application in November 1999.

l Conditional Fai!ure Probability WA comparison of the probability of failure of vertical welds shows JAF lower than for plants, Clinton and Pilgrim, which are expected to bound the vertical weld concern for all BWR's, as follows:

j Plant Probability JAF 4.78 E - 03 Clinton 1.55 E - 02 Pilgrim 1.05 E - 02 L ]

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Attichm:nt i JPN-99-026 Page 6 of 8 Both plants, Clinton and Pilgrim have completed their ISI exams without relevant indications.

Although no credit may be taken to reduce the probability of weld failure for the inspected plants, the fact remains that the bounding BWR plants have not shown signs of service degradation for the RPV shell welds.

4 Low-Temoerature Over-Pressure Event (LTOP)

L At an industry meeting on August 8,1997, the NRC indicated that the potential for, and consequences of, non-design basis events not addressed in the BWRVIP-05 report should be considered. Later, in a Request for Additional Information (RAl) to the BWRVIP, the NRC requested that the BWRVIP evaluate the potential for a non-design basis cold over-pressure transient. The BWRVIP responded in a letter to NRC dated December 18,1997. The NRC also considered beyond design basis events, such as low temperature over-pressure (LTOP) events in their PFM analysis. In the BWRVIP response to the RAI the total probability of an occurrence of cold overpressure for BWR-4s was reported as 9E-4. It was concluded that it is highly unlikely that a BWR would experience a cold over-pressure transient. In fact, for a BWR to experience such an event would generally require several operator errors. The NRC described several types of events that could be precursors to BWR RPV cold over-pressure transients.

These were identified as precursors because no cold over-pressure event has occurred at a U.S. BWR. Also, the NRC identified one actual cold over-pressure event that occurred during shutdown at a non-U.S. BWR. This event apparently included several operational errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79*F to 88 F.

The high-pressure injection sources, administrative controls, and operator training regarding a i cold overpressure event for the FitzPatrick plant were addressed in Relief Request 17 (Reference 10), which requested permanent deferral of the RPV circumferential shell weld examinations. Based upon the information provided in Relief Request 17 it is considered that the probability of a low temperature over-pressure event at the FitzPatrick plant is less than or i equal to that used in the staff's July 30,1998, safety evaluation.

RPV IntemalDbstmotionnanterforence l Graph data provided by General Electric (Enclosure 5) shows lowest vertical weld coverage l achieved with the GERIS 2000 tooling for BWR CE Plants, at approximately 60% average for belt-line and non-belt-line welds. The low coverage is attributed to RPV intemal obstructions.  !

No domestic plant has removed these obstructions to increase weld coverage. l As noted before, unusual circumstances at JAF prevent an examination of " essentially 100%" of  ;

the length of all vertical welds. There is an excessive amount of vessel intemal obstructions / ,

interference, which can be seen in the NYPA prepared three-dimensional clearance diagram drawings (Enclosure 6, Sheets 1 to 3) and in the GE prepared Figure 1 based on actual accessibility study surveys for FitzPatrick " Projected Examination Coverage" (Enclosure 7). The

' internal obstructions / interference are listed below:

1- Jet pump assemblies, support plates and gussets restrict access to at least three vertical welds; o 2 Some of the core shroud repair tie-rods restrict access to at least two vertical welds. JAF has an installed 10 tie-rod system; 3 FW sparger and core spray piping restrict significant coverage to at least 34% of length over three vertical welds; 4- Guide rod at 180* restricts access to two vertical welds located at the same azimuth; h+

Att:chmsnt i JPN-99-026 Page 7 of 8 5 Steam dryer brackets obstruct local access for two welds; and 6 The Surveillance specimen holder would have to be temporarily relocated to access Weld W-3C, and subsequently placed back with same orientation. This would be controlled by a plant procedure and would be performed to aide in inspection coverage.

Removal of obstructions other than the surveillance specimen holder, would involve substantial risk, cost and person-rem exposure.

• Estimate for removing /re-installing two tie-rods and one guide rod is 4.106 person-rem exposure, approximately 460 duration hours, and 6,000 person-hours total. These estimates are based on actual hours used in RO-11 to install two tie-rods at JAF and on other plant specific historical data relevant for this work (i.e., Jet Pump Beam replacement). Duration hours are strictly Radiological Control Area accessed hours. It excludes duration hours spent for engineering, tooling / mock-up development, training, installation, etc.

Without the two tie-rods and one guide rod, net coverage for the belt-line area would be increased by approximately 20% to a 72% total, still short of the minimum 90% code requirement. There would be an increase of dose of approximately 4.1 REM at a total cost of over $750,000 due to material and labor and approximately one week of additional critical path time without a compensating increase in safety. Substantial risk is involved with the cutting and removal of parts with remote tooling with the potential of dropping cut material into the vessel but even riskier would be the material condition of the removed parts or components, probably requiring contingency material stand-by. Removal of other vessel intemals would risk permanent damage to the vessel inside wall, potential for loose parts (i.e., metal shavings that could cause fuel damage), it would involve significant amount of person-hours of direct labor with severe impact to the outage schedule, with a substantial increase in person-rem exposure, and without a compensating increase in safety.

Conclusion:

Deferral of the RPV shell weld exams to no later than refueling outage 16 (currently scheduled for 4* quarter 2004) will ensure a higher belt-line UT scan coverage by use of "new generation" tooling. This tooling will also be able to benefit the whole BWR fleet in terms of increased scan coverage. Based on the documentation in the BWRVIP-05 report, the risk-informed independent assessment performed by the NRC staff, the lower neutron fluence, the less challenging design and operational loading for BWRs, the quality of the original vessel fabrication, the lack of significant dagradation mechanisms, the results of the previous vessel examinations, and controls to prevent a cold ever-pressure event, the Authority believes a deferral in completing the inspection of the RPV si' ell welds until no later than RO-16 provides an acceptable level of  ;

quality and safety. '

E. Alternative Examinations

• The JAFNPP alternative pian would require the deferral of the augmented exams to no later

- than refueling outage 16 (currently scheduled for 4* quarter 2004). The Authority will work with interested vendors to encourage development of newer tooling / technology that will provide greater than 90% coverage of the belt-line region vertical welds, and incidental coverage of 2%

to 3% of the intersecting circumferential welds. The inspections can be done over two refueling outages (RO15/16) if new tooling is used.

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Att: chm:nt I JPN-99-026 Page 8 of 8 The newer smaller tooling will be able to access weld areas not previously accessible due to l

access limitation of the present tooling. To date vertical welds examined for the BWR fleet are 68% of the total vertical weld length and 64% of the belt-line. For BWR-4's vertical weld length

, . examined is 72% of the total vertical and 66% for belt-line. Segregating out the CE Vessels, average examination for all_ vertical welds, including belt-line, is approximately 60%. The data

. demonstrate the potential benefits for developing newer tooling / technology to augment {

coverage.

Further examination of the circumferential welds would depend on NRC review, resolution, and l

approval of the Authority's submitted Relief Request No.17 for a permanent deferral of weld examinations for these welds.

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References:

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1. NYPA Letter (JAFP-98 0316) to NRC, Proposed Altematives in Accordance with 10CFR50.55a(a)(3)(i) for RPV Shell Weld Exam, September 29,1998 (RR No.15).

2. NRC Letter to NYPA, JAFNPP-Authorization of Altemative Reactor Vessel Weld Exams (TAC No. MA 1954), November 3,1998.

3. BWRVIP-05 (EPRI TR-105697), BWR RPV Shell Weld Inspection Recommendations, September 1995.

! 4. BWRVIP response to NRC Request for Additional Information on BWRVIP-05, June 8, {

1998, transmitted to NRC on December 15,1998.

5. EPRI TR-109051, Environmentally-Assisted Fatigue Crack Growth initiation in Low Alloy  ;

L Steels, Final Report, November 1997 '

l l 6. EPRI TP-110168, NDE for Life Cycle Management Strategic Plan, May 1999. l 1

l 7. BWRVIP-60 (EPRI TR-108709), Evaluation of Stress Corrosion Crack Growth in Low l Alloy Steel Vessel Materials in the BWR Environment, Final Report, March 1999.

l 8' Memorandum RE-99-223, JAF Vessel Belt-Line Fluence, A.Ramachandran to P. Okas, l

[ April 22,1999.  !

l 9. Memorandum, Exposure Estimate for the Removal / Installation of Two Tie and one l Guide Rod at JAF, A. Stark to P. Okas, May 18,1999. .

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10. - JPN-99-025, NYPA Letter to NRC, " Proposed Altematives in Accordance with 10CFR50.55a(s)(3)(i) for Reactor Pressure Vessel Shell Weld Examinations," Relief Request #17, dated August 5,1999.

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Atta;hment I to JPN-99-026 1

l ENCLOSURE 1 RPV SHELL WELD INSPECTION COVERAGE (SH.1-4) !

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, NEWYORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT l DOCKET NO. 50-333 1

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• LIMITED EXPERIENCE WITH THI

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(SEE TABLE 2 WITH NOTES)

• WILL MAXIMIZE EXAM COVERA

• VERSATILITY OF 10/00 EQUIP g B

• FOR PREVIOUS 00 INSPECTIONS (SEE TABLE 1 WITH NOTES)

• LIGHT WEIGHT EQUIPENT .-

• FOR PROPOSED TOOLING (SEE TABLE 2 WITH NOTES)

• WILL PERFORM FULL MOCidlUP g

• ASSUES 67% CREDIT FOR FLA L VEE COVERAGE . MANAGEABLE LAYOOWN AREA ON

• ASSUMES 100% CREDIT FOR PHASED ARRAY COVERAGE wE -

R Scitout.E. SADVANT

• IT IS ESTIMATED THAT TE COMPLETION OF THESE TECHNIQUES + 00 SCANNING SUBJECT TO CLE N

A (DEVELOPENT. FABRICATION MOCK UP TESTING Af0 OVALIFICATION

+ WIll REQUIRE M T R ING

• DEMONSTRATIONS) WILLREQUIREAPPROXIMATELY 12-18 MONTHS A FROM CONTRACT AWARD.

+ WOULO REQUIRE CONSTANT

[ -

PERFORMING IVVII REFUEL AC Z r-e e

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ir -

h te i NI TABLE a - PREYIOLS OD INSPECYICpd actual tIELD LIMBfM 0.3. Utf1pKD n.exi [

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[ GISTIlUCTIp4 ComeENTS I. ) D '

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C. CAfasenmass (@)

9. saavleafts (@)

TOT AL SELfL1f( WERTILAL RELDbf EsitatATEDI 3 302 TOTAL vtRf tCAL. ICLDE EEsitMATEDI 2 402 -

(FROM 1D/00)

ET INCREASES PROBABILITY OF SUCCESS.

?S N37 REQUIRE EAVY LOAD ANALYSIS 3

- 99or/do73 42 ~

a* e jam,s A. PrTZPATnecK pa s sen some mLaw?

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W W G I N TION COVERAGE - VDOOR 3

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(ABOVE SHOWN COVERAGE BY EXISTING T d ADVANTAGES (VEPOOR 2 Y GEPERAL NOTES,

$

• ESTIMATED BELTLINE EXAM COVERAGE (WITH *EW TOOLING) = 80% - 90%.

• SUBJECT TO FEASIBILITY STUDY FOR A THIteER TOOL (LESS THAN I .5*

• RESPECTABLE EXAM COVERAGE.

• LIGHT WEIGHT EQUIPMENT -

THICM)IF FEA*,IBLE. IT WOULO ALLOW SCANNING OF THE ENTIRE LENGTH j B 0F WELD BEHIto T>E JET PUMP RESTRA!>ER BRACKET.

• PDI OUALIFIED j

• ESTIMATED TOTAL VERTICAL WELDS 2 80%

• PROVEN DATA ACQUISITION SY

$ e SMALL LAYOUT AREA ON REFUE

• COULD USE TWO SCAbNERS(EXI

= ScHEcutE,

• REOUIRES 12-16 MONTHS TO COMPLETE FEASIBILITY STUDIES. DESIGN, U FABRICATION. AND MOCM-UP TESTING.

• VERY HIGH DEVELORING COST

• REQUIRES FEASIBILITY STUDY

• EXISTING SCANbER HAS LIMIT g g (PERFORMED WELL IN TESTING

• NE DIMENSIONAL TOOLING.

+-W 6"

• ~~ 8 7 6 5

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I roral vretttAL urtos otw v0altus Estrotri > oox W ,, g 4- M... M...a 3 I

$ f air E 2 - OttatLED VERTICAL MEl.D COVERAGE - ESigaeATE WEP00R 2 (EMISTIDG TOOLING)

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• e01-902 C

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con ocxmais ecm.

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E73 C 4 ESTDenfED 5.0. cINEinns . voop 3 F NEW SCANNER IS FEASIBLE M acorrios cocaer ir sem enoposto voolse is ruastas.

S NOT REQUIRE HEAVY LOAD ANALYSIS.

8 NEW DESIGN) SIMULTANEOUSLY

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"$ e ESTIMATED BELTLINE APO TOTAL VERTICAL EXAM COVERAGE = 85-90%

(SEE TABLE 2 WITM TOTES)

• MAY BE USED BY SAME CREW PE

• RESPECTABLE EXAM COVERAGE 8 e PROPOSED NEC TOOLING - FLErIBLE MA$7 TITANIUM S.S. LIPES MAGNETIC WHEELS APO fELESCOPIC ARM WITH EITE R UT TRANSDUdER OR PHASE ARRAY. ARM MAS 20" LATERAL MOTION.

yto 5T _ SOEDULE,

• REQUIRES 12-18 MONTHS INCLUDING TOOLING DEVELOPE NT APO w OUALIFICATION, AFTER CONTRACT AWARD. DISADVANTAGES (VE M e E W TOOLING - WOULD REQUIRC g

$e& n A

b n a N 8 7 6 5

4 3 2 1 P

NNp i TABLE 4 - SWVIC1.85 CD ENSPECTION actual WELD LENGTee 0.0. UBf DED sAAess ( us i accuessi noeueton j maoutgap aos a s ievan l 94.TE ) l i MWW W S MSMNNVMWMNNVMMMN y;u g ( p 794, q,

4-f.s'WNWNNNMWCWWWWWWNNWJ v

- -3 644 < 0 7:o stop i d> i42'ropei 45'e01 44'

-a-N v:,'-3 w, ( 0 7:o stao (5 36' 36'

^ WC .L-d 646. W D 7;oa 00 (S t 37' 3'

W.gC 4 W- a 50 - t i I t t'o* . -

50 ( 2 b i to" -

(100El } vv - s e sa- -

N I W-ic iso -

M* t4, N .N

)

/**. W-2A vv -as 650 s to (5 } i d-to' -

1 / {i' ' W -at 650

y. 'Q s to (513P -

,,,[f ,,f,, W - Ma e6' W-18 So E E ' i i 5" l d vv-3c C ;N

'N Sc i w-4A lo (J)Se* O '

s 3 i ace' co.E.: w-4a Se

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W d i i s aoie

% W Ww-7 tR M L]

3 taste 2 - OET AILED WELD COVERAGE - ESTIMATE -

w .

g.n f WE1.D 1.D. IM[gg[ UDSTmET!Os COMPOf78 l H

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.& Q M.4W 4 .. .,.

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us nEcouMD CBE E80f DOICATI@s AT O' upstROINAL af@ Af 48' p(AR WAWE. NO PE.AS#ia8LE LDGTH OR WIDM i

he r 9m ryd; i yN ya arr W(mk 9ma r.r*

g2B HEcofEED FWD Sp0TS DOICATIQN Af e Lipsth0D4AL. fC SEAR $1m8LE t.ENDM QR WIDfM.

e thl_q . fair ssa w wtsE stoE a w e oro scir esE m Asc cxxE a cxpEmmaE. O tel AECGIEED A SDELE 9807 DOICAfD38 At r MA2 SCmN AT geID-MALL WITH ESIENTIALLV 90 LEND 1H Cil DgIW IsmLL (EPM.

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o=-e #4JCLEAR PCWER PLANT asse on.

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<=s j ENCLOSURE 3 "9 JAFNPP REACTOR COOLANT CHEMISTRY DATA ,

1 I

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

1 ENCLOSURE 3 JAMES A. FIT.ZPATRICK NUCLEAR POWER PLANT REACTOR COOLANT CHEMISTRY 1989 - 1998

(>10% power)

CONDUCTIVITY EPRI LIMIT 1 CHLORIDE EPRI LIMIT 1 SULFATE EPRI LIMIT 1 YEAR ( S/cm) (ppb) (ppb) (ppb) (ppb)

( S/cm) 1998 0.067 0.30 0.5 <5 2.5 <5 1997 0.066 0.30 0.5 <5 2.6 <5 1996 0.071 0.30 0.5 <5 3.4 <5 ,

1995 0.074 0.30 1 <5 4.2 <5 1994 0.095 0.30 2.3 <5 3.9 <5 1993 0.100 0.30 4.4 <20 6.5 <20 1992 Note: Plant was Shutdown Allof 1992 1991 0.112 0.30 1.8 <20 5.6 <20 1990 0.310 0.30 3.5 <20 17 <20 1989 0.271 0.30 4.1 <20 9.2 <20 Note: 1 - EPRI BWR CHEMISTRY GUIDELINES -1996 Revision. EPRI TR 103515-R1 (BWRVIP-29)

August 1999

Attachment I to JPN-99-026 ENCLOSURE 4 1 JAFNPP BELT-LINE WELD FLUENCE - 32 EFPY NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

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Attachment I to JPN-99-026 ENCLOSURE 5 GENERAL ELECTRIC DATA GERIS 2000 WELD COVERAGE GRAPHS l

l l

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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT  !

DOCKET NO. 50-333 DPR-59

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Attachment I to JPN-99-026 ENCLOSURE 6 JAFNPP RPV STRETCHOUT WITH INTERNAL OBSTRUCTION AND CLEARANCE DIAGRAMS (SH.1-3) i 1

4 1

l I

l l

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT

! DOCKET NO. 50-333 DPR-59 t ,

%~

30*GE' VC-F-1 EL,703-VV-1A STEAH DRYER BRACKET (34')

60*GE VC-1-2 EL.553" ,

~2^

FW SPARGER m p ,

l C 5" CORE SPRAY SURVEILLANCE L SPECIPEN BRACKET (30*) ,

15 gy

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, t i g SHROUD REPAIR TIE ROD (TYP) '

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m

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REACTOR VESSEL STRECHOUT O* TO 120*

M (FROM INSIDE LOOKING OUT)

~ _ _ _ _ . . . . . .

INSPECTION AREA CRITERIA RESULTS

, GUIDE NOD AT t 80* DIAW TER OF GUICE ROD 3.5**/*.76*

Ol&TANCE OF RG) FROM RPV WALL t.5**/ .25" Ott OF ROD ACk I 25 -I

1. 2l 5Wa#4%JS 48P' SIDES OF GUIDE ROOe

"#: !!E : ll: Fun's:

TIE RODS AT AS' APO 166* S E ROD RPV WALL - 4,s*

BOTTOM - 3.26' OT R 14 TO Y DIAE TER OF TIE RODS a.S* TO S*

NOTES:

1. ALL DIENSIONS ARE APPR0xIMATE

- ,,-- s

1. • VC-F- t EL .703*

pFT&Ql ck ,,,,,,, y 4 [,% s b~*

.0 en

,f [*';

VC-I-2 EL.SS3' h ,

[ SURVEILLANCE SPECIMEN

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\

VC-2-3 EL . 403*

s N -

3 l/2'$ CORE 4.s*

y SPRAY SPARGER M 74270R (LOCATED IN SWOUD) 4e 6 i*e- )

"' ~

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\

RISER BRACE (TYP)

) 4.-

~=

- 84.75*(TYP)

REACTOR VESSEL ~ - ll*(TYR)

WELD VV-4A (40') 3.25*

. 60 REACTOR VESSEL Vc 4-8H-l EL.103.el' ,

(

h YPICAL SECTION

~

03 .

RISER eRact -af w No.e 6 7 TIE RODS AT 45' ANO 165*

ET M NO.1049 R{Ag, PART PLAN NEW YORK POWER AUTHORITY SMOUD REPAIR TIE ROO(45' ) JAF - MJCLEAR POWER PLANT REACTOR VESSEL STRECHOUT AND JET PUMP RISER BRACE WITH INTERNAL OBSTRUCTIONS AND CLEARANCE DIAGRAMS SHEET nri I OF 3

150' GE f

120' GE VC-F-1 EL.703" GUIDE ROD VV-lB STEAM ORYER - -

BRACKET (146*) d

VC-1-2 EL.553" ,

~

i 6" FW SPARGER.

! (TYP)

~

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(TYP) 62 T  !!!

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VC-2-3 EL.403" h-TOP BELTLIE EL.365"

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$l L EL.0'-0" PLANT EL.285'-6" l20* TO 240*

(FROM INSIDE LOOKING OUT)

- eg6 8.d _

rua Notes src sur i r?hb g/3,QG s '

c. ., ,

SIEAM DRYER 60&#f,g6 BRACKET (214*) pp ,fo x

VC-F- i R . 703*

myy . _.

a m en e st

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s

\

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PEW YORK POWER AUTHORITY PART PLAN JAF - NUCLEAR POWER PLANT SW OUD REPAIR TIE ROD (I65*) REACTOR VESSEL STRECHOUT APO JET PUMP RISER BRACE WITH INTERNAL OBSTRUCTIONS AND CLEARANCE DIAGRAMS SHEET 2 OF 3

270* GE r

240*GE VC-F-5 EL.703" VV-1C 300*GE VC-1-2 EL.553" ,

6"$ FW SPARGERS (TYP) 5" CORE SPRAY C N SKET ver=r TEMPORARLLY' -

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255 3,g.M34N g TOP BELTLINE EL.365, W- T

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== 3

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o x V;gsSEL EL.O'-0" REACTOR VESSEL STRECHOUT

'-6" 240* TO O' N .

(FROM INSIDE LOOKING OUT)

Hili

a-FOR NOTES SEE SEET I goGE

--- STEAM DRYER N BRACKET (326') ho b ,f])

. z, c u

g. n ,9 V

p\c.;)p;3,3

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I

\ VC-4-flW-1 EL.103.et*

9908(Obl89 - O TYPICAL SECTION 5" CORE SPRAY PIPING APO 6* FEEDWATER SPARGERS PEW YORK POWER AUTHORITY JAF - NUCLEAR POWER PLANT REACTOR VESSEL STRECHOUT WITH INTERNAL OBSTRUCTIONS APO CLEARANCE DIAGRAMS SHEET J OF 3

=

l Attac'.nient I to JPN-99-026 ENCLOSURE 7 GENERAL ELECTRIC PREPARED PROJECTED EXAMINATION ,

COVERAGE FOR JAFNPP (SH.1-2) i l

1 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

p

l l

GENE B13-01869-081 Revision 0 l June 1997 1

1 1 TABLE 1 PROJECTED WELD EXAMINATION SHROUD REPAIR IN PLACE l

l Projected Projected Weld Exam Exam Weld Length Coverage Length ID (in) (%) (in) Obstmctions WC-F-1 686.3 94.7 649.9 Guide Rod @ 0 and 180 WC-1-2 686.3 94.7 649.9 Guide Rod @ 0* and 180

- WC-2-3 686.3 80.2 550.4 CS Downcom., Shroud Repair Sury. Spec. 1

. WC-3-4 686.3 18.3 125.6 CS Downcom., Shroud Repair  !

JP Riser Brace, Surv. Spec.

W-I A 150 89.2 133.8 Steam Dryer Bracket W-1B 150 39.5 134.3 Steam Dryer Bracket W-lC 150 100 150 None W-2A 150 76.2 114.3 FW Sparger, CS Piping W-2B 150 0 0 Guide Rod @ l80' l l W-2C 150 76.2 114.3 FW Sparger, CS Piping j W-3A 150 62.4 93.6 JP Riser Brace, Shroud Repair W-3B 150 0 0 Guide Rod @ 180' W-3C 150 41.9 62.9 JP Riser Brace, Sury. Spec.

- W-4 A 150 0 0 Shroud Repair, JP Riser W-4B 150 0 0 Shroud Repair, CS Downc.

W-4C 150 74.1 111.2 JP Riser, Manipulator Lower Scan '

Limit l TOTAL 4545 64 2890 i

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!! .t 9 REFERENCE GE DRAVING NO.197R564, y as CE DVG NO. E-233-230, E-233-231, E-233-234, C-233-234, E-233-240 PURPOSE BEV DATE PREPARED REVIEVED INIT APPROVED INIT

, 0 6-03-97 DHC/ CHM J. GILL!ARD m D. BRAGG gg _ _ - __

Attachment ll to JPN-99-026 Relief Request 19 Relief Request from ASME Section XI Code Regarding inspection of Shell to Flange Weld NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

Att: chm:nt ll JPN-99-026 Page 1 of 2 Relief Request 19 Relief Request from ASME Section XI Code Regarding inspection of Shell to Flange Weld A. Component identification:

The component for which relief is requested is the shell to flange weld, Examination Category B-A, item Number B1.30 of IWB-2500, Table IWB-2500-1, ASME Section XI,1989 Edition. (All future references to ASME Section XI requirements are taken from the 1989 Edition.)

B. Examination Requirements:

ASME Section XI, IWB-2420(a) requires the repetition of the sequence of component examinations which was established during the first inspection interval during successive intervals.

ASME Section XI, IWB-2500, Table IWB-2500-1, Examination Category B-A requires a volumetric examination of the shell to flange weld (item B1.30) during the first inspection period of the interval and during each successive inspection interval.

C. Relief Requested:

Relief is requested from performing the code required volumetric examinations on the shell to flange weld during the first period of the third ten year inspection interval. Rather, the entire shell to flange weld examination will be deferred to no later than the third period of the inspection interval. The examination will be performed in conjunction with the RPV vertical weld ,

inspections (Attachment 1).

D. Basis for Relief:

Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that the specified l requirements would result in hardship or unusual difficulty without a compensating increase in  :

the level of quality and safety.  !

The third ten year Inservice inspection (ISI) plan for the James A. FitzPatrick plant states that 50 percent of the shell to flange weld (Weld VC-F-1) will be inspected during the first inspection period and the remaining 50 percent will be inspected during the third period of the interval.

This inspection schedule complies with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-A, item B1.30.  : ,

This relief request defers examination of the entire shell to flange weld until the third inspection period of the interval, in conjunction with the RPV vertical weld inspections. However, deferral of the entire exam to the third period does not follow the sequence of examinations followed during the previous intervals. Therefore, relief is also requested from section IWB-2420(a) of ASME, Section XI. This deferral will allow the inspection of the shell to flange weld to coincide with the augmented inspections of the reactor pressure vessel (RPV) shell welds. Performing the inspection of the shell to flange weld during the same outage as the RPV shell welds affords the following advantages.

. The inspection of the shell to flange weld, in conjunction with the inspection of the RPV shell welds, reduces the radiation exposure to plant workers, if the shell to flange weld is inspected as currently scheduled (50 percent during the first inspection period and 50

1 Attachmcnt ll JPN-99-026 Page 2 of 2 percent during the third period of the inspection interval), these examinations will be completed manually and estimates of total radiation exposure are approximately 2 person-

- Rem. If the inspection of the shell to flange weld is deferred until the latter portion of the interval, then the inspection can he done in conjunction with the RPV shell welds. As stated in Attachment 1, a remote conhbd, automated tool will be used to examine the RPV shell  !

welds from inside the refueling ca sity. This automated tool can also be utilized to examine  !

the shell to flange weld if it is exarained during the same outage as the RPV shell welds.

The use of this tool to inspect the entire shell to flange weld, rather than manually inspecting the welds during two different outages, is expected to reduce exposure by approximately 2 person-Rem.

. Use of the automated equipment to examine the shell to flange weld and the other RPV .

shell welds improves the reliability and reproducibility of examinations, and therefore l

provides reasonable assurance of the structural integrity of the shell to flange weld. j e The inspection of the entire shell to flange weld during the latter portion of the interval, in conjunction with the RPV shell welds, reduces the outage time and cost associated with this inspection. The automated tool will be inside the reactor cavity to inspect the RPV shell welds and can then be utilized to inspect the shell to flange weld with a minimal incremental impact on cost and outage schedule. In contrast, a manual inspection of the shell to flange weld during the first and third inspection periods would incur higher costs and have a greater impact on outage schedules. Specifically, it is estimated that use of the automated tool, rather than inspecting manually, will result in a two to three shift (24 - 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />) outage savings in critical path time.

In conclusion, deferral of the examination of the reactor vessel shell to flange weld to the end of the inspection interval will provide an acceptable level of safety and quality. JAF's shell to flange weld was manually examined during the second ten year interval with fifty percent of the examination completed in 1990 and fifty percent completed in 1995. These exams did not reveal any rejectable indications. Therefore, based upon a lack of any rejectable indications, ,

deferral of the third interval exams until the third period does not constitute a safety hazard. l Therefore, requiring a partial inspection of the flange weld during RO 14 would constitute an 1 exposure, economic and schedule hardship without a compensating increase in quality or safety.

E. Alternative Examination (s):

JAF will perform the code required shell to flange exam using a remote controlled, automated inspection tool no later than the third period of the inspection period, in conjunction with the RPV shell welds.

F. Implementation Schedule: i This relief request, if approved, will be implemented during the current Inservice inspection (ISI) {

interval for JAF. The Authority would like to use this relief in the upcoming refueling outage (RO 14), and therefore requests disposition of this relief request prior to December 15,1999.

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