ML20196A336

Forwards Suppl RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions, for Plant,Units 1 & 2.Licensee Will Respond within 60 Days of Receipt of NRC Ltr

ML20196A336
Person / Time
Site:	Nine Mile Point
Issue date:	11/19/1998
From:	Hood D NRC (Affiliation Not Assigned)
To:	Mueller J NIAGARA MOHAWK POWER CORP.
Shared Package
ML20196A342	List: ML20196A336 ML20196B944 ML20196B980
References
GL-96-06, GL-96-6, TAC-M96836, TAC-M96837, NUDOCS 9811270116
Download: ML20196A336 (13)

Text

i g j NUCLEAR REGUL.ATORY COMMISSION WASHINGTON, D.C. 30866-0001 l

k..... November 19, 1998 l

l

Mr. John H. Mueller VIAEM Chief Nuclear Officer

~

Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Operations Building, Second Floor P.O. Bcx 63 Lycoming, NY 13093

SUBJECT:

SUPPLEMENTAL REQUEST FOR ADDITIONAL INFORMATION REGARDING GENERIC LETTER 96-06, NINE MILE POINT NUCLEAR STATION, UNIT NOS.1 AND 2 (TAC NOS. M96836 AND M96837)

Dear Mr. Mueller:

The NRC staff is reviewing your submittals of October 29,1996, and January 28, February 7, and December 16,1997, responding to Generic Letter 96-06," Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions," for Nine Mile Point Nuclear Station, Unit Nos.1 and 2 (NMP1 and NMP2). The NRC's staff's review also includes Gupplement i to NMP2 Licensee Event Report (LER) 96-16 forwarded under cover of a letter dated February 24,1997, and Supplement 1 to NMP1 LER 96-13 forwarded February 28,1997.

In addition to the information requested in our letter dated June 1,1998, (to which Niagara Mohawk Power Corporation (NMPC) responded August 28,1998), we find that additional information requested in the enclosure is needed to complete our review.

As shown in Enclosure 2, on October 2,1998, a draft version of Enclosure 1 was faxed to ', j NMPC, along with a copy of our October 1,1998, letter to Mr. D. Modeen of Nuclear Energy Institute, regarding review of Electric Power Research Institute's Technical Report TR-108812,

" Response of isolated Piping to Thermally Induced Overpressurization DurHg a Loss Of Coolant Accident." The fax requested a phone call to discuss the response date. Based on the phone call of November 4,1998, and Mr. Leonard's subsequent e-mail of November 5,1998 (Enclosure '

2), the following mutually agreeable response schedule was determined for NMP1:

1. For the three NMP1 penetrations already modified, NMPC will respond within 60 days of receipt of the NRC's letter (i.e., by January 20,1999).

2. For the seven NMP1 penetrations remaining to be addressed, NMPC will respond by September 30,1999. NMPC plans to address five of those remaining penetrations by analysis and two by modifications (to be installed at refueling outage 16).

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J. Mueller I Based upon an e-mail dated October 22,1998 (Enclosure 2), Mr. Leonard's response for NMP2 resulted in a mutually agreeable response schedule. NMPC will respond within 60 days of receipt of the NRC's letter (i.e., by January 20,1999).

If you have questions regarding the enclosures or if you are unable to meet the established response date, please contact me by phone at (301) 415 3049 or by e-mail at dsh@nrc. gov.

• l- Sincerely, D A ttr 4 Darl S. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-220 and 50-410

Enclosures:

1. Supplemental Request for Additional Information

2. Correspondence Regarding Response Date cc w/encis: See next page l

l-I i

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l l

• J. Mutiltr ' November 19, 1998

. . i BisId upon en e-miil ditId October 22,1998 (Enclosurs 2), Mr. Leonrrd's responsa for NMP2 i resulted in a mutually agreeable response schedule. NMPC will respond within 60 days of l receipt of the NRC's letter (i.e., by January 20,1999). i If you have questions regarding the enclosures or if you are unable to meet the established response date, please contact me by phone at (301) 415-3049 or by e-mail at dsh@nrc. gov.  !

Sincerely, ORIGINAL SIGNED BY:

Darl S. Hood, Senior Project Manager -

Project Directorate 1-1 Division of Reactor Projects - l/ll Office of Nuclear Reactor Regulation Docket Nos. 50-220 and 50-410

Enclosures:

1. Supplemental Request for Additional 1 Information I

2. Correspondence Regarding Response Date

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l cc w/encls: See next page i DISTRIBUTION: ,

Docket File SLittle OGC  !

PUBLIC . KManoly ACRS I PDI-1 R/F DHood CCowgill, R1 JZwolinski BWetzel GHammer SBajwa JTatum DOCUMENT NAME: G:\NMP1-2\N1296836.RA3 To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy

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OFFICE PM: POI-1 lE LA:P01-3ll l 0:P01 1 _ , , l l uAME onood/r t h Jly stietF ss.Jw E p-DATE 11/)T/98' 11/ N /98 11/ /(/f8' Official Rec 0rd Copy i

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. , _ _ ._. ___ . _ _ . _ ~ _ _ . . _ . _ . _ _ _ . . _ _ _ . . _ _ _ . _ . _

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John H. Mueller Nine Mile Point Nuclear Station Niagara Mohawk Power Corporation Unit Nos.1 and 2 j cc:

j Regional Administrator, Region l Charles Donaldson, Esquire

. U.S. Nuclear Regulatory Commission Assistant Attorray General 3

475 Allendale Road New York Depart. rent c aw King of Prussia, PA 19406 120 Broadway New York, NY 10271 4

Resident inspector

U.S. Nuclear Regulatory Commission Mr. Paul D. Eddy P.O. Box 126 State of New York Department of i Lycoming, NY 13093 Public Service i Power Division, System Operations Mr. Jim Rettberg 3 Empire State Plaza New York State Electric & Gas - Albany, NY 12223 Corporation Corporate Drive Mr. Timothy S. Carey Kirkwood industrial Park Chair and Executive Director i P.O. Box 5224 State Consumer Protection Board i Binghamton, NY 13902-5224 5 Empire State Plaza, Suite 2101 ,

s Albany, NY 12223 l

Supervisor j Town of Scriba Mark J. Wetterhahn, Esquire l Route 8, Box 382 Winston & Strawn Oswego, NY 13126 1400 L Street, NW 2 Washington, DC 20005-3502 Mr. Richard Goldsmith )

Syracuse University Gary D. Wilson, Esquire College of Law Niagara Mohawk Power Corporation i E.l. White Hall Campus 300 Erie Boulevard West '

Syracuse, NY 12223 Syracuse, NY 13202 Mr. John V. Vinquist, MATS Inc. Mr. F. William Valentino, President P.O. Box 63 New York State Energy, Research,

)

Lycoming, NY 13093 and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 l

I i

l

SUPPLEMENTAL REQUEST FOR ADDITIONAL INFORMATION l REGARDING RESPONSE TO GL 96-06 l NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION. UNIT NOS.1 AND 2 i DOCKET NOS. 50-220 AND 50-410 i

i The NRC staff is reviewing responses by Niagara Mohawk Power Corporation (NMPC) to j Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity

, During Design-Basis Accident Conditions." In its February 7,1997, letter, NMPC identified 10 i Unit 1 pipe segments and 21 Unit 2 pipe segments that were susceptible to thermally-induced

! pressurization and were evaluated for operability. The NRC staff requests the following additionalinformation regarding NMPC's evaluation of these pipe segments:

j 1. Provide summaries of NMPC's planned corrective actions for the 31 pipe segments l determined to be susceptible to thermally-induced pressurization.

2

2. For any of NMPC's modifications that involve heat transfer and/or structural analyses of the pipe sea.nents, provide the following information for these segments:

j 2.1 Specify the applicable design criteria for the piping and valves. This should include the required load combinations; 2.2 Submit a drawing (or identify a previously docketed drawing) of the piping run between the isolation valves. This should include the lengths and thicknesses of the piping segments and the type and thickness of the insulation; 2.3 Provide the maximum-calculated temperature and pressure for the piping run.

Describe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model and the basis for the heat transfer coefficients used in the analysis.

1 l

Enclosure 1

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pa arooq p t UNITED STATES s* j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. ma m 4

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FACSIMILE TRANSMISSION DATE: 6 A Z,/fffr TO: [ M // M FAX NO: 3 /5- 3 t+f- / + c' o TEL NO: 315- 3 +y . y29g FROM: OdM/M U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION FAX NO.: (301) 415-2102 TEL NO: 3c/- 4/S- 3 o47 PAGE 1 OF PAGES 7 REMARKS: )

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Enclosure 2

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 0001 l Mr. John H. Mueller Chief Nuclear Officer -

Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station I

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r Operations Building, Second Floor P.O. Box 63 Lycoming, NY 13093

SUBJECT:

SUPPLEMENTAL REQUEST FOR ADDITIONAL INFORMATION REGARDING l . GENERIC LETTER 96-06, NINE MILE POINT NUCLEAR STATION, UNil NOS.1 AND 2 (TAC NOS. M96836 & M96837)

Dear Mr. Mueller:

~

! The NRC staff is reviewing your submittals of October 29,1996, and January 28, February 7, and

! December 16,1997, responding to Generic Letter 96-06, " Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions," for Nine Mile Point Nuclear i

Station, Unit Nos.1 and 2 (NMP1 and NMP2). Th. staff's review also includes Supplement 1 to NMP2 Licensee Event Report (LER) 96-16 forwarded under cover of a letter dated February 24, 1997, and Supplement 1 to NMP1 LER 96-13 forwarded February 28,1997. In addition to the information requested in our letter dated June 1,1998, we find that additional information requested in the enclosure is needed to complete our review.

The enclosure was discussed with Ms. D. Wolniak and other members of your organization during a telephone conversation on .1998. Ms. Wolniak stated that the response to this letter would be submitted within days of its receipt. Accordingly, your response is expected by .1998.

If you have questions regarding the enclosure or if you are unable to meet the committed response date, please contact me by phone at (301) 415-3049 or by e-mail at dsh@nrc. gov.

l Sincerely, i

RA FT Dart S. Hood, Senior Project Manager Project Directorate I 1 l Division of Reactor Projects - t/11 Office of Nuclear Reactor Regulation Docket No. 50-220 and 50 410 t

Enclosure:

Supplemental Request for Additional

Information 4

, cc w/ encl: See next page

- .. .. - . . - - . - ~ - - . . - . - - . - - . . . - - . - . . - - . - - . _ - - _ - -

% John H. Mueller Nine Mile Point Nuclear Station Niagara Mohawk Power Corporation Unit Nos.1 and 2 I

cc: .

. Regional Administrator, Region I Charles Donaldson, Esquire U.S. Nuclear Regulatory Commission Assistant Attomey General

! 475 Allendale Road New York Department of Law

! King of Prussia, PA 19406 120 Broadway i New York, NY 10271 '

Resident inspector _

s U.S. Nuclear Regulatory Commission Mr. Paul D. Eddy P.O. Box 126 State of New York Department of Lycoming, NY 13093 Public Service Power Division, System Operations Mr. Jim Rettberg 3 Empire State Plaza  :

New York State Electric & Gas Albany, NY 12223 l Corporation Corporate Drive Mr. Timothy S. Carey  ;

Kirkwood Industrial Park Chair and Executive Director l P.O. Box 5224 State Consumer Protection Board  !

Binghamton, NY 13902-5224 5 Empire State Plaza, Suite 2101 l l Albany, NY 12223 '

i Supervisor Town of Scriba Mark J. Wetterhahn, Esquire Route 8, Box 382 Winston & Strawn Oswego,NY 13126 1400 L Street, NW Washington, DC 20005-3502 i Mr. Richard Goldsmith Syracuse University Gary D. Wilson, Esquire College of Law Niagara Mohawk Power Corporat;on E.1. White Hall Campus 300 Erie Boulevard West Syracuse, NY 12223 Syracuse, NY 13202 Mr. John V. Vinquist, MATS Inc. Mr. F. William Valentino, President P.O. Box 63 New York State Energy, Research, Lycoming, NY 13093 and Development Authority .

Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 1

1

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1 l

l i

2 .

4 i

SUPPLEMENTAL REQUEST FOR ADDITIONAL INFORMATION REGARDING RESPONSE TO GL 96-06 NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION. UNIT NOS.1 AND 2 DOCKET NOS. 50-220 AND 50-410 The NRC staff is reviewing responses by Niagara Mohawk Power Corporation (NMPC) to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions." In its February 7,1997, letter, NMPC identified 10

, Unit 1 pipe segments and 21 Unit 2 pipe segments that were susceptible to thermally-induced pressurization and were evaluated for operability. In its December 16,1997, letter, NMPC states that corrective actions for these pipe segments, including required modifications, will be completed before restarting from Units 1's refueling outage (RFO) 16 (spring 2001) and Unit 2's RFO 7 (spring 2000). The NRC staff requests the following additionalinformation regarding NMPC's evaluation of these pipe segments:

1. Provide summaries of NMPC's planned corrective actions for the 31 pipe segments determined to be susceptible to thermally-induced pressurization.

2. For any of NMPC's modifications that involve heat transfer and/or structural analyses of the pipe segments, provide the following information for these segments:

2.1 Specify the applicable design criteria for the piping and valves. This should include the required load combinations; 2.2 Submit a drawing (or identify a previously docketed drawing) of the piping run between the isolation valves. This should include the lengths and thicknesses of the piping segments and the type and thickness of the insulation; 2.3 Provide the maximum-calculated temperature and pressure for the piping run.

Describe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model and the basis for the heat transfer coefficients used in the analysis.

DRWT Encicsure

.* .' ** 2e l s  :* NUCLEAR REGULATORY COMMISSION

. wAswiwarow, o.c. sosse.coes l

% October 1, 1998 l

4 Mr. David J. Modeon Director, Engineering Nuclear Generation Division Nuclear Energy Institute 1776 l Street, NW, Suite 400 i

Washington, D.C. 20006 '

SUBJECT:

REVIEW OF EPRI TECHNICAL REPORT TR-108812, " RESPONSE OF ISOLATED PIPING TO THERMALLY INDUCED OVERPRESSURl2ATION -

DURING A LOSS OF COOLANT ACCIDENT (TAC NO. MA0695)

The Nuclear Energy Institute (NEI) submitted Electric Power Research Institute (EPRI) report TR-108812, " Response of Isolated Piping to Thermally induced Overpressurization During a i Loss of Coolant Accident," to the NRC on January 15,1998, for staff review. This report was developed to provide technical support for a proposed American Society of Mechanical Engineers (ASME) Code Case addressing the thermal overpressurization of isolated sections of piping issue in NRC Generic Letter 96-06, " Assurance of Equipment Operability and i Containment Integrity During Design-Basis Accident Conditions." The staff provided comments on the report to NEl in a letter dated February 23,1998. The staff met with representatives from NEl and EPRI to discuss these comments in a meeting on March 25,1998.

Subsequently, NEl transmitted EPRI's written response dated June 10,1998, to the staffs comments as an attachment to a June 15,1998 letter. ,

i The EPRI response has not resolved all of the staffs concems with TR-108812. The staff does not believe that the EPRI testing provides a sufficient technical basis to support the strain limits I proposed in the September 3,1998, draft of ASME Code Case N 584. The staffs specific comments regarding NEl's June 15,1998 letter are enclosed. If you have any questions j regarding this issue, please contact me at (301) 415-1355. i Sincerely, v N.

Beth A. Wetzel, Sen r Project Manager Project Directorate lll-1 Division of Reactor Projects-Ill/IV -

Office of Nuclear Reactor Regulation

Enclosure:

As stated cc wl encl: Mr. Kurt Cozens Project Manager Nuclear Generation Division Nuclear Energy Institute 1776 l Street, NW, Suite 400

- Washington, D.C. 20006

-__..s c . - - - - - - - - - - . .

'O eo l Staff Assessment of EPRI June 10,1998, Response to NRC Comments l Regarding EPRI Report TR-108812 1.a. The staff indicated that the EPRI testing did not address the impact of other design loads on the predicted strains. These other design loads may be sustained loads due to deadweight or suppressed thermal expansion of the pipe run, or they may be dynamic loads due to seismic events.

In response to the staff comment, EPRI performed an assessment of the impact of a sustained axial stress on calculated hoop strain in a pressurized pipe segment. The results are shown in Figures 1 and 2 of the EPRI response. On the basis of this assessment, EPRI concluded that the contribution of sustained loads on predicted hoop strain is negligible. The staff does not agree with this conclusion. Review of Figure 1 indicates that at a hoop strain of 4%, the addition of a 10 ksi axialload would increase the predicted hoop strain to 5.5%. This is a significant increase. The Code allowable stress limit for sustained loads and the Code-allowable stress limit for thermal expansion loads are both greater than 10 ksi. Consequently, the staff believes that bending stresses on the order 10 ksi or more due to the combination of deadweight and thermal expansion loads are possible.

With regard to dynamic loads such as seismic events, EPRI argued that the concurrent combination of seismic and loss-of-coolant accident (LOCA) loads is highly unlikely. 1 However, the staff concem was with the case where LOCA and seismic loads would l have to be combined to meet a licensing basis commitment. The EPRI response did not address this issue.

1 1.b,c. The staff indicated that the impact of local attachments on predicted strains has not been assessed. Mcny of the piping runs of concern contain test connections. .The staff also questioned the applicability of the test results to pipe runs containing fittings such as elbows and tees.

In its response, EPRI stated that, although the local attachment will incrasse the localized peak strains due to the stress concentrations caused by the discontinuity, the local discontinuity does not affect the general plastic membrane strain. Hawever, no test data was presented in support of the argument. The staff is still concemed that attachments and other fittings could experience local membrane strains (not necessarily peak strains) which are greater than the average hoop membrane strain in the pipe.

The staff believes that further testing and/or analysis is needed to address this concern.

1.d. The staff indicated that the testing did not uitress the impact of potential flaws in the piping on the predicted hoop strain at fracture. l l In response to the staff comment, EPRI performed an evaluation of impact of potential flaws on the calculated hoop membrane strain at failure for carbon steel pipe. The results of the evaluation indicate that only relatively deep circumferential cracks impact the calculated h'oop membrane failure strain. However, the EPRI evaluation indicates ENCLOSURE

4 h.,'

,4 2

that long axial cracks can have a significant impact on the calculated hoop membrane failure strain. Figure 8 of the EPRI response shows the hoop membrane strain at the calculated failure pressure for long axial cracks in the carbon steel pipe. Figure 8 indicates that the membrane hoop strain at failure would be less than 4% for long axial cracks greater than 15% through the thickness. The EPRI evaluation indicates that

there is little tolerance for long axial cracks in carbon steel piping at a hoop membrane l strain of 4%. The staff believes that the strain acceptance criteria must have adequate margin to accommodate potential material variations, additional loads not accounted for in the evaluation, and potential flaws that may exist in the piping system.

2. The staff indicated that the actual test data for the tensile specimens should have been provided in the EPRI report. In addition, the staff indicated that variability of the ultimate stress and ultimate strain values for the carbon and stainless steel materials listed in Section 2.2 of the EPRI report should have been discussed.

l l In response to the staff comment, EPRI provided the data from the tensile tests. In l addition, EPRI provided test data from the literature on stainless and carbon steel

! materials. EPRI concluded that the material properties in its test program are consistent with the comparable material properties published in the literature. The staff was concerned that a very limited amount of test data was used to assess margins associated the strain limits in the proposed ASME code case. As discussed in item 1.d, the staff believes that the potential variation of material properties should be considered in establishing the acceptance criteria. For example, the ASME Code Appendix F i inelastic criteria allow the use of material strain curves developed from test data.

l However, the criteria also require the results be adjusted to account for code minimum l

properties. This accounts for potential material property variations of installed components. The staff believes that material property variations should be considered l in establishing strain limits.

3. The staff indicated that additional detailed measurements of the pipe specimens that

were burst tested should have been provided in the report. In addition, the staff l

indicated that the method used to calculate the hoop strain values at burst should have l been provided.

EPRI responded that the initial dimensions were nominal and that post burst diametrical measurements were deemed to be of little value. The staff was interested in the detailed pipe diameter measurements to determine whether the hoop strain in the specirnen remained uniform during the burst test. This is an important consideration if the strain in the pipe is compared to the volumetric expansion of the fluid. A nonuniform strain distribution along the length of the pipe would result in a greater maximum hoop strain for a given volumetric expansion of the fluid. In discussions with the staff, EPRI indicated that the radial deformation of the pipe was uniform in the areas away from the structural discontinuities in the test specimens. The staff believes that detailed diameter i

measurements along the pipe would have provided usefulinformation in assessing this l issue.

l l

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! 3 EPRI also indicated that the hoop strain values at burst were based on the displacement j measurements obtained up to the point of rupture. This provided the information that j the staff requested.

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4. The staff questioned the assertion that the loading addressed by GL 96-06 is an " energy j controlled condition." The term " energy controlled condition" was used to justify an

acceptance criteria proposed in EPRI Technical Report NP-1921, ' Rationale for a j

Standard on the Requalification of Nuclear Class 1 Pressure-Boundary Components." l The acceptance criteria are based on the concept that an allowable limit of 70% of the failure strain for an " energy controlled condition" is equivalent to the ASME Code 4

Appendix F criteria of 70% of the ultimate stress for a load controlled condition.

i l

The EPRI response contains a discussion of the failure strain due to a load controlled j test verses the failure strain due to a deformation controlled test. This discussion and i the discussion in item 6 indicate that a larger failure strain would have been measured in a deformation controlled test. The staff does not agree with the inference that the pipe specimens would have failed at a higher hoop strain value if tested with a GL 96-06 loading. The staff believes that test data is needed to substantiate the EPRI claim.

I

5. The staff indicated that the ASME Section ll1 Special Working Group on Faulted Conditions considered the criteria proposed in EPRI Technical Report NP-1921 approximately 10 years ago. The NRC staff representative voted negative on the proposal. These criteria were never adopted by Section til for incorporation into Appendix F of the code. It appears that the technicalissues regarding the' criteria were never resolved by the special working group.

EPRI responded that it was not aware of any staff concems with EPRI Report NP-1921.

It is the staff's understanding that the Appendix F working group discussed the criteria in NP-1921 for use with dynamic impact loads. Whereas the use of energy considerations for evaluating impact loads is a common engineering design practice, extrapolation of criteria based on this concept to hoop membrane stress caused by intemal pressure is highly questionable. The staff's technical concem regarding NP-1921 involved equation 5-24. This equation is based on a theoretical failure criteria developed by F. A.

McClintoc'K. The theory had not been compared to test data to verify its applicability to materials used in nuclear power plants. In addition, the staff did not agree with the concept of using 70% of the failure strain as an acceptance criterion.

6. The staff indicated that the measured hoop burst strain for the carbon steel specimen reported in Table 2-5 of EPRI Report TR-108812 is less than 9%. The staff also indicated that the strain computed using the underlying theory presented in NP-1921 does not appear to correlate very well with test results for hoop strain.

The EPRI response discusses the concept of load-controlled verses energy. controlled loading. EPRI concludes that the measured burst strain is consistent with the theory for a load controlled failure. However, the discussion infers that a greater strain would have been measured for an " energy controlled loading." As discussed in item 4 above, the staff believes that the concept that the pipe would have failed at a greater strain if it were subjected to a GL 96-06-type loading condition needs to be confirmed by testing.

.