ML20217F518

Responds to Re Concerns of Several Oswego County Residents Re Vertical Cracks in Welds of Core Shroud at Plant.Nrc Informs That Existing Shroud Does Not Create Undue Risk to Public Health & Safety

ML20217F518
Person / Time
Site:	Nine Mile Point
Issue date:	07/21/1997
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	James Mchugh HOUSE OF REP.
References
NUDOCS 9707240238
Download: ML20217F518 (5)

Text

. July 21, 1997 g * ;' e, j , ' c The. Honorable John M. McHugh '

. United Stat:s House of R:presentativ:s Washingt::n, D.C. 20515 ,

Dear Congressman McHugh:

I am responding to your letter of May 6,1997, that forwarded a letter dated April 22,1997, from your constituent, Ms. Claudia Smart, of Protect the Residents of Oswego County. Ms.

Smart asks several followup questions pertaining to information discussed at the April 14, 1997, public meeting at the Joint News Center in Fulton, New York. At the meeting, NRC representatives and members of the public discussed the weld cracks in the core shroud at Nine Mile Point Nuclear Station, Unit 1.

On April 23,1997, the same letter from Ms. Smart was sent by facsimile to Mr. Darl Hood who chaired the meeting for the NRC. I am enclosing a copy of Mr. Hood's reply of April 28,1997, and more detailed answers to Ms. Smart's questions. I am also enclosing a copy of the letter and safety evaluation that the NRC issued on May 8,1997, finding that (1) a proposed modification to the shroud tie rod assemblies to restore their intended functions is acceptable and (2) the existing shroud with cracks in vertical welds will continue to meet the codes and standards required by the Commission's regulations and, thus, is suitable for continued operation for a period of 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (about 14-1/2 months),

subject to certain conditions for reactor coolant chemistry control. Accordingly, the NRC is satiefied that operation with the existing shroud in the manner proposed for the prescribed period does not create an undue risk to public health and safety in the Oswego area.

I trust you will find this information useful, if I can be of further assistance, please contact me.

Sincerely,

@@ Signec by t L.J.Cimm L. Joseph Callan Executive Director W W made gg )

for Operations g 4( (

Enclosures:

WM

1. Additionalinformation addressing C. Smart's questions lhanks, nAan '

2. D. Hood letter to C. Smart dated April 28,1997 Og)Q '//99(97

3. NRC letter and safety evaluation ,

dated May 8,1997 lhlhb! !!!fblllhfflb DOCUMENT NAME: G:\NMP1\MCHUGH2.GT *See previous concurrence To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E " =

Copy with attachment / enclosure "N" = No copy 0FFICE PM POI 1 lE LA POI 1 l D:DDI71 l TECM ED* l l NRR:DE l WRR:DRPM* l NAME OMood/rst* Stittle* ADromerick(A)* W0llu J. Strotnider* C. Milltr DATE 06/27/97 06/27/97 06/27/97 05/15/97 05/19/97 05/21/97 0FFICE 0 DRPE ADPR:WRR l DINRR* l EDO* l OCA* g/ # l Chairman ]

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%,.... / July 21, 1997 The Honorable John M. McHugh United States House of Representatives Washington, D.C. 20515

Dear Congressman McHugh:

I am responding to your letter of May 6,1997, that forwarded a letter dated April 22,1997, fro',1 your constituent, Ms. Claudia Smart, of Protect the Residents of Oswego County. Ms.

F nart asks several followup questions pertaining to information discussed at the April 14, s997, public meeting at the Joint News Center in Fulton, New York. At the mating, NRC representatives and members of the public discussed the weld cracks in the core shroud at Nine Mile Point Nuclear Station, Unit 1.

On April 23,1997, the same letter from Ms. Smart was sent by facsimile to Mr. Darl Hood who chaired the meeting for the NRC. I am enclosing a copy of Mr. Hood's reply of April 28,1997, and more detailed answers to Ms. Smart's questions. I am also enclosing a copy of the letter and safety evaluation that the NRC issued on May 8,1997, finding that (1) a proposed modification to the shroud tie rod assemblies to restore their intended functions is acceptable and (2) the existing shroud with cracks in vertical welds will continue to meet the codas and standards required by the Commission's regulations and, thus, is suitable for continued operation for a period of 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (about 14-1/2 months),

subject to certain conditions for reactor coolant chemistry control. Accordingly, the NRC is satisfied that operation with the existing shroud in the manner proposed for the prescribed period does not create an undue risk to public health and safety in the Oswego area.

I trust you will find this information useful, if I ca 1 be of further assistance, please contact me.

Sincerely,

, J s ph Calian Exe ive Director for Operations

Enclosures:

1. Additionalinformation addressing C. Smart's questions

2. D. Hood letter to C. Smart

[g Ha 7 W N ' y ljvfg N 8- g 1P I

, dated April 28,1997

3. NRC letter and safety evaluation dated May 8,1997

AdditionalInformation Addressing Questions in Ms. Claudia Smart's Letter

1. The NRC stated at the meeting that it should consider the financial investment that Nimo has in this facility. Yet let's look at it. When it was built almost 30 years ago the expectation for its usefullife was 17 years. Haven't they received a bonus on their original investment?

The NRC has no position regarding "a bonus" on the original investment in Nine Mile Point Nuclear Station. As stated in Mr. Hood's letter dated April 28,1997, to Mrs. Smart, remarks regarding financialinvestment and consideration of the best interest of customers were made by representatives from Niagara Mohawk Power Corporation. NRC decisions regarding the operation of Nine Mile Point Nuclear Station will be based on the technical matters that relate to safe operation of the plant in accordance with the Commission's regulations.

2. When they came to Oswego County what benefits did the residents receive? Other than the jobs directly associated with the company itself, they have given this area some of the highest electric rates in the nation. They are using us as a cash cow but making it difficult to bring in other industry or commercial projects to lessen our dependency on their tax dollars. Is this to ensure a govemmental reluctance to oppose their policies? In fact, the County legislature forced a sales tax on the entire county on April 11,1996, because poor Nimo "needed it."

l The NRC has no comment regarding either benefits to the residents of Oswego County or the economic and fiscal policies of Oswego County and the State of New York as they relate to the continued operation of Nine Mile Point Nuclear Station.

3. The New York State (NYS) Health Department has been reluctant to do a causal study of the higher than usual cancer rates in Oswego County, is this because of what they are afraid to find?

The NRC has no comment on issues or questions conceming the New York State Health Department. These issues are best addressed to representatives from the New York State Health Department.

The NRC notes that the National Cancer institute conducted a study of the people living near U.S. nuclear power plants, inclu ing Oswego County, and found no ill-effects. (See S. Jablon et al., Cancer Populations Living Near Nuclear Facilities, NIH Publication 90-874, July 1990).

4. Has the NRC contacted the Canadian govemment and the communities around the lake that uses this as a water supply for their input about possibly jeopardizing their water supply?

No. The NRC ensures that activities associated with use of nuclear materials at Nine Mile Point Nuclear Station pose no undue risk to public health and safety.

To this end, licensees are required to control, monitor, assess, and report Enclosure 1

t 2-radiological effluents released to uncontrolled areas. The NRC receives and reviews annual radiological environmental operating reports from licensees describing the results of radiological environmental surveillance activities during the year and apassing the observed impacts of the plant operation on the environment. The NRC l also receives and reviews radioactive effluent release reports identifying the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit during the previous 6 months of operation. With respect to potential releases due to accidents, the NRC staff has concluded that reasonable assurance exists that operation of Unit 1 with the existing core shroud for a specified period will not endanger the health and safety of the public because minimum margins of safety required by Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code will continue to be met. Continued operation as authorized by the NRC's letter of May 8,1997, poses no undue risk to use of the lake is a water supply to Canadians or other communities around the lake.

5. Has the NRC contacted the tourist industry in the NYS, Canada, Penr.sylvania, Ohio, Michigan, Vermont, New Hampshire, and Maine about what type of financial risk this whole sector of the country is taking to help out one company?

No. The NRC ensures that activities associated with use of nuclear materials at Nine Mile Point Nuclear Station pose no undue risk to public health and safety.

The potential effects on tourism and other economic or financial risks, both local and regional, associated with operation of commercial nuclear power plants, are, in general, outside the purview of the NRC.

6. What would the impact be on the entire state if the economics of much of upstate is destroyed through a nuclear accident.

The NRC has no comment on the economic impacts to the State cf New York in the unlikely event of a nuclear accident. However, it is important to emphasize that the likelihood of an accident is very low, in part, through the " defense-in-depth" regulatory approach to design, construction, and operation of nuclear power plants. Multiple barriers to prevent "ission product release, diversity and redundancy in the design of safety systems, operator qualification and training programs, the testing and maintenance of plant systems, and the regulatory oversight provided by the NRC help assure there is no undue risk to the public.

7. - A rumor circulating in this area is that they need to keep running so that they can unload it on NYS. Does NYS really need to take on a white elephant when we are already such an over taxed state?

As previously stated, the NRC has no position regarding the economic and fiscal policies of Oswego County and the State of New York as they relate to Nine Mile Point Nuclear Station.

DisTRIBUT: M t Becket;Fi'e,(50-120)!w/ original incoming PUBLIC(w/ incoming)

EDO #G970350 EDO R/F L. Callan H. Thompson .

E. Jordan P. Norry J. Blaha S. Collins R. Zimmerm/F.

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, S. Burns 1- K. Cyr PDI-l R S. Varg/F a (w/ incoming)

B. Sheron A. Dromerick M. Thadani M. Boyle (e-mail only)

OGC OPA I NRR Mail Room (EDO #G970350 w/ incoming)

N. Olson C. Norsworthy D. Hood (w/ incoming)

S. Little H. Miller, Region !

L. Doerflein, Region !

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% April 28, 1997 -

Ms. Claudia Smart, Spekesperson protect the Residents of Oswego County 698 Silk Road '

Fulton, NY 13069 ,

Dear Ms. Is,

artt '

By letter of April 22 1997 you provided several written questions and 1997 consentsinresponselothe,NRC'smeetingwiththepubliconApril14,You,,

rosarding the Unit I core shroud at Nine Nile Point Nuclear Station.

indicate that your questions are directed to the NRC and elected officials. I appreciate thi; opportunity to respond to the questions that fall within the NRC's. jurisdiction.

You indicate that NRC stated at the meeting that the financial investment that Niagara Mohawk power Corporation has in th's facilit should be considered in whether Unit I should be restarted. As recall the remarks regarding andconsiderationofthebeskinterestofthe determininfiancialinvestment(NR customersh were not made b the C, but by Mr. 8. Ralph Sylvia, Executive Vice pres 4 dent Generation usiness Group and Chief Nuclear Officer of Nhgara Mohawk. The NRC's decision regarding Niagara Mohawk's request to operate with the modified tie rod assemblies for about 14-1/2 months before re-inspecting the shroud will be based upon technical matters that relate to safe operation in accordance with the Commission's regu14 tion.

You state that when Unit I was built almost 30 years ago, the expectation for its useful life was 17 years. Actually Unit I was designed and constructed for a useful life of at least 40 years. In accordance with the standard practice of the Atomic Energy Commission at the time, the initial operating f.e. provisional Operating license issued in 1969 was a provisional1969) to prov<license (de a,perit,J for observation Licensa and No. DPR-17 improvement. This wasdated-August converte 72,d to a full-term operating license (Facility Operating License No. DPR-63) in late 1974 with an effective expiration date of August it 2009. The issuance of a provisional license did not mean that the expected us,eful life was less than 40 years.

Your questions also reflect a concern for the potential tapact a major accident could have u>on use of the lake as a water supply and to the economy of New York and neigh >oring areas. The NRC is conducting a careful review of

. Niagara Mohawk's analyses and plans to assure that such operation will not have a significant adverse affect upon the probability or consequences of an accident, and we will not authorite resumption as proposed unless we find upon completion of our review that the proposed modifications are acceptable and meet the codes and standards required by the Commission's regulations.

Enclosure 2 M00 [

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1 .

~

1, Ms. Claudia smart. - 1- .

As some of your questions were discussed during the April 14 I as enclosing a copy of the NRC's meeting summary. ForadditIonaldetail1997, an meeting,d as noted in the enclosed summary, video tapes of the meeting are available for

- a fee from the NRC Public Document Room, the Gelman Buildi  !!!O L street, Washing W.dnrc.

pdr .

gov) ton, DC 20555 (phone 800-397-4209, fax 102-63413 ,

. I a preciate your interest in this matter and your attending the April 14, 199 , meeting

. Sincerely, Darl 5. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosure:

Meeting sumary 9

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. g NUCLEAR REQULATCRY GOMMISSION g, cAsm: stow, o.c. en n

% ,,g April 25, 1997 LICENSEE: Niagara Mohawk Power Corporation FACILITY: Nine Nile Point Nuclear Station, Unit No. 1 '

SUBJECT:

SumARY OF MEETINGS WITH LICENSEE AND PUBLIC ON APRIL 14, 1997, REGARDINGCORESHROUD(TACND.M98170) l Dn April 14, 1997, the NRC staff participated in a meeting with Niagara Mohawk l Power Corporation licensee end NMPC regarding the Unit I core shroud. The meeting,heldfrom(5:00to7:30p.m.),wasfollowedbyanNRCmeetingwiththe public from about 7:45 p.m. to 10:30 p.m. on the same subject. The meetings were located at the Joint News Center,10' Airport Road in Fulton, New York.

The agenda and a list of HRC attendees are given in Enclosure 1. Participants for HMPC included Messrs. R. Sylvia, R. Abbott, M. McCormick C. Terry, and N.

Rademacher. ContractorpersonnelincludedDr.R.SmithofAltranCorporation, Dr. M. Manahan, Sr. of MPH Technologies, and Dr. S. Ranganath of General Electric Nuclear Energy. Both meetings were well attended by state and local officials, members of the public, and local media.

The urpose of the meeting with HMPC was to review the letter to the NRC dated Apri 8, 1997. To introduce the technical discussions Mr. Hermann and Ms. Kavanagh of HRC provided background discussions, including related generic activitiesbytheBoilingWaterReactorVesselInternalsProject(BWRVIP),

descriptions and functions of the core shroud an explanation of tntergranular stresscorrosioncracking,andareviewofrelevantNMP1andindustry operating experience. Enclosure 2 presents the viewgraph slides and handouts l used by Mr. Hermann and Ms. Kavanagh.

In the April letter and meeting, NMPC discussed recent inspection findings of cracking in the heat affected zones of some vertical and horizontal shroud welds, and anomalies associated with the installation and design of the shroud tie rod assemblies. The licensee discussed root cause and corrective actions, reviewed design documentation and analyses regarding the acceptability of the

,as-found vertical weld cracking for a period of at least 10,600 operating hours, proposed a weld re-inspection schedule, and described actions taken to restore the tie rod assemblies to the as-designed condition. The licensee's corrective actions for the tie rod assemblies include a modification of the lower wedge retainer clip design, for which the licensee has requested NRC approval under 10 CFR 50.55a prior to restart. Details of the licensee's presentations are given in the April 8 letter and are not repeated here.

Enclosure 3 presents the viewgraph slides and handouts used by NMPC and its contractors. .

CONTACT: D. Hood, NRR 301-415-304g g8

~t-The meeting with the public. included introductions of local officials and members of various organizations by Ms. Barbara Brown, Legislator of Oswego l County. Numerous questions and expressions of concern for shroud integrity were received and discussed by the N E staff. Ms. C. Scott of Volney, New York, expressed a preference that the shroud should be replaced before restart and provided the NRC 4 signed petition to this end. Mr. P. Guenther stated his alief that cracks associated with vertical welds had extended into the 1 base metal of the shroud and felt that this condition represented an I unreviewed safety question. Dr. J. Johnsrud of Penns asked questions re regulatory policy.garding aging, operationalmanagerial history,ylvania Some individuals expressed roncerns for the present attitudes and Sta financial health of NMPC and concerns for the impact that a major accident could have on the local economy. Some employees and union members indicated their confidence in the licensee's analyses and their support for continued operation with shortened inspection intervals as proposed by the licensee.

' Asked abopt the restart plans, Mr. Sylvia replied that although the refueling efforts would probably be completed by the end of April, the unit will not be restarted until the NRC has completed its review and approved the modified shroud repair. Several people expressed appreciation for the meeting and requested that more meetings on issues of local concern be held in the future.

The meeting was video recorded and copies of the three VCR cassette tapes are available for a fee from the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washin 3343, e-mail pdrfnrc. gov).gton, DC 20555 (phone 800-397-4209, fax 202-634-att oo Darl S. N5od, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Agenda and NRC' attendees

2. NRC Slides by Mr. Hermann and Ms. Kavanagh *

3. NMPC and contractor slides cc w/encls: See next page

Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit No. 1 cc!

Mr. B. Ralph Sylvia Resident Inspector Executive Vice President U.S. Nuclear Regulatory Commission Generation Business Group P.O. Box 126 and Chief Nuclear Officer Lycoming, NY 13093 Niagara Mohawk Power Corporation Nuclear Learning Center Charles Donaldsor., Esquire 450 Lake Road Assistant Attorney General Oswego, NY 13126 New York Department of Law 120 Broadway Mr. Richard B. Abbott New York, NY 10271 Vice President and General Manager -

Nuclear Hr. Paul D. Eddy Niagara Mohawk Power Corporation State of New York Nine Mlle Point Nuclear Station Department of Public Service P.O. Box 63 Power Division, System Operations Lycoming, NY 13093 3 Empire State Plaza Albany, NY 12223 Mr. Martin J. McComick, Jr.

Vice President Mr. F. William Valentino, President Nuclear Safety Assessment New York State Energy, Research, and Support and Development Authority Niagara Mohawk Power Corporation Corporate Plaza West Nine Mile Point Nuclear Station 286 Washington Avenue Extension P.O. Box 63 Albany, NY 12203-6399 Lycoming, NY 13093 Mark J. Wetterhahn, Esquire Mr. Kim A. Dahlberg Winston & Strawn General Manager - Prpjects 1400 L Street, NW Niagara Mohawk Power Corporation Washington, DC 20005-3502 Nine Mlle Point Nuclear Station P.O. Box 63 Supervisor Lycoming, NY 13093 Town of Scriba Route B Box 382 Mr. Norman L. Rademacher Oswego, NY 13126 Plant Manager, Unit 1 Wine Mile Point Nuclear Station Gary D. Wilson, Esquire P.O. Box 63 Niagara Mohawk Power Corporation Lycoming, NY 13093 300 Erie Boulevard West Syracuse, NY 13202 Ms. Denise J. Wolniak Manager Licensing Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 13093 Regional Administrator, Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406

AGENDA April 14, 1997 Meeting on Nine Mile Point Nuclear Station Unit 1 Core Shroud

!. NRC SESSION WITH NIAEAAA N0 HAWK POWIR COP 70 RAT!0N (NMPC) 5:00 NRC Opening Remarks Darl Hood Purpose Introduction of Participants ,

5:05 Background on Core Shroud Issue Kerri Kavanagh Robert Hermann 4

5:15 NMPC Review of April 8 1997, Letter Martin McCormick to NRC and Supplementa Information et al.

l Introduction Core Shroud Stabilizer Assemblies (Tie Rods)

Core Shroud W61d Inspections and Evaluations Conclusions 6:30 NRC Questions /Coments 6:50 Break

!!. NRC SES$!0N WITH PUBLIC ON CORE SHROUD 7:00 NRC Opening Statements Darl Hood 7:10 Questions /Coments from Audience 9:30 NRC Closing Remarks Sini;h Bajwa Richard Wessman e

G DC45URE 1

NRC ATTENDEES Office of Nuclear Reactor Regulation, Rockville, MD:

Richard H. Wessman Chief Mechanical Engineering Branch Divis\onofEngineer<ng Singh5.Bajwa Acting Dirsctor Project Directorate 1-1 Darl 5. Hood , seniorProjectManager Project Directorate 1-1 Robert A. Hemann Senior level Advisor-Materials science Materials and Chemical Engineering Branch Division of Engineering Kerri A, Kavanagh Reactor Systems Engineer Reactor Systems Branch Division of Systams Safety and Analysis l William H. Koo senior Materials Engineer Materials and Chemical Engineering Branch Division cf Engineering Jai Raj N. Rajan Mechanical En inter Mechanical En ineering Branch i

Division of E gineering Region I, King of Prussia, PA:

Lawrence T. Doerflein Chief Project Branch 1 DivislonofReactorProjects Barry S. Norris Senior Resident inspector

, Nine Mile Point Nuclear Station Diane P. Serenei Senior Public Affairs Officer Public Affairs Staff

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MEETISU ON CORE SHROUD CRACKISU AT MNE MII E POINT USIT 1 .

April 14,1997

~

Robert A. Hermann, Senior Level Advisor g Division of Engineering y Office of Nuclear Reactor Regulation

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BACKGROUND 1

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o Core Shrotid Cracking

+ first detected in U.S. plants in 1993

-* GL 94-03, "Intergranular Stress Corrosion Cracking of Core Sieroods in Boiling Water Reactors," issued July 25,1994

-* all resg,.s evaluated and SERs issued l 0 CATEGORY C (22 Plants) l -* All Cc;c ory s C plants' core shrouds inspected per GL 94-03 (or initiated gw..g; c

repairs)'

[ -* 13 plants installed core shroud repairs (11 tie-rods and 2 clanops) i l

l c CATEGORY B (6 Plants)

-* All Category B plants core shroods :.wied per GL-94-03, met ASME structural .

l integrky criteria for at least one operating cycle

~

i -* No repairs l 0 CATEGORY A (8 Plants) ~

- Limited VT inspection perfoi...cd at 2 plants

' except Browns Ferry 1, which has been in an extended shutdown eeydetery Internetless Centerewe 2 april 2. 19 W i

. ST1TUS OF BWRVIP REPORT REViHWS o BWRVIP-03, Reactor Pressure Vessel and Internals Examination Guidelines o BWRVIP-05, BWR Vessel Shell Weld hpcetion Recommendations o BWRVIP-06, Safety Assessment of Reactor Internals o BWRVIP-07, Guidelines for Reinspection of BWR Core Shnrxis o BWRVIP-14, Evaluation of Crack Growth in BWR Stainless 2PV ,

Internals o BWRVIP-17, Roll / Expansion of Control Rod Drive and In-Core '

Instrument Penetrations in BWR Vessels Segutetery Inforestles tenterence april 2,199F

s

~

STATNS OF BWRVIP REPORT REViHWS (con't.) -

O BWRVIP-18, Core Spray Internals Inspection and Flaw Evaluation Guidelines o BWRVIP-19, Internal Core Spray Piping and Sparger Repair Design Criteria o BWRVIP-25, Core Plate Inspection and Flaw Evaluation Guideline o BWRVIP-26, Top Guide Inspection and Flaw Evaluation Guideline .

o BWRVIP-28, Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld Cracking -

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Core Shroud Stabilizer (Tie Rod Findings) . . . . . . . R. Corieri/G. Deaver Core Shroud Vertical .

Weld Assessment . . . . . . . . G. Inch / Dr. R. Smith .

Dr. M. Manahan/Dr.S. Ranganath Summary . . . . . . . . . . . . . . . . . M. McCormick Closing Remarks . . . . . . . . . . . R. B. Abbott 1

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EERf Meeting Purpose The purpose of the meeting is to:

e Discuss the details of recent inspections of the core shroud and stabilized assemblies { tie rods).

e Discuss the analyses supporting the 10C1'K50.55a submittal for proposed tie rod retainer clip l modification.

e Discuss analyses which demonstrate that the shroud.

and tie rods were operable and safe during the previous cycle.

e 4

- . . _ _ - _ . _ . . g

EEll Shroud Repair Background I

e The BWRVIP developed industry standardized shroud repair criteria which was approved by the NRC. The NMP1 repair was designed to meet standardized criteria.

e NMPC evaluated the industry experience related to core shroud horizontal weld cracking and concluded that the NMP1 core shroud could be susceptible to similar cracking..

e NMP1 took a pro-active approach with this issue and decided to install a shroud repair during the Spring 1995 refuel outage.

3

4

! lilElll Summary ofResults l

i j e The Unit 1 Core Shroud Stabilizers have been restored to

) the as-designed condition.

l e Redesigned lower spring wedge retainer clips have been j mstalled to improve tie rod operation.

e Steady state and transient thermal expansion has been .

! analyzed and proper function of the tie rods and their

{ components is assured. .

i i

4 4 .

.i EElll Summaryof Results l

i e A~ baseline inspection of the shroud vertical welds has been completed.

! e The as-found condition has been analyzed,

taking no credit for the integrity of the l horizontal welds and applying conservative j crack growth rates, and demonstrates the l continued structuralintegrity of the shroud.

l I

i

'l

i. - __ -

EEll! -

Recommended

~

Re-inspection Schedule -

~

e NMPC requests operation for at least 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (141/2 months) before re-l inspection. -

l e A safety evaluation, based on l conservatism with regard to analytical ,

l parameters, concludes no unreviewed -

! safety questions with regard to tie rod .

! repairs and vertical weld integrity.

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1 l EElll ShroudRepairDescription l e Shroud repair designed to structurally .

replace the shroud circumferential welds.

l e Four tie rod assemblies are placed around the shroud (azimuths 90 ,166 ,270 ,350 ).

l e Vertical restraint is provided by an alternate ~

load path between the top of shroud and l shroud support cone. .

e Horizontal restraint of the shroud is provided through the use of linear springs and limit stops.

. 10 e

,. - . . . . - , . - - s m. . .--, ..- . - .. -- .__ . _., ____ __

~

EEll! Spring 1997 ShroudRepair Inspection Plan e Prior to the 1997 refueling outage, NMPC submitted its shroud repair inspection plan to the

~

NRC for approval. ~

e The plan was in accordance with the BWRVIP-07 guidance.

e Visual inspection of all four stabilizer assembliss to:

- Verify the general mechanical and structural condition.

g

EElli .

As-Found Condition e Tie Rod Assemblies

- The tie rod assemblies were found to be in place and functional at the time of theireption with some anomalies.

e Tie Rod Nuts

- Allnutloddng devices wereintact.

- A torque died on the 270 der; tie rod nut identified a lack of the original installation mechanicalpreload.

- The torque ad de:mmined that an axial h.nce in the tie axi assembly on the order of 0.06" existed.

o kwsSpring Wedge and Latch

- 90*: latch Latured and k,we wedge re-positioned down on wedge guide .

- 166*: latch and k,ws spring wedge normal

- 270*: latch potentially damaged and lower spring wedge normal

- 350*: latch damaged and k,ws spring wedge =1/8" below normal position U

O e

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e ERll

~

AdditionalInspections e Based on the as-found conditions, additionalinspections were determined to be required.

• A comprehensive procedure was developed to interogate the condition of each of the tie rod assemblies.

e Remote operated underwater tooling and inspection equipment was designed and fabricated to implement the procedure.

e The procedure was also intended to obtain data to validate the root cause theories associated with the degraded latches and the lack of preload in the 270* tie rod.

e As a result it was determined that the tie rod assemblies at the 90,166 and 350 degree azimuths also had some amount of axial clearance which ranged from 0.054" to 0.151".

U

>ime Elli .

Root Cause

~

e The evdation of tk as-found condition shows that both the latch failum and the loss of de rod preload wem minted.

e The design of the lower spring contact implicitly assuined that the lower spring contact would slide along the Reactor Pmssum Vessel l (RPV) wall.

e Them wem two conditions causing diffemntial movement that wem not expected:

l

- The lower support assemblies wem able to shift up the sluvud cone toward the shroud due to original installation clearances between l

the toggle bolts and the cone holes. The impact of the clearances was not mcognized.

- Diffemntial motion could also be caused by the deflection of Fu C -

spring under tie rod load for heat up. This could also cause stresses in the latch, although somewhat less than in the pmvious case.

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EElli Consequence of Tie Rod Anomalies

~During the Past Operating Cycle e No plant operational anomalies noted during the past cycle.

e All plant operating design cases evaluated.

All stresses are within ASME Code limits.

Bypass leakage does not affect plant operation or safety functions.

Core cooling operability unaffected.

- Safe shutdown capability unaffected. .

e Flow induced vibration did not occur.

e As found shroud horizontal weld conditions were safe without tie rod repair in place.

e

Conclusion:

no safety concern; no adverse affect on tie rod '

repair hardware. .

b

EEll . Corrective Actions 1

e Removed clearance between the lower support toggle bolts and the shroud side of the cone holes.

e Re-torqued the tie rods to their original

~

designinstallation torque.

~

e Installed new modified latches which are more tolerant of differential vertical

displacement.

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EElli Latch Design Objective o Supportlower wedge dead weight

! loads e Accommodate potentialvertical

! displacem.ents betweenlower wedge

! and lower spring -

o Prevent release of the lower wedge and .

loss oflower spring contact a

. -l EEll Potential Sliding Cases- ]

l l

e Only sliding at vessel wall / lower wec ge interface.

) e Only sliding at lower wedge / lower spring interface.

o Combination of the above. .

I i

19

EEll Lower Wedge / Lower Spring Sliding Scenario Event Surface Assumed to Latch Displacement Slide (inches)

Initialheatup and SpringInterface 0.042 hydrotest Remainder of heatup SpringInterface 0.090 to fullpower operation

~

Loss of Feedwater SpringInterface 0.132 Heating

O EEll Combined Sliding Scenario

~- Event Surface Assumed to Latch Displacewait Slide (inches)

Initialheatup and SpringInterface 0.042 hydrotest Remainder of heatup SpringInterface 0.090 to full power operation Cooldown to VesselInterface 0.115*

Ambient (70*F)

Heatup to Full Power SpringInterface 0.182 Operation less of Feedwater SpringInterface 0.224 .

Heating

• Maximum displacementis limited by the amount of travel down the 5 degree angle of thespring.

11

EElli Stress Analysis Results e

Sliding Plant Displacement CalculatedStress Condition Operating Condition ADowableStress Sliding only at Normal .090"* 33 %

lower Operation wedge / lower spring interface IDFWH .132" 43 %

Operation Sliding at both Normal .182" 60 %

interfaces Operation LOFWH .224" 73 %

Operation .

• 'Ihe stress results reported are for a 0.100" displacement,,

whichis conservative.

~~

EElll Stress Corrosion Evaluation ]i e Stress Rule Index Methodology utilized.

e For probable sliding case, stress corrosion will not occur for remaining l

life of plant.

e For worst case sliding, stress corrosion j will not occur in the next operating .

cycle.

S

_ _ _ _ L _ _ _ _ __ _ _ _ _ . . ___ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _

Mll Comparison ofLatch Designs l e The improved latch design stresses are 8 to 12 times lower than the original design

- membrane + bending ~

ratio = 8.6 .

- membrane + bending + peak ratio = 12.8 .

~

e No permanent deformation in new latch design even under worst case conditions 24

e EEE Core Shroud Vertical Weld Inspection and Evaluation

• l G. B. Inch NMPC Engineering

EEll Expanded Shroud Inspection Goals e Baseline shroud vertical and horizontal welds.

o Obtain comparison between vertical and horizontal IGSCC cracking patterns

- Horizontal cracking at the H4 location consistent with other BWR-2 cracking and NMP1 H4 analysis predictions.

~

o Obtain H8 UT re-inspection data

- Re-inspection of H8 confirms no significant IGSCC cracking -

which could impact core shroud support function. .

Structural capability assured based on inspection

- Sample inspection of H9 with EVT shows no indications.

a .

EElli Expanded Shroud Inspection Goals e Determine actual core shroud structural margin presentin horizontalwelds.

e The tie rod installation assumed horizontal welds not present.

- Inspection shows significant margin.

e Based on structural capability of H4 and H5 ~

establish the margins associated with vertical weld cracking

EEll Additional Assessment Initiatives e Obtain comprehensive material condition assessment using all available inspection tools (enhanced visual examination / ultrasonic volumetric examination).

e Assessment of the shroud vertical cracking performed by severa1 independent IGSCC experts to compare cracking to other industry shroud cracking.

e Advancedcomputermodelingof thefabrication process to better define the most probable residual stress state which could explain OD dominant cracking.

o Obtain metallurgical sampling of vertical welds (two boat samples).

a

~

EEll Additional Assessment

~

Initiatives untinuce e More refined analysis of the vertical weld crackingis expected to increase the inspection interval to one operating cycle.

e Re-inspection of vertical weld cracking most probably will show deeper cracking is arrested.

e The industry has never seen through wall cracking.

s

O EEll Basisfor Vertical Weld Analysis Loads e Welding residual stresses and the welding l process fitup related induced stress in the weld create built in stresses which drive IGSCC cracking.

~

o Pressure stress dominates fracture. .

e The pressure stresses are defined by .

reactor internal pressure difference calculations.

e

~

EEll Root Cause ofCracking e Vertical weld cracking is IGSCC.

e Potential for irradiation enhanced material sensitization in the HAZ, which, coupled with enhanced stress relaxation, can affect crack growth.

e All findings show that IGSCC consistent with basis for BWRVIP established and NRC approved methods for analyzing core shroud cracking and establishing re-inspection requirements.

e Conclusion is that the BWRVIP core shroud inspection and evaluation guidance applies to the NMP1 vertical weld cracking.

Ji

EElll Ti1ermal Hydraulics Assessment o Potentialverticalweld through-wallcracking could result in (negligible) diverted core flow.

e Anticipated transients (potentially increased carryunder has favorable effect on thermal limits).

eLOCA

- Potential leakage has no impact on - a spray .

flow. .

- Core cooling is assured through core spray.

M .

e **

EEll Vertical Weld V9 and V10

~

Crack Growth Margins e Uncertainty associated with variables like stress intensity, neutron fluence are basis for bounding crack growth rates  !

of Se-5 inches / hour.

o Detailed crack growth analyses which account for all the above variables .

define V9 and V10 crack site specific ~

growth rates which demonstrate that Se-5 inches / hour is conservative.

n

l ERR Evaluation o Cracking in Vertical Welds i

i 1

i Dr. R. Smith Altran 4

i al .

.-_ - --------- --_- --- - - - s

dP l Elli .

Purpose

-, l l

e Careful examination of cracking patterns and other information.

e Develop a plausible explanation of what happened.

! EEll -

l Cracking Patterns Provide Evidence o the Reasons or .

Cracking

..}

5 9

EElli Observations @ V9 and V10 e Citacking characteristics typical for

~

shroud e Cracking predominantly on OD e Cracking remains axial predominantly in the weld HAZ e Cracking density favors one plate e Cracking deeper at top /more shallow towards bottom

O ' hw i

EEll -

Considerations o Parameters for IGSCC are wellknown e Welding and fabrication practices alter residual stresses i

N .

. . i

4 9

EEll Residual Stress Sources 4

e Welding e Surface metalworking e Fabrication and fitup e

EEll Welding Residual Stresses

~

o Extensively studied e Predicted by FEM e Confirmed by measurements .

su .

\ _

EEIII Through-Wall Stress Pattern e Depends on heat input and weld sequence o High OD stress predicted for low heat input welds e Fitup shaping adjustments (diameter squeeze) e Combination produces a stress pattern thatis consistent with the cracking observations

~

EEll Time Dependent ~

Irradiation Effects ,

e Increase electrochemical potential (ECF) e Enhances material susceptibility .

i e Reduces residual stresses by a creep l mechanism '

l 4

EEll -

Core Flux Pattern Suggests a Possible Reasonfor Crack Depth Diferences Top to Bottom a

b dl

ERII -

Conclusion

• e Shroud fabrication practices pro vide a plausible explanation of vertical weld cracking observations e Time dependent irradiation effects can help explain crack depth profiles top to

~

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EEll Qualitative Characterization o ~

V9 and V10 Cracking e Tlie cracking is almost exclusively on the OD side of the weld within the HAZ e Most of the cracks run longer in the axial dirch and are cwe.cci -d to a short horizontal crack segment a h axial cracks (driven by hoop stus) are deepest near the H4 weld where the fast neutron flux is highest e Both the left and right sides of V10 are cracked e h left side of V9 is cracked with little cracking on the right side

. e h depth of cracking correlates with fast (E > 1 MeV) neutron fluence

Conclusion:

h evidence suggests that the cocking niedmism is -

irradiation enhanced - intergianular stress corrosion cracking (IE-IGSCC).

L

~

EElll Cause o Predominant OD i

Cracking at V9/V10

Nretrospective analysis of the cracking ohred has been i ,

performed to obtain an in-depth understanding of the '

! fabrication processes which contributed to the observed '

I cracking behavior. The approach involved the following-e weldingsimulations(WELD 3) i e shoploadsimulations(ALT 3D) e weldrepairsimulations(ALT 3D?

Conclusion:

It can be demonstrated that a combination oflow heat input and a diametral squeeze (dead weight and/or jacking) produce a stress field which would explain the cracking behavior. .

& ~

l e e

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EEll StructuralMargin Assessment o V9 or Continued Operation Model Desuiption .

e Cmdit was not taken forcrack arrest e LEFM, EPFM, and limit load calculations wem performed e Crack growth rates were calculated using GE fluence dependent model e Initial crack depths which bound the measured depths were used e Variation of fluence through the wall was modeled using plant-sgeine fluxes e K ys. a data wem calculated using finite element weeds for a ~

representative sirs field e Cracks initiate under axial strer grow 0.3 to 0.5 inches deep, and then grow under hoop stress e Cases with, and without, credit for integrity of the H4 and H5 welds were analyzed 49

~

EEE StructuralMargin Assessment of V9for Continued Operation 1

Conclusions l

l

^

o The bounding 5 x 10-5 in/hr crack growth rate i

is conservative o The analyses show that safe operation can be ensured for at least an additional 2 years of

, hot operation L

! Q.

[

d>

EElll Structural Evaluation of the Shroud Vertical WeldIndications 1

1 Dr. S. Ranganath GENE -

I

EEll! Impact of Vertical Weld Cracking on the Tie Rod Func~ tion e Tie rod repair design basis does not require vertical welds to be crack free

- Any flaws should be within the allowable size

- No credit for horizontalwelds -

o Structural analysis of NMP1 vertical weld indications based on separate stand-alone cylindrical model

- Acceptability demonstrated assuming horizontal welds to be fully cracked

- No adverse effect on the tie rod repair function Vertical Weld Cracking does not feed to Violation ofTie RodRepairDesign Basis al -

.i EElll Efect ofTie Rod Loading on the Vertical Weld Cracking e Xnalysis performed to determined whether tie rod loading can cause stresses which could cause crack growth in the vertical welds

- 3D finite elementmodeling e Results confirm that the stresses due to tie -

rod loading are negligibly small

~

, I Tie RodRepair has noImpact on i

Vertical Weld Cracking \

w

! EEll Structural Evaluation

~

~ . e Two types of evaluations performed; with common features:

- Fracture and LimitLoad considered

- ASME Code safety factorsincluded i

e Screening Criteria Approach I

- Assumes through wallcracking

- Analysis for16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> i'

l e Detailed Analysis using UT Depth Data l - Creditfor uncracked ligaments j - Maximum period of operation based on allowable K t L

M .

EElli Crack Growth Rate Assessment e Several predictive models evaluated

- BWRVIPcorrelation

- GE PLEDGE model

- SKIcrack growth model

- NRC crack growth rates e NRC accepted growth rate of 5 x 103 in/hr is bounding

- Irradiation effects are bounded by the NRC curve

. - BWR shroud field cracking data confirms thatactual growth rates are lower; 2 x103in/hr bounds data

- NMP1 water chemistry during the last cycle has been excellent (less than 0.1 micro-siemen/cm)

NMP1 Crack Growth Rates Expcted to beMuch less than the Bounding Crack Growth B2te used in the Analysis

t .

EEll Screening Criteria Analysis .

Technical Approach

.e C9acks assumed through wallin all uninspected regions s Where indications found (UT/VT), through wall flaw assumed e LEFM and Limit Load analysis e ASME Code safety factors

~

- 3.0 Normal and upset;1.5 Emergency and faulted e Uncertainty factors for UT and VT included e Evaluations performed for 16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />

- Crack growth rate of 5 x 10-8in/hr '

- Indications acceptable if final length less than allowable value All Welds except V4, v9 and V10 shown acceptable by theScreeningCriteria Analysis

& 9 Elli Detailed Evaluation of V9, V10, and V4 Welds

^

e Detailed evaluations for the V9, V10, and V4 indications

- Credit for remaining ligament after crack growth of 5 x 104 in/hr and inspection uncertainty factors

- LEFM and limit load analysis with ASME code safety factors -

- Covers normal / upset and accident conditions

- Acceptable period for continued operation deisudned ~

e Analysis shows that continued operation is justified for at least10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> bl

9 e

EEll Structural Analysis Conclusions e tie rod repair design basis maintained even with the observedverticalweld cracking

- No credit taken for horizontal weld integrity

- e Structuralmargin demonstrated for continued operation for atleast10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />

- ASME Code safety factors maintained

- Bounding crack growth rates used .

- Conservative flaw sizing assumed Required Structural Margins Maintained IP

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' _ ,7 k UNITED STATES 9

j NUCLEAR RESULATORY COMMISSION W ASHINGTON. o.C. 20540@

May 8,1997 Mr. B. Ralph Sylvia Executive Vice President, Generation Business Group, and Chief Nuclear Officer Niagara Mohawk Power Corporation Nuclear Learning Center '

450 Lake Road Oswego, NY 13126

SUBJECT:

MODIFICATIONS TO CORE SHROUD STABILIZER LOWER WEDGE RETAININ AND EVALUATION OF SHROUD VERTICAL WELD CRACKING, NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 (TAC NO. M98170)

Dear Mr. Sylvia:

During the 1995 refueling outage for Nine Mile Point Nuclear Station, Unit No. 1 (NMP1), Niagara Mohawk Power Corporation (NMPC) installed four core shroud stabilizer assemblies consisting of tie rods, brackets, springs, wedges, and other parts. This repair measure was implemented to ensure the structural integrity of the core shroud with respect to the function of the circumferential welds that experienced cracks in some boiling-water reactors (BWRs) as discussed in NRC Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," and other communications. The 1995 shroud repair was designed as an alternative to the l

requirements of the American Society of Mechanical Engineers Boiler and-Pressure Vessel Code (ASME Code) pursuant to 10 CFR 50.55a(a)(3)(1) and was approved by the NRC staff by letter dated March 31,_1995.

l During the 1997 refueling outage, NMPC inspected core shroud vertical welds in accordance with the BWR Vessel and Internals Program (BWRVIP) document BWRVIP-07, " Guidelines for Reinspection of BWR Core Shrouds," dated February 1996, and observed cracks in excess of_the screening criteria. NMPC inspected the four shroud stabilizer assemblies and found that the tie rod nuts had lost some preload and that the lower wedge retainer clips on tiiree of the stabilizer assemblies had experienced damage. NMPC ti. formed the NRC staff of the inspection findings by telephone calls on March 20 and April 2,1997.

By letter dated April B,1997, NMPC described the inspection findings, provided analyses of the vertical weld cracking, and preposed a plan to restore the stabilizer assemblies to function as intended by the original design. In the April 8 letter, NMPC discussed '.oot cause and corrective actions, provided design documentation and evn uations to demonstrate the acceptability of the as-found vertical weld c. racking for at least 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of hot (above 200 *F) operation, propoced t weld re-inspection schedule, described actions taken to restore the .tabilizer assemblies to the as-designed condition, and proposed a modification to the lower wedge retainer clip design. NMPC requested the NRC <taff't approval of the modified lower wedge retainer clips as an alternate repair under the ASME Code Section XI definition of repair or replacenent pursuant to 10 CFR 50.55a(a)(3)(i) for the proposed operating i.ycle of 10,600 houts. To support the design assumptions for the repair (i.e., tFat the horizontal welds were completely cracked for Enclosure 3

3. Sylvia .

i for the_ vertical welds demonstrating that  ;

S60')Ccde,

,,NMPC includedSection a flaw-analys sXIided margins 26, 27 (two letters), would 30, and be maintained additional information by To support the-request for approval, NMPC prov ASME )

letters dated April ll, 23, 25'(two letters ..

May 7,-1997. (SE),--the NRC staff has lower wedge retainer clip design and As discussed in the enclosed safety ssembliesevaluation to the intended The proposed modification dis, reviewed the proposed by the NRC staff. modification to thefinds that- it a d

function previously approved t ff has also reviewed the vertical wel s an ration as proposed.

therefore, acceptable. The sembly NRCrepairs s afinds have beenthemof acceptable for 1 The NMP3- core shroud cracks andd statements stabilizer of asblic, local offici considerableThe interest to members NRC-staff received of several the pu questions an organizations. lton, New York, with the public on April 14,T concern during meetings in Fu ceived at this meeting and in 1997,. and in related correspondence.on the shroud and stab related correspondence. i Directive on Chemistryl and alues The NRC recognizes ing procedures that your i

Nuclear have Divisincorporated on applicab e vP Chemistry Department monitor " and that you will continue toWe further and action levels -issued by the Electr perating c cycle.

Water Chemistry Guidelines l tion of the metallurgical (" boat")

-1996 Revisionfollow thes recognize your intent to complete eva uad provide a report to th samples of the1 shroud cracks anWe request il 30, 1997, thatwillyou keep us informed that you 1997.

tion plans for the shroud and We acknowledge, basedeon theyourend of letter dated Apr operating the 10,600-hour provide us with copies off your reinspec stabilizer assemblies 3 months1)be or is contingent upon (1) cycle.

ithin the tguidelines set-forth in This NRC approval under 10 CFRTR-103515, 50.55a(a)(3)(

"BWR Water maintaining reactor coolant chemistry dance w with the commitment-in your Electric Power:Research Institute f i g thetechnical commitment reporof May 7, 1997, Chemistry Guidelines-1996 Revision," in accor30,1997, and (2) sat i for a license amendmentk that letter dated April rent TS conductivity limits for to submit, witnin 60 days, an applicat nalysis on assumptions for core shroud cracill rende h

addresses the difference reactor coolant chemistry and t e a between the curFailure to growth rates. approval null and void.

t.

.B. Sylvia ;This completes the NRC staff's efforts under TAC No. M98170. If you have -

questions, please contact me by phone at (301) 415-3049 or by electronic mail at dsh9nrc. gov.-

Sincerely, l'

@[ '? e Darl Hood, Senior Project Manager Project Directorate I Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation-Docket No. 50-220-

Enclosure:

Safety Evaluation cc.w/ enc 1: See next page 6

i k

k :

EXECUTIVE

SUMMARY

This safety evaluation assesses the 1997 reinspections and analyses-by Niagara Mohawk Power Corporation (NMPC and the licensee) for its core shroud-at Nine Mile Point Nuclear Station, Unit 1 (NMP1) to determine whether the shroud meets the structural integrity requirements of the American Society of Mechanical Engineers Boiler and Pressure Yessel Code (ASME Code),Section XI.

During the current spring 1997 refueling outage NMPC inspected vertical welds in the core shroud in accordance with the Boiling Water Reactor Vessel and Internals Project-07 " Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07)." NMPC found vertical weld cracking that exceeded the screening criteria of BWRVIP-07. Additionally, NMPC's inspections of four tie rod assemblies, installed as a pre-emptive repair during the last outage in 1995, showed that the tie rod nuts had lost some preload and that the lower wedge retainer clips on three tie ads were damaged.

NMPC found that the degraded repair was operational during the past cycle.

! Through its structural analyses, NMPC has concluded that continued operation I at NMP1 is justified for at least an additional 10,600 hour0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />s- (about ly-1/2 months). The allowable hours of operation until the next inspection ensure that the required ASME Code margins will continue to be maintained.

NMPC implemented design changes to the tie rod assemblies to correct the problems that were found during the inspection.

The NRC staff has concluded, based on the inspection and analyses performed and the modifications made to the tie rod assemblies, that the NMP1 core shroud is acceptable for an additional 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of operation. This is the NRC's approval under 10 CFR 50.55a(a)(3)(1) which is contingent upon NMPC (1) maintaining reactor coolant chemistry within the guidelines set forth in Electric Pow'er Research Institute technical report TR-103515, "BWR Water Chemistry Guidelines-1996 Revision," in accordance with its commitment by-letter dated April 30, 1997, and (2) submitting, within 60 days, an application for a license amendment that addresses this matter in accordance with its commitment by letter dated May 7,1997. Failute to satisfy either of these conditions will render this approval null and void.

'NMPC will submit to the-NRC an inspection plan for the next outage at least 3 months before the outage is scheduled to begin.

I

p CIC f UNITED STATES 4 1E NUCLEAR REGULATORY COMMISSION

-) f WASHINGTON, D.C. 2065 Hoot

\p..ci4/

..+

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING RESULTS OF THE REINSPECTION OF THE CORE SHROUD NIAGARA MOHAVK POWER CORPORATION NINE MILE POINT NUCLEAR STATION. UNIT I DOCKET NO. 50-220

1.0 INTRODUCTION

1.1 Purpose l This sPf'ety evaluation (SE) assesses the reinspection results and analyses submitted by the Niagara Mohawk Power Corporation (NMPC and the licensee) for its core shroud at Nine Mile Point Nuclear Station, Unit 1 (NMP1) to determine whether the core shroud meets the structural integrity requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME  ;

Code),Section XI.

The NRC staff considers the core shroud repair by means of shroud stabilizer assemblies (also called tie rod assemblies) to be an alternative to an ASME Code repair under 10 CFR 50.55a(a)(3)(i). Since the assumption for the repair is that the circumferential welds are fully cracked, the vertical welds need a minimum ligament to maintain the structural integrity of the shroud. The NRC staff reviewed the analysis of the vertical welds to determine whether the results supported the alternative repair. The alternative repair and Code requirements are discussed in Section 2.0 of this SE.

1.2 Background

By letters dated January 6 and January 23, 1995, NMPC submitted an application for repairs to the NMP1 core shroud. The shroud repairs and the use of stabilizer assemblies (tie rods) were submitted as an alternative, as discussed above. The NRC staff approved this repair in its 5E dated March 31, 1995, according to which NMPC was also to address certain issues by the upcoming spring 1997 inspection (refueling outage 14). NMPC committed to submit its plan for reinspection of the core shroud repair assemblies and the core shroud following refueling outage 13. Also, as noted in the SE, the NRC staff recommended that NMPC perform certain activities to qualify the ultrasonic testing (UT) techniques used to inspect weld H8 and to develop an effective method to locate the segment welds of the top guide support ring.

The NRC staff also noted in the SE that NhPC would reinspect all reported indications on the top side of the H8 weld during refueling outage 14. The reinspection would verify the postulated crack growth of these indications.

During the post-installation inspection of the shroud repair, NMPC identified conditions that differed from the intended repair design and reported them in O

its letters of March 25 and April 30, 1995. NMPC committed to take appropriate corrective actions.during refuelingNMPC outage 14 to restore the proposed the final design shroud stabilizers to their original design.

modifications in its letter of August 14, 1995, as part of the long-term corrective action plan. In its SE of March 3, 1997, the NRC staff reviewed the anomalies in the installed shroud repair hardware and approved the proposed corrective actions.

By letter dated October 4,1995, NMPC submitted its insoection plan for the core shroud and its repair assemblies. The plan described the inspections projected for the core shroud stabilizer assemblies, the repair anchorage, the H8 weld, the top. guide ring segment welds, and the vertical welds. In its letter of February 7, 1997, NMPC confirmed its intent to conduct shroud inspections in the spring 1997 outage according to the inspection plan it previously submitted. NMPC also presented adGitional information about the inspection plan, addressed the issues in the SE of March 31, 1995, and provided the fracture mechanics analysis that determined the next required ultrasonic inspection of the H8 weld.

In its letter of February 28, 1997, NMPC updated the plans for inspecting the vertical welds described in the October 4,1995, letter. The inspection scope discussed in the letter was based on a draft version of the Boiling Water Reactor Vessel and Internals Project-07 " Guidelines for Reinspection of BWR Core Shrouds," EPRI Report TR-105747 (BWRVIP-07). NMPC stated that its current plans are in conformance with the BWRVIP-07 guidelines, with one exception related to the expansion criteria for inspection of vertical welds.

NMPC also corrected statements made in its letter-of February 7,1997, about the scope of inspection of the H8 weld.

On March 3, 1997, the NRC staff issued its SE of the proposed inspection program. It found that NMPC addressed the issues noted in the SE-of March 31, 1995. It also found NMPC's plans acceptable, but stated that the inspections for this outage should be conducted in accordance with-the BWRVIP-07 guidelines without the exception. Although the NRC staff has not-completed its review of these guidelines, it accepted their use for this NMP1 refueling outage.

NMPC inspected core shroud vertical-welds in accordance with the BWRVIP-07 guidelines. On March 20 and April 2, 1997, it notified the NRC weldstaff (by that telephone) of the inspection findings. NMPC found vertical cracking exceeded the screening criteria of BWRVIP-07. Additionally, its inspections-of the four tie rod assemblies showed that the tie rod nuts had lost some preload (clamping force) and that the lower wedge retainer clips on three tie rods were damaged.

In its submittal of April 8,1997, supplemented by letters of April 25 (two letters), 26, and 27 (two letters), and 30, 1997, NMPC reported the findings of its inspections. It presented a root cause analysis, corrective actions, design documentation to establish the acceptability of the vertical weld cracking, a weld reinspection schedule, and details of the actions taken to restore the tie rods to the as-designed-condition and to modify the lower ,

]

wedge retainer clip design. 1 i

1

l L 2.0 CODE RE0VIREMENT AND ALTERNATIVE in accordance with ASME Code Section XI, 1983 edition, Summer 1983 Addenda, NMPC chose to perform corrective measures for the cracked circumferential shroud welds on a pre-emptive basis. The repair for cracked welds to satisfy the repair rule of Section XI, IWA-4000 would be to remove the defect and conduct a repair by welding. This is impractical since the core shroud is irradiated to an extent that might cause-the shroud to crack further if it were welded. Therefore, as an alternative to IWA-4000, NMPC chose to repair the core shroud with mechanical tie-rod assemblies that serve as a replacement for the circumferential welds.

The NRC staff has approved the tie-rod assembly as an acceptable alternative repair in an SER dated March 31, 1995. The current request, which is the subject of this SE, is to modify a component of the tie-rod assembly (i.e.,

the lower wedge retainer clip) pursuant to 10 CFR 50.55a(a)(3)(i) and institute a revised installation procedure to ensure that pre-load of the tie-rod assemblies will not be lost by unanticipated motion of the lower toggle bolt supports.

The design assumptions for the alternative repair are that circumferential welds are cracked through-wall for 360' and that the tie-rod assemblies will accommodate the vertical and lateral loads for the " stacked cylinders." The cylindrical sections are fabricated from rolled plates. Two vertical welds are used to join the sections, completing the cylinder. Therefore, vertical weld structural integrity is needed to ensure that the cylinders will remain as cylinders. The crack growth rate used as a part of the fracture mechanics analysis for assessing the needed weld integrity is considered bounding if the plant water chemistry is maintained in accordance with the EPRI BWR Water Chemistry guidelines. However, the NMP-1 current TS is not consistent with the BWR Water Chemistry guidelines. NMPC has committed to submit an application for a license amendment within 60 days to address this matter.

The NRC staff has reviewed the modification and the fracture mechanics analysis-that ensure the structural integrity of the vertical welds for the requested operating period of 10,600 hot operating h;urs and finds that the  ;

lilternative repair is acceptable pursuant to 10 CFR 50.55a(a)(3)(i). This '

approval is predicated on the condition that NMP-1 is operated in accordance with the BWR Water Chemistry guidelines, Electric Power Research Institute technical report TR-103515, 'BWR Water Chemistry guidelines-1996 Revision."

Details of the review are discussed in the following sections.

3.0 SHROUD INSPECTIONS Figures 1-1 and 1-2 show the location of the shroud welds.' Table 2-2 shows the findings for vertical weld inspections. NMPC did not present a table of findings for the horizontal welds, but provided examination indication maps

• Att figures and tables attached to this htC staf f SE are extracted, without change, f rom reports submitted by hMPC. Therefore, figure and table twrt>ers do not necessarily correspond to the section ru @ering within this $E.

( -

and discussed the results in a qualitative way in Appendix C of the report by General Electric Nuclear Energy (GENE), titled " Assessment of the Vertical.

Weld Cracking on the NMP1 Shroud," GENE-523-B13-01869-043 (portions attached).

The scope, methods, and findings of the inspections are discussed hereinafter in Sections 3.1, 3.2, and 3.3, respectively.

3.1 Inspection Scope As discussed in Section 1.0, NMPC stated that its current plans were in accordance with the BWRVIP-07 guidelines, but with one exception. The exception concerned the inspection of vertical welds. For the vertical weld inspections scheduled for the spring 1997 outage, NMPC proposed to modify the method for sample expansion presented in BWRVIP-07 as Option B.

According to Option B, if the cumulative cracking in either the original sample or the expansion sample exceeds 10 percent of the equivalent length of weld inspected, the inspection scope is to be expanded to verify the minimum required uncracked length for each vertical weld that is not structurally replaced by existing hardware or the repair or both.

NMPC proposed to expand the inspection scope to verify the minimum required uncracked length for each vertical weld not structurally replaced by existing hardware or the repair or both only after finding the cumulative cracking in either the 50-percent expansion sample or the 100-percent expansion sample greater than 10 percent of the equivalent length of weld inspected.

The NRC staff did not grant the exception, finding that NMPC did not present adequate technical bases to deviate from the BWRVIP-07 guidelines. The NRC staff's judgment was that the provisions of Option B should be followed and hence, a finding of more than 10 percent of the equivalent length of weld inspected being cracked would merit additional inspections to further qualify the degree of cracking.

NMPC inspected the NMP1 vertical welds'according to the sampling option of BWRVIP-07. This option specified a visual inspection of 25 percent of the equivalent total vertical weld length from either the outside diameter (0D) or

'inside diameter (ID).

The rina segment welds (labeled in Figure 1-2 as V5 and V6) were excluded from the vert: cal welds requiring inspection based on a GE analysis of the ring segment welds submitted to the NRC staff for review by letter dated February 7, 1997.

The initial inspection of the vertical welds with enhanced visual techniques found cracking over the entire OD length of the V1.0 weld. NMPC then expanded the inspection plans'to establish the minimum required uncracked ligaments on the vertical welds that are required to meet the shroud stabilizer repair design-basis assumptions. It performed the inspection using an enhanced visual inspection (EVT) method supplemented by UT.

NMPC found extensive cracking on the 00 of vertical welds V9 and V10. As a result, NMPC performed a complete baseline inspection of the accessible parts

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, of most horizontal and vertict1 welds to assess the overall material condition of the NMP1 core shroud. NMPC examined the accessible areas of the ring but was unable to locate the welds V5 and V6. NMPC inspected the hurizontal welds H2, H4, H5, H6a, H6b, ar.d H7, even though ins)ection of horizontal welds was nt,t required because a preemptive repair had >een installed. Such re) airs are designed assuming the circumferential welds are fully cracked (throug1-wall for 300 degrees), obviating the need to inspect them.  ;

The inspection scope is summarized in Table 2-2 and Appendix C. The scope was

.ufficiently comprehensive tu identify degradt. tion that could invalidate design assumptions for the tiv rod repairs. In expanding the scope of inspections, and inspecting from the ID and the OD of the weld, NMPC determined whether each vertical whid contaired the minimum ligament needed for maintainthg the strut.tural integrity of the shroud.

1 The relation of the results of vertical weld inspections to the design assumptions for the tie rod repairs is discussed previously in Section 2.0 of this SE.

3.2 Inspection Methods ble 2-2 and Appendix C indicate the inspection method used for the various ds. EVTs aro performed with a video camera qualified in tests to detect a j-mil wire. Visual inspections cannot determine the depth of a flaw. They 6 e uted to detect surface flaws on either the ID or 0D. Ultrasonic inspections are volumetric examinations used to detect flaws and determine their size. Accessibility of the weld determines which method is used. Table 2-2 shows that each of the vertical welds was examined from the OD and ID either by EVT or UT or both. Supplemental eddy current inspections were also used to determine if defects were present on the ring segments containing welds V5 and V6. These welds are difficult to locate by UT or visually because the support ring was machined after fabrication and welding.

The UT for examining the core shroud at NMP1 was performed with a GE delivery device, the suction cup scanner.- Three transducers were used including a 45' shear, 60' longitudinal, and an OD creeping wave. The inspection device was demonstrated and qualified to detect and determine the size of cracks on a

' core shroud mock-up at the EPRI Nondestructive Exanination (NDE) Center. This device has been used to perform core shroud inspections at a number of other BWRs.

The crack size (length and depth) measurement uncertainties for this device were determined in accordance with guidelines in the " Reactor Pressure Vessel and Internals Examinations Guidelines (BWRVIP-03)' (EPRI TR-105696), October 1995. In measuring length, uncertainties arise from the delivery system and the NDE technique. The qualificatte of the system at the EPRI NDE Center included quantification of measurement uncertainties. The uncertainty factors from the delivery system and the NDE technique were determined to be 1.106

-inches and 0.364 inches, respectively. The combined uncertainty factor of 2.94 inches ;(1.106 inch 4 0.364 inch) x (2 ends)) was applied to each ultrasonically measured flaw length including the calculated critical crack length. The 60' longitudinal wave transducer was used in depth measurement.

The uncertainty factor associated with the depth measurement was determined to

6-be 0.108 inch (no contribution from delivery system), and was applied to each UT depth measurement.

The EVT performed for the core shroud inspactions was qualified by The demonstrating the capability of resolving a 0.5-mil-diameter wire.

uncertainty factor for the length measurement by the EVT was determined to be 1.2 inch. To account for the measurement uncertainty, flaw length.

2.4 inches (for two ends) were added to each visually measured 3.3 Inspection findings NMPC presented detailed descriptions and examination indication maps of the vertical weld cracking. The results are summarized in Table 2-2 and are included in Appendix C (attached). The inspections showed significant Minor cracking was found on welds V3, V12, l

cracking V15, and V16.on welds V4, V9, and V10.No cracking was found on the accessible areas of welds and Vll,. Welds V5 and V6 were not located.

The weld with the worst cracking was V9. It had numerous indications on more Weld V10 had than 90 percent of the weld and up to 80 percent of the depth.

numerous indicatioas on more than 80 percent of the weld.

NMPC tried to examine the ring segment welds V5 and V6, welds known to beUsin difficult to examine given current technology. However, EVT of thn inspections, NMP(, -ould not locate the specific welds. accessible outer surf The NRC staff approved NMPC's inspection plans for these welds in its SE of March 3,1997. Those plans contained an analysie that showed that if NMPC could not locate these welds, the shroud repair would remain effectivo even if the welds were significantly cracked. NHPC's evaluation showed that the structural integrity of the ring would be maintained even with cracking up to However, the tie rod re] air may be less 95 percent of the segment welds. effective if ring segment welds are totally cracked a Under this horizor.tal welds, H2 and H3, are also completely cracked. -

condition, the stiffness of the core shroud may ci.ange and, consequently.

. during affectnormal the amount operation.

of the thermal preload applied by the tie rod assembliesT ,

length exists in the H3 weld to support operation, even if the ring segment welds were significantly cracked.

NMPC submitted indication maps of the horizontal welds and a qualitative description rather than a table of quantitative results. The installed stabilizer assemblies were designed to ensure shroud integrity even if all the horizontal shroud welds were cracked through the wall around the full circumference of the shroud. Thus, the condition of the horizontal shroud welds did not need a quantitative assessment to support continued operation.

The findings of the inspections are, however, included in Appendix C (attached).

The structural significance of the indications is assessed in Section 4.2 and the potential for bypass leakage in Section 4.3.

7 4.0 EVAtVATION OF INSPECTION FINDINGS 4.1 Root Cause NMPC determined that the cracking in the vertical welds was caused by intergranular stress corrosion cracking (IGSCC). NMPC's findings agree with l

research and operating experience with this phenomenon.

It is well known that the core shrouds of all boiling water reactors (BWRs) are susceptible to IGSCC. The following relevant factors affect the cracking:

operating time, coolant conductivity, material carbon content, plate l

orientation, fabrication-related surface cold work, neutron fluence, residual ,

stresses resulting from welding and fabrication and operating stresses. The NMP1 shroud is susceptible to IGSCC because it Is made of high-carbon Type 304 l

stainless steel that is sensitized by welding and subjected to residual stresses from welding and fabrication. NMP1 is classified by industry guidelines as a Category C plant, a category that contains the most susceptible core shrouds. Category C plants have core shrouds made of Type 304 stainless steel and more than 6 hot operating years, regardless of reactor coolant conductivity.

The location of the cracking in the NMP1 shroud is consistent with IGSCC as shown by inspection data. The cracking was located at heat-affected zones (HAZs) sensitized by welding and in areas with residual stresses in which fabrication-related welding or grinding was apparent. In a few limited instances, cracking was found to extend up to 1.5 inches away from the weld perpendicular to the longitudinal axis of the weld. A few crack indications not associated with HAZs were found in the base metal. However, these cracks initiated from the cold-worked areas resulting from grinding to remove the attachment welds (lugs) during construction.

The cracks are expected to stay confined to the HAZs or in the severely cold-worked area. Radieion, when exceeding a threshold value, can sensitize materials, making them susceptible to IGSCC. However, offsetting this effect, is the reduction of tensile residual stresses outside the affected areas.

Unless high tensile stresses, such as those that result from grinding or cold work are present, there is not enough driving force to propagate a crack beyond the affected areas. The depth of the cold-worked layer in the base metal resulting from grinding is generally very shallow and, therefore, the cracking in the cold-worked areas will also not grow very deep.

NMPC contended that although some cracks extended beyond the HAZ, the/ did not-differ from those seen in other BWR plants in the past. To check the contention, the NRC staff reviewed inspection data for circumferential welds, considering data for the H5 weld at Brunswick Unit I as an example. The inspection report for that plant of June 21, 1995 (Brunswick Steam Electric Plant No. 1. Docket No. 50-325/ License No. DPR-71 NRC Generic Letter 94-03,

• Intergranular Stress Cctrosion Cracking of Core Shrouds in Boiling Water Reactors," letter dated June 21, 1995, from R. Anderson of Carolina Power and Light Co. to the NRC) for the 10th refueling outage states that H5 was visually examined 100 percent of the ID and 45 percent of the 00. Five of the cracks on the OD were circumferential. 0.5 to 3 inches long. The remaining

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cracks Long circi we'; axul, extending less than 9 inches from the toe o Baseline depth measurements were taken by UT in two locations were on tha 10. The depth ranged from less than 0.3 inches to benchmark future inspections.

to 0.6 inches.

That inspection report further states that during the 10th refueling In outage,

" punch marked crac(s" on the ID were visually reinspected for length.

addition, two areas inspected by UT during the 9th refueling outage were ultrasonically reinspected to determine crack growth and were co previous or length. sizing data.This conclusion supports the contention that cracks are not li<ely to propagate outside the cold-worked area and ,that the crack growth rate i slower than the NRC's bounding rate of 5 x 10' inch per hour.

The NRr. staf f considered the question of the adequacy of base NMPC metal inspections. HAZs are typically narrow and adjacent to the weld.

reported that the HAZs associated with the verticalHMPC welds near the beltline visually extended about 0.8 to 1.0 inches from the weld fusion line.

examined Considering the about extent  ?.5 inches of the of base examination metalwith together onthe each sidetheof each v fact that driving force for cracks is greatly reduced outside the HAZ, the NRC staff concluded that NMPC examined enough area outside the HAZ to determine the condition of the base metal.

The observed.

NRC staff considered the effect of radiation fluence on was in a range for which irradiation-assisted stress corrosion cracking (IASCC)conditionscanoccur.',ThegstimatedfluenceatweldsV9andV10wasNM n/cm (>l HEV),

in the range of 2 to 4.5 x 10 was a contributing mechanism to the cracking by analyzing boat samples of This work will be completed and the results will be cracked material. This issue does not need to be resolved submitted to the NRC for review.immediately because the crack growth r Crack growth rates are evaluated in Section 4.2.

of fluence.

7he NRC staff finds that HMPC's analyses reasonably explain the observatio they showed that the welding and The/fabrication also showedprocesses caused that the pattern of the crackin pattern seen on the vertical welds.

crack depth is consistent with the calculated fluence axial and radial profiles. That is, both welds V9 and V10, the welds with the worst cracks, are between welds H4 and H5 and are exposed to neutron irradiation along entire length. The axial cracks in weld V9 were found to be the deepest near the H4 weld where the fast neutron flux is highest.

4.2 Structural Integrity Assessment The methodology for assessing the structural first,integrity the size ofof athe flawed core sh axial weld, is b-oken down into a number of steps.

l flaw at the time of the analysis (the initial flaw size, a ) is determined by l ultrasonic or visual inspection of the weld and this flaw ks conservativel I

assumed to be present completely through the thickness of the core shroud

.g.

I wall. This value, ai, represents the best estimate of the flaw size.

However, some uncertainty exists in a because of the method used tc' determine the flaw size. When the uncertainty due to the inspection methodology is added to the best estimate of the (discussed initial flaw size,in detail a new in Section value, a 2.0)is determined.

This value, aa , is taken >

as the initial flaw size in the s$,ructural analysis since it represents a j bounding value for the size of the existing flaw. Then a crack growth rate l (da/dt. is determined. For this analysis, a bounding value of the IGSCC l growth)ratewasassumed(5x10 4 inch per hour) and multillied by the amount of operating time to be justified (t) to determine a bouncing valut for the flaw size at the end of the. proposed operating time (the final flaw size,- a,).

It is this final flaw size, a that must be shown to be less than some critical flaw size determined,,by prescribed loading conditions (with appropriate safety factors) according to structural integrity assessment methodologies.

The flaw sizes based on reported measured values and the uncertainties associated with these measurements are discussed in Section 3.3. Section 4.1 discusses assumed crack growth rates. Section 4.2 discusses structural integrity assessment methodologies. Section 4.3 summarizes NMPC's structural integrity assessment, and Section 4.4 discusses the hRC staff's independent analyses.

4.2.1 Evaluation of Crack Growth Rates l

GE, in its analysis for NMPC, assumed a bounding crack growth rate of 5 x 10*5 inch per hour. This crack growth rate bounds laboratory test data for a variety of water chemistries, bounds field experience with IGSCC of other components of the reactor coolant system, and has been used by the NRC for flaw growth evaluations. Even after considering potential IASCC contributions, the actual crack growth is expected to be lower than this bounding value.

GE aredicted the crack growth rate for the vertical welds using several metiods, it considered the effects of radiation on crack growth rates, comparing predictions of rates for unirradiated material to those for irradiated materials at the same values of reactor water conductivity,

, and initial sensitization. The com)arison electrochemical motential showed that at tie fluence leve (ECP)ls at BWR shrouds, the predicted crac( growth rate is similar for both the unirradiated and the irradiated materials. GE attributed the similarity in rates to offsetting effects. Although irradiation increases the susceptibility of the material, it also relaxes the weld residual stresses, the driving force for crack growth. Ultimately, GE concluded that the crack growth rate over the range of shroud fluences is bounded by the NRC bounding crack growth rate of 5 x 104 inch per hour. The NRC staff agrees with GE's determination.

The NRC staff considered the effect of the distribution of residual stresses on the crack growth rate. Similar results to those reported by MPM Technologies in its

• Analysis of Nine Mile Point Unit 1 Shroud Welds V9 and Weld V10 Cracking,' Report No. MPM-497439, April 1997, were obtained by researchers performing NRC-sponsored work at Argonne National Laboratory

(ANL).

Researchers at ANL analyzed the distribution of residual stresses from welding in shroud welds. The work at ANL, unlike that of MPM, did not include residual stresses induced by fabrication, but did find residual stresses from welding to be highly tensile at the surfaces and compressive in the center of the wall of the cylinder. Additional observations include that actual conditions like *out of roundness" would bias cracking to a particular surface. Further, if stresse, were uniform Even across the thickness, cracks would assuming multiple crack initiation grow at about a 2:1 aspect ratio.

sites, the aspect ratios found in the cracking at NMP1 imply that the stresses decrease sufficiently through the wall to reduce the stress intensity. The dec,rease in stress intensity is clearly more than sufficient to show that 5 x 10' inch per hour is a conservative rate of crack growth.

The NRC staff examined the question as to whether crack growth rates in The rates are vertical welds differ from the rate for horizontal welds.

The cracking on both the expected to be similar for the following reasors:

horizontal and vertical welds is caused by the same mechanism, IGSCC, in analogous locations, that is in weld HAZs, which are areas of high residual stress. The residual stre n patterns are similar for both kinds of welds.

They are tensile on the outer surfaces and decrease, even become compressive, within the wall and with distance away from the weld.

At NMP1, no field data on crack growth are available for vertical welds. Of the horizontal welds, field data are available only for the H8 weld. During the current outage, NMPC located a flaw found by UT during the previous (13th) refueling outage as well as an additional flaw in the same area. Through UT, NMPC determined that the flaw was of less through-wall depth than in the 13th refueling outage. NMPC concluded, after reviewing the previous data, that the earlier sizing was very conservative. NMPC drew no conclusions from these data about the crack growth rate at NMP1, and the bounding rate of 5 x 10**

inch per hour was used in NMPC's analyses.

The NRC staff examined the question of whether the water chemistry at NMP1 conferms to the requirements under which the bounding rate is applicable.

This rate was derived from laboratory test data. The reference for the data is a letter from NMPC to NRC,

• Responses to NRC Staff Questions Provided During Telephone Conversations of April 22, 1997, on Core Shroud Cracks and Repair," dated Aprfl 27, 1997. The crack growth tests for sensitized stainless steel were performed at a range of conductivities from 0.3 to 1.5 microSiemen/cm.S The data from these tests are bounded by the crack growth rate of 5 x IO' inch per hour.

In its letter of April 30, 1997, NMPC committed to continue operating in accordance with the EPRI Water Chemistry Guidelines. These guidelines are also incorporated into BWR'ilP-07; that document assumes that plants meet the requirements specified by the EPRI Water Chemistry Guidelines Action Level 1.

According to Action Level 1, if the conductivity exceeds the limit of 0.3 microSiemen/cm, the conductivity must be reduced to that value or lower within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. Although the conductivity limit in NMP1 Technical Specification 3.2.3 is 5 microsiemen/cm (when steaming rate are equal to or greater than 100,000 pounds per hour), NMP1 has been operating at a much lower conductivity, less than 0.10 microSiemen/cm, during the pastTherefore, three cyclestheand less than 0.3 microSiemen/cm during the last seven cycles.

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bounding crack growth rate applies to HMP1 because this rate is conservative with respect to Action Level 1 in the EPRI Guidelines. By letter dated May 7, 1997, NMPC has committed to submit a license amendment within 60 days that will address the difference between the current reactor coolant system chemistry requirements of NMP1 TS 3.2.3 and the coolant chemistry criteria referenced for core shroud crack growth rates as described in NMPC's letter of April 8, 1997. The NRC staff finds this commitment acceptable.

4.2.2 Structural Integrit Assessment b Linear Elastic Fracture Mechanics and Limit Load Anal sis Methodolo les i Either of two analysis methodologies, limit load analysis (LLA) or linear clastic fracture mechanics (LEFM), may be applied to demonstrate the structural integrity of flaws in ductile materials such as Type 304 stainless steel. Either of these methods seeks to determine the length of flaw that can be tolerated (or, conversely, the amount of uncracked ligament that must remain) for the location to support the applied stresses under )rescribed loading conditions. These loading conditions are divided for t1e purpose of this analysis (and by the ASME Code) into two general categories: (1) normal operation or upset conditions and (2) emergency or faulted conditions. The significance of this is that the ASME Code analysis procedures require that a safety factor of 3.0 be applied to the loads determined for normal operation or upset conditions, while a safety factor of 1.5 is to be applied for emergency or faulted conditions.

The first methodology, LLA, determines the amount of uncracked ligament that must remain to keep the location from being pulled apart by full-section plastic yielding. An LLA is usually applicable to materials that retain a high level of fracture toughness and demonstrate no crack extension under loadings up to those that cause plastic yielding. The results of this method are then compared to the results of an LEFM analysis to determine which method controls the evaluation (i.e., gives the smallest acceptable flaw size).

LEFM is a conservative method for assessing the potential for crack growth in a ductile material. LEFH assumes that the material will behave in a nonductile fashion (which reduces its flaw tolerance) and compares the stress Jntensity due to the flaw and the applied loading to the material's fracture toughness (a quantitative measure of the material's flaw tolerance). LEFH would normally be expected to be appropriate for materials that have lost ductility, for example, from irradiation damage.

4.2.3 NMPC's Evaluation of Structural Integrity GE performed the analysis that determined the structural margin of the vertical welds.

The assumptions for the shroud repair are that horizontal welds are cracked through-wall. Under this assumption, vertical weld cracks are acceptable, as long as the crack lengths are less than the allowable flaw sizes or as long as the structural integrity of the vertical weld can be shown. Typically, the allowable crack sizes are large and approach or exceed the length of the weld itself.

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The analysis determined that the primary stress that could cause vertical weld failure would result from the internal pressure. Consistent with ASME Code practice (Appendix C,Section XI), the analysis considered internal pressure as the only load for axial cracks. It considered the internal pressures under all conditions--normal, upset, and accident events--with the appropriate ASME Code safety factors.

The analysis did not take credit for the horizontal weld integrity in detennining the alloweble vertical weld flaw sizes. The horizontal welds are assumed to be cracked completely through the wall, and the cylinder between any two horizontal welds is assumed to be " stand-alone." The calculations also assumed simultaneous cracks at the diametrically opposite welds in a

'given cylinder. The analysis applied both LEFM and limit load to determine '

the allowable flaw sizes. Calculations for the required minimum ligament allowed for crack growth and inspection uncertainty.

In the analysis, GE first determined whether the vertical welds met the screening criteria specified in BWRVIP-07. GE then performed a more detailed fracture mechanics analysis to demonstrate _ the integrity of vertical welds with indications that did not conform to the screening criteria.

Allowable shroud vertical crack lengths were calculated on the basis of both LEFH and LLA. The high fluence region in which there is potential irradiation embrittlement is limited to the shroud section between horizontal welds H3 and H6a. Therefore, both LEFM and LLA were used for vertical welds in this region. LEFM was found to be governing for welds V9, V10, Vll, and V12, and LLA was governing for welds V7 and V8. The vertical welds in other regions were governed by LLA.

Table 5-2, Columns 3 and 4, (see enclosure to this SE) shows the allowable crack length as well as the required uncracked ligament lengths for each vertical weld, in accordance with BWRVIP-07, allowances for crack growth and uncertainty in the inspection were added to the required uncracked ligaments (Column 5 of Table 5-2).

Cracking of radial ring welds was separately evaluated and found to have

. negligible impact. Tne only effect of radial ring weld cracking is on the thermal preload, but the analysis found that even with 90 percent of the weld essumed to be cracked, the effect on preload was insignificant.

Column 6 in Table 5-2 lists the values of uncracked ligament lengths determined from the UT and EVT of the various shroud vertical welds. A comparison of these lengthr with those required by the conservative screening criteria lengths (Column 5) shows that each of the vertical welds except for 4 V4, V9, and V10 meets the required uncracked ligament length criteria.

The analysis showed that each weld except for V4, V9, and V10 is acceptable for continued operation for at least a fuel cycle of 16,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. These three welds were analyzed further. In this more refined ant. lysis, credit was taken for uncracked ligaments for part-through cracks after accounting for crack growth.-

Of the three welds, V9 was determined to be the most limiting case and was, therefore, evaluated first. The analysis applied the required safety margins ,

consistent with the ASME Code Section XI criteria for piping. The analysis -

justified continued operation for at least 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (approximately 14-1/2 months).

Weld V4 was evaluated by LLA because it is located abova the top guide support ring where the fluence is low enough that the material is not embrittled.

Table 5-2 shows the required uncracked ligament. The analysis showed that the limit load margins are satisfied for the v4 weld for a period of at least 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />.

The structural analysis concluded that continued operation at NMP1 is justified for an additional.10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (about 14-1/2 months). The allowable hours of operation ensure that the required ASME Code margins are maintained, considering both LEFM and LLA. The NRC staff finds this conclusion acceptable. The analysis was performed according to methodology approved by the NRC staff. The methodology is detailed in the General Electric Nuclear Energy document GENE-523-113-0894 Revision 1, *BWR Core Shroud Inspection and Flaw Evaluation Guidelines," March 1995, and has been used for circumferential welds, it is incorporated by reference in BWRVIP-07, 4.2.4 Independent NRC Staff Structural Integrity Analyses The NRC staff has rerformed an independent, quantitative assessment of the largest axial flaw in the NMP1 core shroud. The NRC staff evaluated this flaw using both LLA and LEFM techniques. The NRC staff found that the LEFM analysis, presented below, controls the evaluation.

The NRC staff accepted NMPC's best estimate for the largest initial flaw length of approximately 74 inches. The NRC staff accounted for the uncertainty in the sizing of the initial flaw by adding the uncertainty factor in GE's ultrasonic testing method (1.5 inch) to each end of the best-estimate flaw. Therefore, the NRC staff used a value of 77 inches for-a o, the bounding initial flaw size. The NRC staff then applied the bounding crack growth rate (5 x 10'S inch per hour) for 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (approximately 14-1/2 months) to determine a bounding final crack length.

The NRC staff then applied the stresses that were reported to result from faulted loading conditions with the safety factor of 1.5 (as these were-reported to be the most challenging loading conditions) and used an analytical equation that contained a small correction factor for long cracks that was not included in GE's LEFM analysis. This resulted in a calculated stress intensity of about 175 ksilin at the tip of the bounding final crack length.

The NRC staff compared the result with the fracture toughness of 150 kst/in reported by GE from test results for material removed from a foreign reactor.

-Further, the NRC staff has determined that additional test data for weld material from another foreign nuclear power plant indicate similar fracture toughness to that discussed. In addition, an elastic / plastic fracture analysis was performed by Structural Integrity Associates with the fracture toughness of the base metal reduced by about a factor of 3 to bound weld metal properties. The analysis by Structural integrity Associates produced results

that were similar to the LEFM analysis performed by GE. On the basis of the information reported by NMPC and the NRC staff's analysis, the bounding final crack length for the limiting core shroud flaw is slightly larger than would be acceptable under LEFH analysis. This independent evaluation by the NRC staff used a handbook solution that is not as precise as NMPC's finite element analysis. The results from the analysis is that the assumed material toughness is slightly exceeded. Considering the test data from the two foreign nuclear sower plants yielding similar values for fracture toughness, the results of tie elastic / plastic analysis serformed by Structural Integrity Associates and the uncracked ligaments for w11ch credit was not taken in NMPC's analysis, the NRC staff concludes that the more refined analysis performed by NMPC appears reasonable. Finally, the NRC staff confirmed that the LEFM analysis was controlling by determining the acce) table amount of remaining ligament according to LLA to be 7.0 inches, muc1 less than that required by the LEFM analysis. Furthermore, evidence exists that the materials removed from shrouds that have been tested are failing in a ductile indicating that LLA or elastic / plastic analysis may be more manner, appropr fate anklysis methods.

4.3 Assessment of Potential for Bypass Leakage NHPC evaluated the potential crack opening and associated leakage for the V-9 and V-10 vertical welds, as documented in General Electric Nuclear Energy Report GENE-B13-01869-043, Revision 0, " Assessment of the Vertical Weld Cracking on the NMP1 Shroud," April 1997. The evaluation also considered whether the leakage would be detectable. The NRC staff's review of NMPC's ,

leakage evaluation is presented below.

4.3.1 Leakage Estimate By conservatively assuming that both the V-9 and V-10 welds were cracked through-wall, NMPC estimated that the postulated vertical weld cracks would provide 3-square inches of leakage flow area (approximately a 0.003-inch opening along the entire length of the welds). On the basis of this leakage flow area, the total calculated leakage from the V-9 and V-10 welds was '

estimated to be 200 gom at 100-percent rated power and 100-percent rated flow.

.This statistic is equivalent to approximately 0.11 percent of the core mass flow.

In a letter dated March 31, 1995, the NRC staff presented its evaluation of the estimated leakage with the tie-rod assemblies installed and with the postulation of through-wall cracking of all the horizontal welds. The NRC staff found that with the tie-rod assemblies installed, the estimated leakage was 0.33 percent of core flow at 100-percent rated power and 85- to 100-percent rated core flow. The NRC staff notes that the postulated 360' through-wall cracks in the horizontal welds with the tie-rod assemblies installed yield a higher estimated leakage than the total estimated leakage from the V-9 and V-10 welds cracked through-wall. NMPC estimated that the total leakage resulting from the tie-rod assemblies installed and the postulated through-wall cracking of the horizontal welds and the V9 and VIO ,

i welds was 0.65 percent.of the core mass flow at 100 percent rated power and 100-percent rated core flow. The NRC staff coes not consider this a

15-significant leakage rate since it is less than 1 percent of the total core flow and is, therefore, acceptable.

The NMP1 emergency core cooling system (ECCS) consists of the single-train feedwater coolant injection (FWCl) system, the automatic depressurization system (ADS), and the two-train core spray (CS) system. M FWCI system requires limited offsite power to be functional. During a postulated loss.:of-coolant accident (LOCA), the CS system transfers water from the suppression pool to the reactor pressure vessel (RPV) where the water is sprayed upon and cools the core and returns to the suppression chamber by way of the break. CS l system water is introduced at the top of the core using circular spargers l located inside the core shroud and containing distribution nozzles pointed i- radially inward and downward to produce fine spray droplets directed onto all  ;

i fuel assemblies. On the basis of this description of the CS system, the NRC l

staff notes that the estimated total leakage from the installation of the tie-rod assemblies and the postulated through-wall cracks in the horizontal welds and the V9 and V10 welds does not affect the performance of the CS system.

Because NMP1 is a BWR-2, it does not have jet pumps and does not depend upon reflood for a postulated recirculation line or steam line break. Thus, leakage of the shroud is not an issue for a recirculation line or steam line break since the core would be cooled by spray cooling. Therefore, ECCS performance is not affected by the increase in postulated leakage as a result of the V9 and V10 welds.

4.3.2 Detection NMPC evaluated whether the postulated through-wall cracking of the V9 and V10 welds would be detected as a result of changes in reactor operating parameters. The nuclear industry has established that shroud cracks may be detected if the estimated leakage produces a power anomaly of 2 percent in the rated power. The NRC staff notes that approximately 3-percent leakage is required to produce a 2-percent power reduction. For the estimated leakages discussed above, the combined leakage from (1) the installation holes from the tie-rod assemblies, (2) the postulated cracking in the-horizontal welds, and (3) the postulated cracking of the V9 and V10 welds would not be detectable.

Any change to other leakage indicators, such as recirculation loop temperature

'and core support plate pressure differential, would also be too small for the reactor operators to detect.

NMPC postulated that if vertical welds V-9 and V-10 did develop a larger leakage flow area, i.e., a crack opening of 1.5 inch for the entire length of the weld, the power decrease would be more than 2 percent and, therefore, would be detectable. NMPC maintains a Special Operating Procedure N1-SOP-2, Revision 5,

• Unexplained Reactor Power Change' that was- revised in 1994 to account for potential shroud cracking or displacement. In this procedure, depending upon where the indication is detected (i.e., below the core plate, between the core plate and the top guide, or above the top guide), the operator is instructed to either scram the reactor or cammence a normal reactor shutdown. At the NRC's public meeting on April 14, 1997 NMPC stated that operators were retrained on this procedure every 2 years.

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On the basis of the estimated leakage rates and the above information, the NRC staff concludes that shroud cracking in the horizontal or vertical welds would not produce effects that could be detected. However, if the crack openings were of significant size along the entire length of the welds (0.25 inches for all horizontal welds or 1.5 inches for both the V-9 and V-10 welds), the NRC staff concludes that these cracks would be detectable, and NMPC has suitable operating procedures in place to shut the reactor down.

5. 0. 3RR_0 TID STABILIZER ASSEMBLIES 5.1 Summary Description of the 1995 Shroud Repair In its SE dated March 31, 1995, the NRC staff approved NMPC's submittal of the shroud repair. -A summary description of the repair is given in the following paragraphs.

The shroud repair consists of four stabilizer assenblies (also called tie rod assemblies) placed at approximately uniform intervals around the shroud (azimuths 90',166',270',and350'). The stabilizer assemblies were designed to structurally replace the circumferential welds in the core shroud. Each assembly functions to vertically hold the shroud to the shroud support cone and to horizontally support the shroud at the top guide and core plate elevations. There are also other horizontal supports that would prevent unacceptable horizontal movement of any shroud cylindrical segment in the event the horizontal shroud welds should fail.

The stabilizer assemblies provide vertical restraint and represent an alternate load path between the top of the shroud and the shroud support cone.

This load path consists of the upper support, tie rod, C-spring, lower support, and toggle bolt. Differential thermal expansion from the different materials used for components of this load path, produces a thermal preload at plant operating conditions. Under postulated failures of the horizontal shroud welds, the thermal preload is sufficient to hold the cylindrical segments of the shroud in place for all normal operating conditions, such as the vertical upward force applied to the shroud by the coolant flow and pressure. The vertical load path is also de:igned to have a vertical saring

' rate that both prevents unacceptable vertical load during plant upset tiermal

> conditions and provides acceptable dynamic response during a plant seismic event.

The shroud is restrained horizontally by linear springs at the top guide and the core plate elevations. At the top guide elevation, the linear spring consists of the upper spring, upper wedge, upper contact, and the upper support. At the core plate elevation, the spring consists of the lower wedge, lower contact, and lower spring. The horizontal spring rates of these springs were designed to produce acceptable horizontal dynamic response during a seismic event. Tae horizontal displacement of the shroud during all events must be limited by these springs to ensu.re control rod insertion and prevent unacceptable leakage, in addition, in the event of failure of the horizontal shroud welds, unacceptable displacement of other cylindrical sections of the shroud would be prevented by displacement limiters, such as the mid-support and top support.

O During the initial installation of the repair hardware, the attachment features were machined into the shroud head and the shroud support cone. The lower support toggle bolts were inserted through the shroud support cone and tightened to 40 foot-pounds. In each assembly, the tie rod nut at the top was also tightened to a specified amount. The lower wedge was machined to provide a 0.01-inch compression of the lower spring.

The mid-support was machined to provide a 0.25-inch horizontal displacement of the tie rod.

The lower wedge is a component that was machined on the basis of actual site sneasurement to fit between the RPV and the lower spring with a small (0.010 inch)-comaression of the lower spring at room temperature. The lower wedge is heldinplacebyalatchdevice. The latch (also referred to as a retainer clip) is a wishbone-shaped piece that is intended to prevent relative motion between the lower wedge and the lower spring. Similar latches are also used to prevent relative-motion-at the mid-support-and at the upper spring. The lower support is an assembly that connects the shroud repair hardware to the l shroud support cone. The upper spring jacking bolt was 4.djusted to slightly

compress the upper spring. All moveable features were locked in place with mechanical devices such as crimps or spring retainers. The lower spring wedge latches are one type of spring retainer Thus, it was intended that, after installation, the upper and lower springs would be com)ressed between the shroud, and the RPV and the tie rod would be tight wit 1 a small preload.- The lower support and toggle bolt assemblies were tightened to the shroud support cone, but (as would be discovered during the spring 1997 outage) their position within the support cone holes had not been properly specified in the installation procedures.. They could be at any location within the hole depending on the hole's size and shape.

As stated earlier, the NRC staff reviewed NMPC's submittal relating to the shroud repair hardware and found it acceptable.

5.2 NRC Staff Evaluation of the Stabilizer Assemblies 5.2.1 Root Cause Evaluation of the Shroud Repair Deficiencies and Corrective Actions During the spring 1997 outage, NMPC performed post-operational inspections on the core shroud stabilizer (tie rod) assemblies, consistent with its commitments and NRC st.iff requirements. Deficiencies were found in the tie rod repairs, which included, to varying degrees, loose tie rod nuts on all four tie rods, damage to the retainer clips on the lower spring wedges, and mispositioned mid-supports. NMPC performed root cause evaluations, additional

' inspections, and testing of the tie rods. NMPC described these matters to the NRC staff in a letter dated April 8,1997, forwarding GE's analysis of the repair.

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i

The The tie rod nut at the 270' location was first discovered to be loose.

nut-locking device was normal and, in accordance with the original design and inst &l14 tion requirements, the not could not be moved without removing the locking feature. However, there was essentially no preload between the nut

' and the tie rod. After the locking feature The was removed, the nut was turned rotation of the nut before with less than 25 fact-pounds of torque. i tightening it at 25 foot-pounds was equivalent to an axialNMPC'sclearance of inspection and approximately 0.1 inch between the nut and the tie rod.

analysis confirmed that the nut had loosened because the toggle bolts had moved slightly within the oversized holes in the shroud support cone during plant operation. Although the holes are round when viewed perpendicularly from the shroud support cone, their projection on a horizontal plane is oval-shaped due to the angle of the support cone.

Thus, the toggle bolts are allowed to move up the inclined surface ot* the shroud support.

NMPC calculated the maximum tie rod looseness.that.could have been caused by movement.of the toggle bolts within the holes in the shroud support cones.

.The maximum upward movement (as measured by turning the tie rod nut on the 90' tie rod) was 0.151 inch; whereas, the differential thermal expansion in the tie rod is 0.155 inch, which is greater than the looseness at the shroud support cone. Therefore, although the initial installation mechanical preload Significantly was lost, some thermal preload would have remained.

rore thermal preload remained at the other three tie rod locations where the measured upward movement ranged from 0.054 inches to 0.093 inches.

Visual inspection revealed that the lower support wedge latch at the 90' location was broken. A piece of the latch that was missing Afterwas later found on examining the lower support cone at approximately azimuth 330'.

photographs and video tapes, NMPC stated that fatigue did not appear to be the failure mechanism. There was no visible evidence of plastic deformation, which would be necessary for a single-overload type of failure. The failure surface appeared to be consistent with a stress corrosion failure under high stress. Results from a metallurgical evaluation at a GE testing laboratory are expected to be available by May 20, 1997. Pending completion of those examinations, NMPC believes that the most likely failure mechanism As is the stress NRC corrosion, although failure from overload has not been ruled out.

staff understands this information, this conclusion appears reasonable.

Videotape inspection of the othat three h wer wedge latches indicated that they are all in one piece, but the latch at the 350' location appeared to be bent. In addition, the lower spring wedges showed minor evidence of local Since the latch material, alloy X-750, is hard contact with the latches.

harder than the lower spring wedge material, Type 316 low-carbon stainless steel, the lower spring wedges are likely to show surface wear before the latches.

A similar latch is used in each mid-support assembly, and two similar latches are used in each upper support assembly. NMPC observed the latches on the Because middle and upper supports, and all 12 appear to be in good condition.

of design differences, these latches cannot be loaded as severely as the lower wedge latches. The contact force between the RPV and the shroud repair is In much smaller at these locations than the contact force at the lower wedge.

addition, these latches are not loaded during plant heat-up. The lower wedge at the 90' location had dropped to the bottom of the post on the lower spring.

The lower wedge at the 350' location appeared to be approximately 1/8 inch below its normal position. The other two wedges were at their normal positions.

Contact sliding marks have been observed on the RPV wall at the 166* location above the lower and upper contact points of the upper spring assembly.

Sliding marks are not evident at any other contact points for the upper springs located at the 90', 270', and 350' locations. Visual inspection of all mid-support contacts confirmed that there was contact with the RPV surface in the cold condition. During normal operation, the mid-sup) ort compression on the RPV increases because of thermal conditions. Thus, t1e functionc of providing a load path from the tie rod to the RPV and increasing the natural frequency of the tie rod assemblies were maintained. During inspections and tests related to the root cause investigation of the lower support wedge latch failures, two mid-supports at azimuths 90' and 166' had to be repositioned.

New mid-supports have been fabricated and installed with the original design preload at these locations. The mid-supports at the 270' and 350' locations have also been verified for proper preload. Thur, the required support configuration will be maintained during future operation.

The root cause for the tie rod degradation affecting both the tie rod nut and the lower latch, is attributed to the movement of the toggle bolts within oversized lower support bolt holes. The installation procedures did not contain specific criteria for locating the toggle bolts during installation of the lower support. The lower support toggle bolts are nominally 4.000 inches in diameter. The measured electric discharge machining (EDM) holes in the shroud cone ranged from 4.090 inches to 4.110 inches Since the position of the lower support within the machined holes was not rocedurally controlled during installation, the relative position of the bo ts within the holes was rcndomly located.

During heat-up, the expansion of the shroud and tie rods generated a force sufficient to overcome the installation friction forces and slightly move the lower support toggle bolt assembly up the shroud cone. This translated into a

' vertical movement of the tie rod. loading the latch on the lower spring wedge and causing its failure. These latches were not designed to accommodate differential movement between the RPV wall and the lower spring wedge during normal and transient conditions in the event the lower wedge got stuck at the RPV wall. This also caused the loss in preload on the tie rod nut, described earlier. The NRC staff has /eviewed NMPC's analysis of the tie rod degradation and agrees with NMPC's conclusions, in the original design of the lower wedge, it was assumed that the wedges would slide on the RPV wall and accommodate differential thermal expansion betwecn the tie rod assembly and the RPV. However, the actual frictional force between the wedge and the RPV was higher than anticipated and sufficient to prevent movement of the wedges at the RPV interface during thermal growth of the tie rod assembly. This caused the retainer clip (or latch) to stretch

0 A portion of the between its attachment points on the tie rod and the wedge.

retainer clip was overstressed and failed because of stress corrosion cracking, i On the basis of its review (as discussed above), the NRC staff believes However, NMPC's even if the preliminary root cause determination is reasonable. failure resulted fro address the failure.

j 5.2.2 Modification of the Lower Support Latch Design The lower support latch (or retainer clip) has been redesigned to accommodate The redesigned latches were movement during normal and transient conditions.The new design is substantially more installed before core reload.

and takes into account the various potential sliding cases. These cases include combinations of sliding at the RPV wall / lower wedge and lower wedge / lower spring interfaces. The largest latch displacement is postulated during the initial heat-up and hydrotest where the wedge is projected Duringtoa slide at the spring interface but remain locked at the RPV interface.

potential cooldown to a ambient conditions the wedge is projected to slide only at the RPV interface, finally, during a heat-up to full power, the wedge is assumed to slide a the spring interface with no movement at the RPV surface. The calculated latch displacement for this worst-case scenario has been determined to be 0.182 inches. The calculated stress in the new latch for this displacement has been determined to be 60 percent of the allowable stress. In addition, the latches have been evaluated for potential damage due to 500. Utilizing a Stress Rule Index (SRI) methodology, NMPC has demonstrated that for the worst-case sliding, stress corrosion will not occur in the next operating cycle. However, if a most probable wedge /RPV/ spring interface sliding scenario is assumed, then no corrosion is projected for the remaining life of the plant.

5.2.3 NRC Staff Evaluation of the Structural Analysis of the Redesigned Lower Support Latch On the basis cf itt review of the new retainer clip design, the NRC staff One concern was that the slightly sloping surface on raised several concerns.

the lower wedge would preferentially cause the wedge to slide in one direction and get stuck in the other direction. NMPC, in response, clarified that the purpose of the angle on tne lower wedge is to facilitate moving the lower This wedge into position and creating the contact force with the RPV wall.

slight slope on the wedge does not influence whether sliding occurs during Therefore, at the wedge / spring interface or the RPV wall / wedge interface.

plant operation, sliding isAsexpected to occur at whichever surface has thesliding is discussed in GENE Report 813-01739-22, lowest friction factor.

likely to occur at the wedge / spring interface because the machined surfaces and dissimilar materials will result in a low friction factor.

In response to the NRC staff's inquiry about the feasibility of lowering the lateral load at the wedges to facilitate sliding, HMPC discussed the The considerations in determining the lateral design loads on the wedges.

intent of the lower spring design was to ensure that positive contact exists

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during shutdown conditions, and the 0.010-inch radial contact at the lower spring represents a minimal condition that meets this requirement.

Recognizing that the ma surements for matching the lower wedge have to be taken remotely from the reactor refueling floor, the 0.010-inch interference represents a value that confidently ensures contact at the lower spring support. Without radial support at the lower wedge, the only potential problem is vibration. However, as demonstrated at the 90' tie rod location ,

during the last o>erating cycle (which lost contact), and by analysis. NMPC i concluded that viaration of the tie rod assembly is not an issue. NMPC's i responses as discussed above, reasonably address the NRC staff's concerns, i lhe maximum possible differential vertical displacement of the lower wedges and the )robable wedge movement has been determined. Since the potential exists t1at the wedges may not always slide at the spring interfaces due to unanticipated forces on the wedja'during various operational transients, the l design of the latches has been based upon the maximum estimated vertical i displacement. The maximum displacement assumed for the latch design exceeds the maximum vertical displacement of the tie rod hardware at the lower spring elevation as shown in Table 4.1 of GENE Report B12-01739-22.

During certain plant operating and test conditions Te.g., during a hydrotest),

the radial contact force on the wedges is likely to be minimal. Under such conditions, the wedges could sotentially slide circumferential1y along the RPV surface and remain stuck in t1at position. In response to NRC staff concerns that this could impose an additional torsional moment on the latch Juring subsequent plant operation, NMPC indicated that these considerations have been L factored into the new latch design. The ratio of the width of the wedge to the potential gap between the wedge and the RPV wall is very large.

Therefore, potential angularity of the wedge in a loss-of-contact event is extremely small and any applied force on the wedge will cause it to readjust the or_tentation of the wedge to distribute the contact forces across the entire face of the wedge surface, and not allow an edge to become stuck. Even if it is assumed that the wedge would become stuck on one edge, the latch mechanism has lateral clearance with the wedge and the spring interface slots, so that the latch will reposition itself without any torsional loads being applied.

The new latch design stresses have been calculated to be substantially lower than the original latch design stresses during various postulated operating, transient, and accident conditions. The membrane plus bending stresses in the new latch design are lower by a factor of 8.6 in comparison with the original latch stresses. The corresponding reduction factor for the membrane, bending, and peak stresses is 12.8. This indicates a substantial improvement in the design margin for the new latches.

5.2.4 NRC Staff Analysis of Shroud and Tie Rod Assembly Section 4 of GE Report GENE B13-01722-40, " Shroud Repair Anomalies--Nine Mile Point Unit 1. RFul4," dated April 1997, contains an evaluation of the safe operation of the shroud and the tie rod assembly with the existing ligaments in the horizontal and vertical welds. NMPC's assessment indicates that the safe operation of NMP1 was not impaired with the as-found condition of the tie

rod assembly and an assumption that all horizontal welds w wall cracked.

the regions where inspection has been performed, no through-wall Therefore, cracks or 360' crack lengths had been found in any of the horizontal welds.

NMPC's evaluation is considered conservative.

NMPC has analyzed the as-found condition of the shroud A conservative interval vertical welds en established that the olant can be operated safely. NMPC has committed to for reinspection of tie welds has been established.

reinspect the tie rods, which will include checking the latch devices and checking the tie rod nuts for tightness,f the structural integrity of the re-operation. On the basis of its review o designed latch and the core shroud vertical welds discussed earlier, the staf f finds the inspection interval acceptable.

5.2.5 Revised Installation Procedures for Tie Rod Assemblies

" Lower Wedge Latch As described in NMPC's Document NMP-SHD-003, Revision 1 Replacement and Tie Rod Torque Checks," including Special Process C Sheet SPCS#01, Revision 1, NMPC has developed improved installation p to ensure that the tie rod assemblies are installed with no gaps at th and of the support cone holes. rod be jacked Jacks areat three placed locations under the during tie gaps associated with installation tolerances.

lower support, on the RPV side of the lower support, to push it um the sh cone.

This removes the clearances between the toggle bolts and tie shroud side of the cone holes.

In its letter to the NRC dated April 27, 1997, " Responses to NRC Staff Questions Provided During Telephone Conversations of April 10,11, and 1997, on Core Shroud Cracks and Repair," NMPC states that there are n operating, transient, or test conditions that will cause the toggle bo slide down the shroud support cone.

would have to be higher t1an the RPV and shroud temperatures, This in addition comalete loss of preload in the tie rods to cause such a movement.

,comsination of events does not occur during operation.

After installation, and in accordance with the revised procedures, inspecti were completed on each tie rod assembly to verify oroper contact by ver the absence of gaps. These inspections revealed t1at the middle support wasThis resu  ;

no longer in contact with the RPV on the 90' and 166' tie i The middle support dimensions were obtained and No new middle su j installed before reload.  !

potential for gaps and non-conforming conditions were inspected.

additional deficiencies were noted.

6.0 FUTURE INSPECTIONS in its lettar of April 30, 1997, NMPC also stated that it will propose an inspection plan for the next scheduled outageThis andplan submit should the plan at least 3 months before the outage is scheduled to begin.

0 23-provide details regarding the inspection of the shroud repair components; the shroud repair anchorages; and the shroud's horizontal, vertical, and ring segment welds. The plan will specify the inspection methods to be used, including the provisions for sample expansion.

7.0 CONCLUSION

7.1 Inspection The NRC staff has determined that the scope and methods used for the inspections of the NMP1 core shroud during the spring 1997 refueling outage met or exceeded the requirements of BWRVI)-07, and are, therefore, acceptable.

7.2 Structural Integrity The NRC staff has concluded that results of the analyses show that the core shroud maintained the required ASME Code margins during the last operating l cycle. The review considered the loss of pre-load for the tie-rod repairs.

Additional analyses submitted by the licensee pursuant to 10 CFR 50.55a implemented by ASME Code Section XI for the vertical welds have been shown to have sufficient ligaments to support the tie-rod repair for at least 10,600 additional hours of operation. The staff performed an independent analysis and determined that the licensee's analyses appear reasonable.

7.3 Tie Rod Repair On the basis of its review of the root cause analysis of the shroud repair anomalies, the NRC staff concludes that the tie rod was loosened by the movement of the toggle bolts within oversized bolt holes. NMPC's new installation procedures include measures to prevent tie rod looseness and maintain tie rod vertical forces as intended in the original design. The root cause of the latch failure was larger-than-anticipated vertical displacements of the 1:'.ch, which overstressed the latch and most likely subjected it to SCC. The new latch has been redesigned to accommodate larger vertical

,d isplacemen t s while maintaining its original function of locking the wedge to the lower spring structure. On the basis of a most probable sliding scenario at the wedge /RPV interface, no failure is projected for the remaining life of NMPl. The calculated stresses are within the ASME Code allowable values and the latch has been analyzed to be resistant to stress corrosion for a minimum of 2 years assuming worst-case displacements in the latch. HMPC's root cause evaluation and corrective actions offer reasonable assurance that the tie rods will perform their intended function. The root cause evaluation will be confirmed in the near term by ongoing evaluations. Future inspections will monitor the conditions of the tie rod assemblies.

The NRC staff has determined that the modified tie-rod repair is an accept Sie alternative to a repair in accordance with ASME Code Section XI, pursuant to 10 CFR 50.55a(a)(3)(i). The alternative provides an acceptable level of quality and safety because the ASME Code margins of safety will be maintained for the operating period. This NRC approval under 10 CFR 50.55a(a)(3)(i) is contingent upon NMPC (1) maintaining reactor coolant chemistry within the il

i I

guidelines set forth in Electric Power Research Institute technical report TR-103515, BWR Water Chemistry Guidelines-1996 Revision," in accordance with its commitment by letter dated April 30, 1997, and (2) submitting, within 60 days, an application for a license amendment that addresses this matter in to failure accordance with its commitment by letter dated May 7, 1997.

satisfy either of these conditions will render this approval null and void.

8.0 BIBllOGRAPHY

1. Letter from L.B. Marsh, U.S. NRC to Niagara Mohawk Power Corporation dated March 31, 1995, and attached Safety Evaluation by the Office of Nuclear Reactor Regulation relating to the Core Shroud Repair at Nine Mile Point Unit 1.

l

2. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 0927),

dated March 25, 1995. ,

i

3. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1067),

dated April 30, 1995.

4. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL lill),

dated August 14, 1996, with enclosure providin9 final design documentation for modifications to the core shroud repair assemblies installed in March 1995.

S. Letter from NRC to Niagara Mohawk Power Corporation,-dated March 3 1997, ,

with enclosed Safety Evaluation by the Office of Nuclear Reactor Regulation relating to the Modifications for Correcting Deviations identified in Core Shroud Repairs at Nine Mile Point Unit 1 (TAC No, M90102).

6. Letter from Niagara Mohawk Corporation to NRC, dated January 6,_1995, with enclosed design report of Core Shroud repair hardware.

Supplemented by letters dated January 23, 26, February 14, 24, 28, March 7, 9,13 (2 letters),14 (2 letters), 23, 27, 28 and 30 1995.

7. Lotter from Niagara Mohawk Corporation to NRC, NMPIL 1200 dated April 8, 1997, (Enclosure 2) GENE B13-01739-40 " Shroud Repair Anomalies Nine Mile Point 1, RF014,' April 1997.

8. GENE-B13-01739-22, Revision 0, Nine Mile Point 1, " Design Report for Improved Shroud Repair Lower Support Latches," April 1997.

9. Niagara Mohawk Document NMP-SHD-003, Revision!!, " Lower Wedge Latch Replacement and Tie Rod Torque Checks," including Special Process Control Sheet SPCS#01, Revision 1.

10. Safety Evaluation of Niagara Mohawk Power Corporation (NMPC)

Reinspection Plans for Nine Mile Point Unit 1 Core Shroud, NRC, March 3, 1997.

.. o

11. "BWR Vessel and Internals Project. Reactor Pressure Vessel and Internals )

Examination Guidelines (BWRVIP-03)," EPRI Report TR-105696, October 1995.

12. "BWR Vessel and Internals Project, Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07)," EPRI Report TR-105747, February 1996.

13. " Reactor Pressure Vessel and Internals Examinations Guidelines (BWRVIP-03)," EPRI TR-105696, October 1995.

14. " Assessment of the Vertical Weld Cracking on the NMP1 Shroud," Report No. GE-NE-523-B13-01869-043 Rev. O, April 1997. Prepared for Niagara Mohawk Power Company by GE Nuclear Energy.  ;

l

! 15. " Shroud Repair Anomalies,-Nine Mile Point Unit 1, RF014," Report No.

B13-01739-40, Rev. 0, Class !!!, April 1997. Prepared for Niagara Mohawk Power Company Ly GE Nuclear Energy.

16.

Nine Mile Point Unit 1 Core Shroud Cracking Evaluation," Presared for George Inch, Niagara Mohawk Power Company, by Richard E. Smit 1, Altran Corp., April 3, 1997.

17. " Analysis of Nine Mile Point Unit 1 Shroud Welds V9 and Weld V10 Cracking," April 1997, Report No. MPM-497439, MPM Technologies.

18. Safety Evaluation 96-018, pursuant to 10 CFR 50.59 Rev. 1.

19. " Concerns Regarding Operation of Nine Mile Point 1," letter dated April 9, 1997, from D. Lochbaum, Union of Concerned Scientists to S. Bajwa, NRC.

20. " Potential Unanalyzed Operation of Nine Mile Point Unit 1 Core Shroud Vertical Cracks," letter dated April 17, 1997, from D. Lochbaum, Union of Concerned Scientists to S. Bajwa, NRC.

,21. "BWR Core Shroud Inspection and Flaw Evaluation Guidelines," GENE-523-113-0894, Rev. 1. General Electric Company, March 1995. Prepared for the BWRVIP Assessment Committee.

-22. "BWR Water Chemistry Guidelines, 1993 Revision, Normal and Hydrogen Water Chemistry," EPRI TR-103515 February 1994.

23. " Response to April 17, 1997, letter from Union of Concerned Scientists to the NRC," letter dated April 25,=1997 - from HMPC to NRC.

24. " Response to Union of Concerned Scientists letter dated April 9,1997 Reg.trding Nine Mile Point Unit 1," letter dated April 26, 1997, from NMPC to NRC.

25. " Responses to NRC Staff Duestions Provided during Telephone Conversations of April 10, II,.and 24, 1997, on Core Shroud Cracks and Repair,"11etter dated April 27, 1997, from NMPC to NRC.

26.

• Responses to NRC Staff Questions Provided during Telephone Conversations of April 22 1997, on Core Shroud Cracks and Repair.'

letter dated April 27, 1997, from NMPC to NRC.'

27. Brunswick Steam Electric Plant No.1, Docket No. 50-325/t.icense No. DPR-71 NRC Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," letter dated June 21, 1995, from R. Anderson of Carolina Power and Light Co. to the NRC.

28. Letter from Niagara Mohawk Power Corporation to U.S. NRC (NMPIL 1217),

dated May 7, 1995.

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Figure 12 NMF1 Shrend Wald Locations I

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, _ _ , . . . , . . . ~ . . . . . . , _ , , . . _ . _ , . _ _ _ , . . . , _ , , _ . . . . , _ . . _ , , _ _ . _ , . _ . , . . , . . , . . . , , _ . _ , . _ _ . __ ,..__,.._..._,...._,,_.-m., _,

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,o cut.sn vises.on n o Table 2 2 '

Ennunary of Recent Shrend Vertical Weld laspections (RTO 14)

Mid %and lasportes shroud Eass Type Fnam 14egta 14sgth Coverage

• ID.OD us)

• V.) 31.25 15'Le8 OD UT IJ' ID, Nght HAZ Ise arpt .

V.4 31.25 22

• lea 0.I*OD.NahtRAZ '

OD UT 22 *ID lea MAZ,  :

11* Naht V.5 ring 1J'!D Nsht RAZ Not located NA NA  ;

NA V4rar Not !-1 NA NA NA V.1 alJ 9* Le8 OD UT Nola6 amens 11' N abt ,

i V4 18.5 5.5 " L4A OD UT Nolahannens 9.$* N aht V.9 90.12 100% ID and EVT.)

shell lahannens se ever 90% OD OD right HAZ Minor ernaking on OD neA side B0* and on ID both sides OD UT Numerous a6mnons se OD, 14A HAZ .

Two minor As*1 on ID, Nght V.10 HAZ 90.12 ION ID and EVT.) Crar.kas se OD, Ash MAZ~ ~

CD i Crasking se ID,LaA and Nght g4" MAZ OD UT Flaws dneced on > 80% on OD, Nght RAZ Flowi dnamed on > 10% on V.Il 63.$ OD L4A MAZ 10W OD ID and EVT.1 Nolassanons

$0%TD OD V 12 63.5 100% OD ID and EVT.1 6' OD, hght HAZ *

$0%ID OD V.15 22.13 11

• LaA OD a UT 6* ID, LaA MAZ .

11' N ght V.16 22.13 2.2

• ID. Naht MAZ 100 % OD EVT.1 '

5* OD.LaA MAZ

' 10.5

• L4 A OD UT $* ID L4A HAZ 20' Nght '

d' ID Nght HAZ 3* ID148 HAZ 6am right side esem Won coveragt may be less Iban in$cated, ligammt length. .

but has boss akan in somsidenk bence tbt awgmed e

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Appendia C -

Shroud Inspection Summary l

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The following is a weld by weld summary detailing the scope of inspections and results of the shmud ev=inations performed to date.

Wald V 3 Perfonned ultrasonic eraminadon of approximately 15 inches of each side of the weld from the abroud OD surface. Approximately 1.5" of Gaw was detected on the ID surface and 0.8" of flaw onthe OD surface. .

Wald V-4 Performed ultrasonic - -taalon of approximately 11" of the leh HAZ aml 22" of the' right HAZ.

ID flaws wm detected over the entire examined length of the left HAZ and 1.5" of flaw was detected on the ID of the right HAZ.

Wald V-7 s

Performed ultrasonic examinadon of approximately 9" of the left HAZ and 11" of the right HAZ.

No flawntre detected during the erammation.

W eid V 8 Performed ultrasonic examinadon'of approximately 5.5" of the left HAZ and 9.5" of the right

- HAZ. No flaws were detected during the exami tation.

Weld V 9 Perfanned ultrasonic examination from the shroud OD surface for approximately the entire length of both the left and right HAZs as well as EVT from both the ID and the OD. Visual cracking was detected over greater than 90% of the right HAZ on the OD anti minimal cracking was detected on the ID in both the left an tight HAZs. Minor cracking was also detected on the OD in the left RAZ.

The cracks detected visually on the shroud ID surface were found to be predominantly transverse to the weld whereas the cracFing detected visually on the shroud OD surface was mostly parallel to the weld with components that branched transverse to the ' weld. Ultrasonic ===minadons of

==<rnei=11y the entire length of the weld was performed from the shed OD surface and detected numerous flaws over the length of the left HAZ emanating from the abroud OD surface. Two small flaws on the ID curface were detected in the right HAZ.

Wald V 10 -

Performed ultrasonic examination from the shroud OD surface for approximately the entire length of both the left and right HAZs as well as EVT from both the ID and the OD. Flaws were detected -

en greater than 80% of the right HAZ on the OD surface and greater than 50% of the left RAZ revealed flaws on the OD surface. The EVT evamination revealed cracking in the left and right C2

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' ct.>,'t.sn-ctsews n o HAZs on the OD surfue for most of the length of the weld and' on the D in both the left and right HAZs. The cruks detected visually on the shroud ID surfee were found to be predominantly transverse to the weld whereas the cruking detected visually on the shroud OD surface was mostly parallel to the oeld with components that b:snebed transverse to the wtld.

Weld V.11 .

. EVT vaiwiens were performed on the ecessible weld length from both the ID and the CD of both the ich and right HAZs. No cruking wu detected during the ev=miwien.

Wald V 12 EVT

• amtwiens were performed on the accessible weld length from both the ID and the OD of both the len and right HA2s. One 6" eruk was detected on the length OD surfue in the right HAZ. No other cruking wu detected.

Weld V-15 Ultrasonle examination was performed imm the shroud OD swfee on approximately 11 inches of both the leR and right HAZs. One 6" flaw wu detected in the left HAZ on the ID surfue and several ID flaws totaling 2.2" in total length was detected on the ID in the right HAZ. No flaw detected in either HAZ was greater than 10% through wall.

WeldV.16 Ultrasonic examination was performed from the shroud OD surfue of approximately 10.5" ofleft HAZ. Two flaws were detected on the ID surfue. One flaw was 5" in length,10% through wall.

The other ID flaw in the left HAZ wu datected from the scan on the right HAZ and wu 3" long and 30% through wall. Apprcximately 2 inches of the right RAZ was examine from the shroud OD surface. One flaw was detected on the ID which measured 4" in length and 21% through wall.

An EVT evnmintion of both HAZs from the shmud OD surfue revealed one cruk in the left HAZ.

Recent inspection Results for Shroud Horizontal Welds In addition to the shroud vertical weld inspections, the horizontal welds H 2, H 4, H 5, H 6a, H-6b, and H 7 were also inspecte4 for analytical purposes, to evaluate the overall integrity of the shroud using assumptions of worst case cra: king of the vertical welds.

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L C/ eld H.2 U)nssonic namI= tion was performe$ from the shroud OD surface of approximately 24 inches cf the upper HAZ adjacent to weld V 4. Approximately 7 inches of intermittent flaws wm detected on the OD surface, with the deepest area having a through wall depth of.22 inches.

W ald U d

===t= don from the shroud OD surface was performed on approximately 60% of the Ultrasonic lower HAZ. ID anNor OD flaws wm detected intermittently throughout the ammi= don area.

Some ID flaws were dessetad in the upper HAZ. Approximately 32 inches of the upper HAZ was ultrau;sically c==t=A. 3 inche.s of shallow OD flaws wm detected in the upper HAZ and one 6 inch long ID flaw was desseted with the maximum through wall depth of.23 inches. . An EVT enmintion of the OD was performed of over 70% of the upper and lower HAZs. Cracks were detected in both the upper and lower HAZs. .

W ald H 5 Ultrasonic examination from the shroud OD surface was performed on approximately 30% of the upper and lower HAZs. OD and ID flaws wm detected in the upper HAZ only. No flaws were detected in the lower HAZ. EVT of approximately 60% of the shroud OD surface revealed cracks interminently in both the upper and lower HAZs. Most of the flaws detected visually on the OD surface wm oriented perpendicular to the weld. No flaws were dessetad in the upper HAZ at the intersections ofwelds V9 or V10.

Weld H4A Ultrasonic enmintion was performed on both the upper and lows: HAZs of approximately 30%

of the circumference from the shroud OD surface. Flaws wm detected on the OD surface of the lower HAZ only. No flaws were detected in the upper HAZ or on the ID of either HAZ.

Wald H4B ditrasonic examination was performed on both the upper and lower HAZs of approximately 30%

of the circumference from the shroud OD surface. Flaws were detected on the OD surface of the upper HAZ only. No flaws were detected in the lower HAZ or on the ID of either HAZ.

W ald B 7 Ultrasonic examination was performed for the shroud OD surface on the upper HAZ on approximately 30% of the circumference. No flaws wm detected during the namintion.

Wald H-8 .

Ultrasonic examination was performed for the shroud OD su: face on the lower HAZ on approximately 30% of the circumference. A flaw which was identified by UT during a prior C4

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• GE.NE. DIS.01sJs-04). An o .

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outage was located as well as one additional Daw in the name area. This Daw wu tdtrasonically sized to be oflesser through wall depth than in RFOl3. A rniew of the pmious data indicates that the previous sizing performed _was very conservative. An EVT was performed on approximately 30% of the circumference from the shroud OD surface. Of the five small cracks visually detected during RF013 only 1 wn: visible during this inspection. The inspection in the area of the other four was hampered by the placement of a Tie Rod support mbich prevented a j good EVT inspection. Cracks were Sisually detected in three new locations in the upper HAZ, The 1srgest of these cracks (9" 12") is located predominantly in the ring segment Upper HAZ and runs >

into the wild toe and back lato the ring segment. .

i j W ald B 9

An EVT ev=midon was performed in one arsa 26 inches long. No indications were noted during l the mmidon.

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• cz.hwumws.n o Tableh2 ADowable F1sw Shas for the Nine Mile Polat Unit 1 .

j Shrood VerttaalWalds '

(1) (2) 9) (4) (8)

Benlana (6)

Wald Min.1Jgsmest Availabk Asewshie Thrwegt wan regelred Wald ID Imagth, Ameletlag ernek ,agakansat erack length,In. Ngsmaat,nn.

As growth (rwe Useraskad years)and' 14smest laspeedes langth,is, s.rnd sanntend '

Unemataty,la.

V.3, V.4 J1.25 . (Note 1)

, 29.97 1.24 J.63 7J (V 3)

W7, W4 18.50 18J Note 2 N.4) 17.78 0.72 3.07 9.0 (V 7)

W9, 90.12 13.40 S.6 N 4) 46.61 14.72

% 10 17.07

' Note 2 (V.9)

%)1, 63.50 38.20 Note 2 N.10) 41.03 4 440

%)2 9JO l 31.75 (V.)1)

%15, 22.13 Note 3 25.1$(W12) 19.53 ~ 2.46 4.41

% 16 Nou 4 (V.15)

$J N.16)

Notes 1, Based on crack grmrth of1.6 tr and UTinspection sneertalaty of 2 x 0.375 inch at each ersck tip forlength sizing. - *'

. 2. Meets regninments based on further evaluation reported la Esbsection 5.3.

3. The minimum ligament for EVT inspection h larger to.secount for greater uncertainty in the visual inspection. The ascertalaty 2netor applied k equal to2x1.2in. - ,

4. De egnkalent length after subtreeting ersck growth and inspection sneertainty k 2.39 in.which h greater than the required ligament of1,46 in.

and thus neseptable.

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