ML20207D314

Ack Receipt of 981203,990325 & 0415 Ltrs,Which Presented Several Concerns & Comments Re Operation of Nmp,Plant,Unit 1 During Last Operating Cycle

ML20207D314
Person / Time
Site:	Nine Mile Point
Issue date:	05/28/1999
From:	Adensam E NRC (Affiliation Not Assigned)
To:	Penn S SYRACUSE UNIV., SYRACUSE, NY
References
NUDOCS 9906030304
Download: ML20207D314 (18)

Text

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[ t UNITED STATES j j NUCLEAR REGULATORY COMMISSION C WASHINGTON, D.C. 20555-0001 o

.....g M3y 28,1999 Dr. Steven Penn Syracuse University 201 Physics Building Syracuse, NY 13244

Dear Dr. Penn:

This is to acknowledge receipt of your letters dated December 3,1998, March 25,1999, and April 15,1999, to the U.S. Nuclear Regulatory Commission (NRC), in which you presented several concerns and comments regarding the operation of Niagara Mohawk Power

, Corporation's (NMPC or the licensee) Nine Mile Point Nuclear Station, Unit 1 (NMP1) during its last operating cycle. This letter further confirms our request that correspondence related to the NMPC's PS ;; Mile Point, Units 1 and 2, should be addressed to the attention of Mr. Darl Hood, the NRC Project Manager for licensing activities associated with these units. >

We apologize for not keeping you better informed as to the progress of this review, and would j like to assure you that we take very seriously all concerns and comments provided to the NRC by members of the public. Our mission and highest priority are to maintain safety. We recognize that, to be a successful regulator and provide effective oversight, the NRC must be responsive to the public, while considering the burden of regulation on the business community, and the effectiveness and efficiency of our regulatory processes. Toward this end, this letter is intended to address your various comments and concerns.

In your March 25,1999, letter, you inquired as to the status of the NRC staff's response to your December 3,1998, letter regarding the NMP1 core shroud. We have completed our review of the comments and concerns in your December 3,1998, letter; and our response is provided in Enclosure 1.

We have determined that your March 25,1999, letter presented several concerns related to both the licensee's September 30,1997, submittal, as supplemented by letters dated January 30, February 27, September 21, and October 22,1998, and the related NRC staff's Safety Evaluation (SE) dated November 2,1998. The NRC staff's SE considered both the Irkrmation provided by the licensee and comments received from members of the public during p alic meetings on September 24,1998. Since many of the comments in your letters were n ommendations for the licensee rather than regulatory requirements, we forwarded your is ers to the licensee, who responded on April 30,1999. Our review indicates that the licensee has adequately addressed the technical concerns in your March 25,1999, letter. We have also requested our contractor from Argonne National Laboratories to review the licensee's responses of April 30,1999, and that review is continuing. If our contractor's review should identify any item of concern or require further explanation, we would advise you accordingly. In Enclosure 2, we have addressed the item in your March 25,1999, letter that is not of a technical nature.

In your March 25,1999, letter, you also requested several actions by the NRC staff.

Specifically, you requested copies, as available, of the (1) NMPC NMP1 core shroud inspection plan, (2) NMPC crack inspection results, (3) NMPC's report to the NRC justifying restart, and .

(4) the NRC's SE approving restart. Regarding item (1), you acknowledged in your April 15, j 9906030304 990528 PDR ADOCK 05000220 {

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Dr. St:v:n Penn May 28,1999 1999,1:tt:r thit you "...hiva revi:w d tho inspection plan (Dock *,t No. 50-220, DPR-63) submitt d by Ni:girs Mohawk Pow:r Corporation (NMPC) on 30 Dac:mb:r 1998 rzgarding the current inspection of the core shroud at Nine Mile Point One (NM1). I have also read the NRC letter (TAC No. MA4491) dated 24 March 1999 authorizing that inspection plan." Regarding )

item (2), the licensee's reinspection results and analysis to disposition these reinspection ,

findings are required to be submitted to the NRC within 30 days of completing the reinspection.  !

The NRC will expedite placing into the NRC's Public Document Room (PDR), and the local PDR (Reference and Documents Department, Penfield Library, State University of New York, l Oswego, NY 13126), all non-proprietary submittals regarding the reinspection results from the licensee and any related correspondence from the NRC staff to the licensee. The NRC does not normally provide copies of docketed correspondence, since these documents are available at the PDRs. Regarding items (3) and (4), decisions regarding restart are the responsibilities of individual licensees and do not normally require the submittal of advanced justification to the NRC. The licensee is not required to obtain NRC approval to restart NMP1, and we do not anticipate issuing any SE approving restart of NMP1.

In your April 15,1999, letter, you discussed NMP1's inspection plan for the core shroud and stated that you "...would like to enumerate the shortcomings of that inspection plan and to present an addition to the inspection plan which would greatly increase our understanding of the extent of cracking in the shroud." Your comment regarding the core shroud plates was addrecced by the licensee in its response of April 30,1999. Our response to your remaining comments in this correspondence, including our disagreement with your conclusions regarding the adequacy of the inspection plan, is included as Enclosure 3.

In addition to responding to your letters on April 30,1999, the licensee, in a letter dated April 20,1999, committed to follow approved NRC staff criteria regarding disposition of the inspection results for the upcoming operating cycle. The licensee has defined its consideration '

regarding repair of the vertical welds in that letter, and has determined that a pre-emptive repair will be installed to reinforce the core shroud V9 and V10 welds.

The NRC thanks you for your expressed concern and your considered opinions regarding the NMP1 core shroud reinspection. If you have any further questions or comments, please contact Mr. Hood at 301-415-3049 or dsh@nrc. gov (e-mail).

Sincerely, ORIGINAL SIGNED BY:

Elinor G. Adensam, Director Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220 DISTRIBUTION: I

Enclosures:

As stated cc w/encis: See next page

% Docket File; PUBLIC S. Bajwa S. Little

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• lin coDV 0FFICE EMCB:DE* lE EMCD:DE* lE EMCB:DE* lE EMCB:DE* lE NAME CECarpenter WHKoo RAHermann WEBateman DATE 05/10/1999 05/11/1999 05/11/1999 05/12/1999

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n Dr. Steven Penn 1999, letter that you "...have reviewed the inspection plan (Docket No. 50-220, DPR-63) submitted by Niagara Mohawk Power Corporation (NMPC) on 30 December 1998 regarding the current inspection of the core shroud at Nine Mile Point One (NM1). I have also read the NRC letter (TAC No. MA4491) dated 24 March 1999 authorizing that inspection plan." Regarding item (2), the licensee's reinspection results and analysis to disposition these reinspection findings are required to be submitted to the NRC within 30 days of completing the reinspection.

The NRC will expedite placing into the NRC's Public Document Room (PDR), and the local PDR (Reference and Documents Department, Penfield Library, State University of New York, Oswego, NY 13126), (!i non-proprietary submittals regarding the reinspection results from the licensee and any related correspondence from the NRC staff to the licensee. The NRC does not normally provide copies of docketed correspondence, since these documents are available at the PDRs. Regarding items (3) and (4), decisions regarding restart are the responsibilities of individual licensees and do not normally require the submittal of advanced justification to the NRC. The licensee is not required to obtain NRC approval to restart NMP1, and we do not anticipate issuing any SE approving restart of NMP1.

In your April 15,1999, letter, you discussed NMP1's inspection plan for the core shroud and stated that you "...would like to enumerate the shortcomings of that inspection plan and to present an addition to the inspection plan which would greatly increase our understanding of the extent of cracking in the shroud." Your comment regarding the core shroud plates was addressed by the licensee in its response of April 30,1999. Our response to your remaining comments in this correspondence, including our disagreement with your conclusions regarding the adequacy of the inspection plan, is included es Enclosure 3.

In addition to responding to your letters on April 30,1999, the licensee, in a letter dated April 20,1999, committed to follow approved NRC staff criteria regarding disposition of the inspection results for the upcoming operating cycle. The licensee has defined its consideration regarding repair of the vertical welds in that letter, and has determined that a pre-emptive repair will be installed to reinforce the core shroud V9 and V10 welds.

The NRC thanks you for your expressed concern and your considered opinions regarding the NMP1 core shroud reinspection. If you have any further questions or comments, please contact Mr. Hood at 301-415-3049 or dsh @nrc. gov (e-mail).

Sincerely, 4 ,

Elinor G. Adensam, Director Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

As stated cc w/encls: See next page

Nine Mile Point Nuclear Station Unit No.1 Regional Administrator, Region i U.S. Nuclear Regulatory Commission j 475 Allendale Road King of Prussia, PA 19406 Resident inspector U.S. Nuclear Regulatory Commission

- P.O. Box 126 '

Lycoming, NY 13093 a

Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy State of New York Department of Public Service Power Division, System Operations 3 Empire State Plaza -

Albany, NY 12223 Mr. F. William Valentino, President New York State Energy, Research, and Development Authority Corporate Plaza West '

286 Washington Avenue Extension

' Albany, NY 12203-6399 Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Gary D. Wilson, Esquire Niagara Mohawk Power Corporation  !

300 Erie Boulevard West Syracuse, NY 13202  !

l Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 l

1 i

r U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION l EVALUATION OF COMMENTS AND CONCERNS PRESENTED BY A MEMBER OF THE PUBLIC BY LETTER DATED DECEMBER 3.1998 In your letter of December 3,1998, you provided to the U.S. Nuclear Regulatory Commission (NRC) staff several concerns related to the proposal by the Niagara Mohawk Power Corporation (NMPC or the licensee) to extend the Nine Mile Point Nuclear Station, Unit 1 (NMP1) operating period from 10,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> to 14,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, and the review of that proposal as presented in the NRC staff's safety evaluation (SE) dated November 2,1998.

The NRC staff has reviewed your concerns, and has found, in general, that the issues you raised are not safety significant and do not warrant further action on th,e part of the NRC staff.

The reasons for this conclusion are enumerated below in the individual responses to your specific comments and concerns.

You indicated in your December 3,1998, letter that the conclusions of the licensee's technical reports could be influenced or biased under political and economic pressures, particularly, when there is a conflict of interest. You stated that ... Niagara Mohawk has based their arguments for an inspection delay on the conclusions of a studyperformed by General Electric. General Electric built the Nine Mile Point One reactor and they are a manufacturer of reactor fuel rod assemblies. Their clear financialinterest demonstrates an obvious conflict ofinterest. In addition General Electric's record for technicalintegrity is less than stellar.... While I do not claim that the report issued by GE includes any biased results, given GE's conflict of interest in this case, I believe that their report should only be utilized by the NRC if a parallel study of equalscope was performed as well. h The technical reports submitted by the licenseo to support its request to use alternative repairs ,

to those contained in Section XI of the American Society of Mechanical Engineers (ASME)  !

Boiler and Pressure Vessel (B&PV) Code have been thoroughly reviewed by the NRC staff, l with assistance from Argonne National Laboratory (ANL). The review does not support a claim that the GE report includes biased results or should only be used if a parallel study exists.

Since you have raised issues in this letter regarding the competency and integrity of the licensee and its contractors, as well as the NRC staff in reviewing this matter, we have forwarded your letters to the appropriate NRC Offices to further evaluate your concerns.

In your letter you stated that the shroud cracking is not a well understood problem and suggested that the NRC should perform a full analysis to determine the associated risk factor as a function of the extent of shroud cracking. You also identified the difficulties in performing ,

safety analyses and recommended the development of a detailed model to facilitate the performance of such an evaluation. The NRC staff agrees with you that to perform such analyses as prescribed in your letter would not be an easy task. The NRC staff's SE was based upon a deterministic review. The core shroud circumferential cracking was repaired using the tie rod assembly to " replace the circumferential welds"in the core shroud. The design assumptions were that the circumferential welds are completely cracked; they are not. Vertical weld integrity is needed to satisfy the design assumptions of the repair; ring segments are held l in place laterally and axially by the tie rod assembly. The structuralintegrity of the core shroud l is maintained with a high degree of assurance since the ASME B&PV criteria have I

Enclosure 1

been satisfied for the design of the repair and for the evaluation of the flawed vertical welds for the period specified prior to the next inspection. For example, the conservatism in the ASME criteria when using limit load analysis for evaluating the vertical welds is that a safety margin of 2.77 or 1,39 be maintained on the design loads for the most limiting conditions for normal / upset or emergency / faulted conditions at the end of the operating cycle. These safety margins are

. consistent with those for design of the original uncracked shroud. Further, the bypass leakage associated with the cracking has been evaluated and found to be negligible. With regard to cooling of the core under postulated, remote likelihood design basis accidents, this unit does not rely upon core reflood capabilities but is, instead, spray cooled by the core spray system.

Therefore, any significant increase in risk with regard to core damage frequency is considered unlikely.

o You have expressed a concern that very little is known about the actual crack growth rate in the NMP1 core shroud because of the lack of fundamental data obtained from performing the measurements of crack sizes at various times. You also indicated that these fundamental data are essential to verifying models designed to predict the crack growth. It is important to understand that the NRC staff's SE is based upon the use of a bounding crack growth rate for a simple geometry such as the vertical shroud welds. The NRC staff's determination that the crack growth rate is bounding did not rely upon a theoretical model but upon an empirical assessment of laboratory and field data. From a safety perspective, it is not necessary to know

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the actual crack growth rate as long as the crack growth rate used in the flaw evaluation is bounding. The NRC staff has high confidence in the validity of the crack growth rate used in .

the flaw evaluation, and this has not been reduced by the preliminary results of the core shroud inspection. Furthermore, extensive laboratory testing to evaluate the crack growth in various simulated reactor environments were performed in this country, as well as in Europe and l' Japan, for unirradiated stair.less steels. Much information is available regarding the effect of the controlling parameters srch as material sensitization, water chemistry, ar,d stress intensity on the intergranular stress corrosion cracking (IGSCC) in the BWR environment. The cracking in the NMP1 core shroud vertical welds is not unique. The calculated fluences at the locations of interest, as well as the results of the metallography, have amply demonstrated that the cracking in the NMP1 core shroud is typical of IGSCC rather than irradiation assisted stress corrosion cracking, j The following are the specific concerns you provided in your December 3,1998, letter, I regarding the NRC staff's November 2,1998, safety evaluation and the NRC staff's responses.

Concem: (Comment on page 5, final paragraph, regarding the method used in calculating the measurement errors associated with the measured crack length) The report discusses how the systematic error is calculated for the measurement of the crack length. The crack length is measured by determining the position of each end of the crack using a detector. The detectoris positioned using a delivery system.

Let use declare, a, the position of the detector according to the delivery system, and, b, the position of the crack end as measured by the detector, in this case the crack length is given by I= (a, + b,)-(a2 + be) where the subscripts refer to the measurements for a given end of the crack. In this case the resulting errorin the length is given by a, = (2 * (a," + co '))"' where a, and a oare the errors in a and b, respectively. Using the errors cited in the report, a, = 1.106 and a, =0.364 inches, J

I~

3-yields a length error ai = 1.647 inches which is about half the figure given in the report. The author assumed that ai = 2(a, + c o) which is incorrect.

Response: The NRC staff is aware of the method you recommended for treating the flaw length measurement errors resulting from measurement uncertainties. However, it is more conservative to add the measurement uncertainties directly to the measured length because the flaw length estimated by this summation method is longer than that using the method you have recommended. The purpose of using the summation method is to estimate an upper bound of the flaw length.

Therefore, to assess the structural integrity of the core shroud, it is conservative to compare the upper bound flaw length with the Code allowable values.

Concern: (Comment on page 6, first paragraph regarding the use of an estimated flaw length for further calculations which may not yield the most conservative value such as used to calculate the crack growth rate) The EVT(enhanced visual test) erroris given at 1.2 inches. I willassume that this error was calculated correctly although I am left wondering ifit does not suffer from the mistake noted on page S.

Nonetheless, while the erroris first listed as being the errorin the length, in the next sentence the author then calculates the total measurement uncertainty as 2.4 inches. I willassume that the author means for the first error to be errorin determining the end of the crack. In this case the crack length erroris ai= 1.7 inches. In any case a; # 2.4 inches. The author then makes the egregious mistake of adding the error to the value of the measured cracklength. Values and errors are two entirely separate types of objects that obey completely different mathematics. For those unfamiliar with the basics of error analysis let me recommend that you read Data Reduction and Error Analysis for the Physical Sciences by P. R. Bevington. This book teaches the well founded andproven method forpropagating errors through a calculation. When one simply adds the error to the value, as was done in your report, then that value is wrong and all further calculations done using this length value willyield the most consewative value (i.e. consider the crack growth rate in this case, or any case where errors are not constant for alldata).

Response: The intent of the NMPC's flaw length estimates is to establish an upper bound of the flaw length by adding the potential measurement uncertainties to the measured length. The estimated flaw lengths are not used to determine the crack growth rate for flaw evaluation. They are being directly used to assess the structural integrity of the core shroud by comparing with the Code allowed values. A consewatively estimated crack length will provide additional margin in the structuralintegrity of the core shroud.

Concern: (Comment on page 6, section 3.3, paragraph 4, regarding how the stress from the tie rods will affect crack growth rates in the vertical welds) "The authors note that NMPC was unable to detect welds V5 and V6 using their current detection methods. They also point out that the tie rod repairloses its effectiveness if these

, welds are completely cracked and welds H2 and H3 are completely cracked. I find it surprising that his point is passed overlightly. One of the two majorpoints that the NRC report addresses is the failed tie rod repair. If these welds are

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. i e l significantly cracked then the tie rods lose thermalpreload. In addition since the l ring segment is mounted normal to the direction of tie rod stress, these welds will l be stressed due to the torque from the tie rods and will thus be subjected to higher crack growth rates. NMPC should model how the stress from the tie rods will effect crack growth rates in the vertical welds. The authors conclude this discussion by stating the inspections reveal that the cracks in weld H3 were not extensive enough to inhibit safe operation. However, the report on the weld inspection, given in Appendix C, showed that weld H3 was not inspected (or the results were not listed in the report).

Response: NMPC has calculated the stresses imposed upon the vertical welds resulting from the tie rods during normal operation. The magnitude of the stresses that would affect the crack growth in the vertical welds is determined to be small and is insignificant when compared with the total of the combined stresses considered in the bounding crack growth rate used in the crack growth calculations.

In this paragraph, the NRC staff stated that the inspections verified that sufficient intact weld length exists in the H3 weld to support operation, even if the ring segment welds were significantly cracked. There is an typographical error in this statement. The referenced H3 weld should be the H2 weld because the H3 weld was not re-inspected during the 1997 refueling outage. The H3 weld was inspected visually during a previous refueling outage (1995) and the reported cracking of this weld was not extensive. The typographical error has no impact upon the conclusion.

Concern: (Comment on page 8, paragraph 2, regarding the large error associated with the crack growth rate determined from the flaw length measurements) The authoris discussing the crack growth data from the Brunswick BWR and comparing the crack size measured using UT during two successive refueling outages. The author notes "that there was no change in length or depth" which is a meaningless statement in the context of a measurement whete the errorin the detector reading is twice as large as the expected signal. Let me explain. If we assume a typical fuel cycle of 14,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and a crack growth rate of 5 x 10' inches / hour, then we might expect a crack length increase of about 0.7 inches. However, as noted above the length erroris cri = 1.647 inches for the method used by NMPC. I will assume that the method used for the Brunswick measurement was of similar accuracy to the method used by NMPC (if not then questions arise as to why NMPC would use a less accurate method). Sinco the erroris large compared to the expected signal one MUST cite the error to understand the measurement.

Even if the exact same length and depth were measured the crack may still have grown. In fact by assuming the values for the length error and the fuel cycle duration the error in crack growth rate would be 16.64 x 10' inches / hour or 3.3 times the NRC limit of S x 10' inches / hour.

Response: Because of large measurement uncertainties associated with the flaw length measurements (2.94 inches for ultrasonic testing (UT) examination and 2.4 inches for enhanced visual examination), it is not meaningful to assess the crack growth rate based upon such measurements if the anticipated crack growth is within the measurement uncertainties. Howaver, the uncertainties associated with the UT

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1 examinations of the flaw depth are much smaller. As stated in the NRC staff's SE,.

pages 5 and 6, section 3.2, paragraph 2, the uncertainty factor associated with the  ;

depth measurement was determined to be 0.108 inch, and was applied to each UT l

depth measurement. The uncertainty was determined by root mean square (RMS) l method during UT technique demonstration. Therefore, based upon UT

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measurements of flaw depth after each approved operating period, we are able to i determine with reasonable confidence whether the actual crack growth is bounded )

by the crack growth rate used in the flaw evaluation. 1 Concern: (Comment on page 8, paragraph 2, regarding the lack of field data on crack growth for vertical welds at NMP1) 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. This fact alone should be the cause of great concern to the NRC l for without data from the system in question we can not truly say we understand i the system. Gathering more data on NMP1 core shroud crack size should be of I paramount importance to both the NRC and NMPC. i l

Response: It appears that this comment pertains to page 10, paragraph 2, not page 8, paragraph 2. The cracking of vertical and horizontal welds in the core shroud is '

not different since the materials, water chemistr'y, stresses and fabrication methods are similar. Therefore, we do not expect the cracks in the vertical welds ,

to behave significantly differently from the horizontal welds. The cracking in the vertical welds was reported to be typical of IGSCC. We have observed IGSCC in the core shroud and recirculation piping of many other BWRs. We expect the IGSCC in the vertical welds to behave similarly to those found in other places.

However, the NRC staff does agree with you that gathering additional data on NMP1 shroud cracking is needed to verify that the rates used in the analyses were i bounding. NMPC was required to inspect the cracked core shroud welds after  !

completion of the approved operating period, and is presently analyzing the inspection results.

Concern: (Comment on page 8, paragraph 2, regarding the crack growth rate determined by measurements performed on weld H8) The single measurement on crack growth t rate was performed on weld H8. The depth of the crack measured during the 14* I fueling outage was less than that measured during the 13* fueling outage. NMPC l drew no conclusions about the data and the NRC simply went along with the '

industry's erroneous thinking. Once again the NRC neglected tho importance of the relevant error analysis in this case. When the measurement errors were considered was the data within the error? If not then how different was the measurementin terms of the error (e.g.107 or2a?). If there was a large deviation then the NRC should be concerned that the detector error being reported by the NMPC is too small. Converselyif the data was within the error then the data should not be disregarded for it is useful data which implies a range of possible growth rates. One is always left to wonder when data is disregarded whether such exclusion is done justly or whether it is excluded because it does not conform to the experimenter's preformed conclusions. The NRC need to be especially watchful of such tainted science when dealing with scientists who may be under pressure to achieve a conclusion which benefits their employer.

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Response: It appears that this comment pertains to page 10, paragraph 2, not page 8, paragraph 2. It would not be meaningful to discuss the crack growth rate for the flaw observed in the circumferential weld H8 since the UT measurements showed a net decrease in depth after operation of one fuel cycle. NMPC attributed the decrease in flaw depth to the very conservative method used in the earlier sizing determination. We cannot make a reliable estimate of the crack growth rate when the initial flaw size is not accurately known. However, we could make a conservative estimate of the limiting crack growth rate by assuming the entire crack was grown during a given fuel cycle. For a crack growth rate of 5 x 104 inch / hour, the crack growth in one fuel cycle (assuming 14,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />) would be 0.725 inch. Since the reported flaw depths in the 1995 and 1997 outages were about 15% (0.23 inch) and 10 % (0.15 inch) of the wall thickness, we are able to e

conclude with reasonab!e confidence that the crack growth in the through-thickness direction, assuming all growth occurred during a given fuel cycle, is

, bounded by the crack growth rate used in the flaw evaluations, if the UT measurement uncertainty of 0.108 inch is added to the measured flaw depth, the conclusion still remains valid.

Concern: (Comments on page 11, last paragraph, regarding the meaning of the statement that the allowable crack sizes are large and approach or exceed the length of the weld itself) The allowable vertical weld flaw size is being modeled assuming that the horizontal welds are cracked thru wall. The last sentence reads, " Typically, the allowable crack sizes are large and approach or exceed the length of the weld itself." What exactly does it mean when the allowable crack size is larger than the -

welditself? If the model truly allows for such unphysicalresults then the relevance ,

of the results of that modelshould be seriously questioned.

Response: Your statement of "unphysical results" referred to the allowable crack sizes in the core shroud. The NMP1 core shroud may be viewed as consisting of five cylinders stacked on top of each other, joined by the circumferential welds. Each cylinder contains two vertical welds. With this core shroud geometry, the length of each of the vertical welds V9/V10 (approximately 90 inches) is limited by the length of the cylinder. If the cracking were to continue through the other cylinders, the crack could be as long as 238 inches. However, this is considered unlikely because the welds are not aligned and, therefore, crack extension into the adjacent plate base metal is considered unlikely. Since the stainless steel material is much tolerant to flaws, the allowable flaw length in the core shroud can be very long. Therefore, it is not an "unphysical result" when the allowable flaw length in the core shroud approaches, or exceeds, the length of a vertical weld in a single cylinder when the length of the joined cylinders is much longer.

Concem: (Comment on page 13, section 4.2.4, regarding NRC's not performing the necessary safety calculations) This section discusses the NRC staff's independent calculation for the largest acceptable axial flaw. According to the authors the NRC performed a calculation for the bounding final crack length and found it to be slightlylarger than would be allowed assuming both LEFM failure mechanism and the ASME criteria. Since the authors do not cite any numerical results the reader has no way of assessing the validity of their conclusions. Nevertheless the NRC

notes that NMPC also obtained a similar result when they performed their calculation assuming that the cracks were through walland that the horizontal welds were fully cracked. The authors then noted that a more detailed NMPC calculation, which included the additional strength from partially cracked welds, was within the ASME safety standards. They thenjump to the conclusion thatif they performed a similarly more detailed calculation that their results would also fall within the safety requirements. The authors may be right, but it is the responsibility of the NRC to do these calculations and not speculate what the results might be. It is the NRC'sJob to perform the necessary safety calculations to ensure that the public's safety is being protected.

Response: The NRC staff has carefully reviewed NMPC's flaw evaluations and its crack growth calculations. The NRC staff provided a discussion of LEFM and limit load analysis methodologies in paragraph 4.2.2. The results of the licensee's LEFM analyses performed using a handbook solution and a finite element solution as well as a limit load analysis are discussed in paragraph 4.2.3. The licensee found that all the vertical welds except V4, V9 and V10 were acceptable based upon LEFM analysis using a handbook solution or limit load analysis based upon a  :

conservative assumption that all cracks were through wall. The three remaining cracks were analyzed considering the depth of the cracking. The LEFM / finite element analysis was used by the licensee for vertical welds V9 and V10 because this was a limiting method considering the fluence. The NRC staff reviewed the finite element model, includi.ig the special crack tip elements. As stated in paragraph 4.2.4, the NRC staff performed its own LEFM handbook analysis and limit load analyses and confirmed that the LEFM analysis is bounding, in performing the independent LEFM handbook analysis, the NRC staff confirrned that the calculated stress intensity,175 ksi(in, was greater than the fracture toughness value of 150 ksi(in, as determined from the test results from a foreign plant. Based upon the results of our analyses and our review of the licensee's LEFM / finite element analysis, we have determined that it would be of no benefit to independently re-perform the computer calculation performed by the licensee when we have reviewed the model and the input assumptions as well as the reasonableness of the results.

Concern: (Comment on page 14, section 4.3.1, regarding the discrepancy in the reported crack area) There is a math errorin calculating the crack area. It states that a crack has an area of 3 square inches for a 0.003 inch wide and 90 inches long crack. Assuming the crack to be 0.003 inches wide, the correct area is 0.27 square inches. The leakage rate should be reduced by a similar factor although this will not affect [ sic] the conclusions drawn in this section of the report.

Response: The NRC staff's SE dated May 8,1997, stated the following:

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 ofleakage 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 calculatedleakage from the V-9 and V-10 welds was estimated to be 200

r t

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1 8-gpm at 100-percent ratedpower and 100-percent rated flow. This statistic is \

equivalent to approximately 0.11 percent of the core mass flow.

The NRC ststf has re-evaluated the calculation for the crack area of the vertical A welds and has confirmed that the estimated crack area of V9 and V10 is 3 square t inches. Unfortunately, neither the staff nor the licensee can determine where the 0.003-inch opening value originated since it does not appear in any docketed information from the licensee and this number appears to be in error.

Furthermore, the crack along the vertical weld may not have the same opening because, as discussed below, the crack opening depends upon the pressure drop across the shroud wall. The crack area and associated leakage estimate were 3

provided in the proprietary report GE NE-B13-01869-043, Revision .1, Q

" Assessment of the Vertical Weld Cracking on the NMP1 Shroud," dated April 1997. The General Electric report stated that the postulated crack openings for j vertical welds V9 and V10 were determined from structural analysis and were '

dependent upon the pressure drop across the shroud wall. The NRC staff requested further information regarding the calculation of the crack area and the  !

estimated leakage flow rates from the licensee by letter dated April 18,1997. The  !

licensee responded by letter dated April 23,1997.

In its response, the licensee stated that the leakage estimate was performed i assuming that the entire length of vertical welds V9 and V10 were cracked through-wall. Additionally, the combined length of vertical welds V9 and V10 was used in the analysis, i.e., the length of the crack was assumed to be 180 inches versus 90 inches. The crack opening area was estimated using fracture mechanics solutions for a through-wall crack in a cylindrical shell. These solutions are discussed in NUREG/CR-3464,"The Application of Fracture Proof Design Methods Using Tearing Instability Theory to Nuclear Piping Postulating 3 Circumferential Through Wall Cracks." A copy of the NUREG is publicly available. I As shown in the NUREG, the fracture mechanics solutions are more detailed than the typical length-times-width calculation of area. Once the crack opening was j calculated, the leakage flow area was determined by multiplying the entire length '

of both vertical welds V9 and V10 by the crack opening and applying other applicable conservatisms. Based upon our review, the NRC staff believes that the analysis performed by NMPC and its contractor, General Electric, provides a conservative estimate of the postulated crack opening area and associated l leakage flow rates.

Concern: (Comment on page 17, section 5.2.1, regarding the loosening of the tie rods and failure of the clasps) In this section the authors review the condition of the tie rod assembly which had been in place for a single fuel cycle. A great deal ofinterest is placed on the tie rod on the 270 location which had vibrated loose and was thus not providing the full design tension. Indeed, the updated design was intended to remedy this defect. However, I was alarmed to leam that the clasp at 90* had snapped and that part of the clasp was thrown over to 330*. In any engineering design, one expects that the hardware willbe designed to withstand stresses at least double the maximum expected stress. I would assume that in a nuclear reactor, the safety margin should be even higher. Why then did that clasp snap?

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F l Was the material of poor quality? Was the calculated stress too low? How did this flaw get past the initial NRC review? These are serious questions which are not addressed and leaves one in doubt concerning in the design's revision.

Response: The root cause for the 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. Because the position of the lower support within the machined holes was not procedurally controlled during installation, the relative position of the bolts within the holes was randomly 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 due to tensile overload. These latches were not designed to accommodate differential movement between the reactor pressure vessel (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. The NRC staff has reviewed NMPC's analysis of the degradation and agrees with its 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 between 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. This aspect was overlooked in the original design. During thermal growth of the tie rod assembly, the lower support toggle bolts moved up in the oversized holes, this movement caused the retainer clip or latch to extend beyond its elastic limit and fail. The failure of the latch was not due to any deficiency in the material properties, but rather due to the movement of the toggle bolts within oversized support holes. The new latches have been redesigned with substantial improvements in design margins to accommodate unanticipated operating conditions. In addition, the procedure for installing the toggle bolts has now been revised to ensure that there is no movement due to thermal growth. Based upon additional analysis, the licensee determined that the remaining shroud repair assemblies would have performed I

their intended function.

Concern: (Comment on page 17, section 5.2.5, regarding the tie rod middle supports)

NMPC reported that the tie rod at 270* had loosened because the base of the rod was not correctly seated in the tie rod anchor on the apron of the core shroud. As part of the design revision they recommended a new installation procedure which, they said, wouldinsure that the rods were properly seated. However, two l

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paragraphs later, the NRC notes that upon inspection afterinstallation, NMPC found that the tie rod middle supports at 90* and 166* were no longer in contact with the RPV. They attributed this loss of contact to " movement of the lower support assembly up the cone toward the shroud." It is just such movement that the new installation procedure was supposed to eliminate. This description suggests that the new design does not correct for the loosening problem.

Response: This comment appears to be a result of misunderstanding the nature of the

" movement of the lower support assembly up the cone toward the shroud

• and its effect upon the gap at the middle supports. The tie rod middle supports were machined to dimensions based upon the actual gap between the tie rod and the RPV wall. This gap increased when the lower support toggle bolts were moved up in the oversized holes after the new installation procedure. Therefore, during the j inspection after the implementation of the new procedure, the middle supports at 90* and 166* were no longer in contact with the RPV wall, as might have been expected. New middle supports based upon actual gap size were installed. This change in the gap of the middle support due to lateral motion was not accidental or unexpected movement during installation. Rather, that lateral motion was part of the design modification.

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CONFLICT OF INTEREST AND INDEPENDENT ANALYSES CONCERNS PRESENTED IN MARCH 25.1999. LETTER l

Comment: As I discussed in my letter of 3 December 1998, General Electric (GE) and NMPC have a conflict ofinterest which prevents calculations performed by these l corporations and their paid consultants from being viewed as truly objective research. Therefore all critical measurements and calculations performed by these corporations which were used as a basis for a safety evaluation should be repeated by the NRC orits paid consultants. Moreover, ALL data and methodology should be available for public review. I find it particularly disturbing that some of the supporting information for this SE have been withheld from public exposure by General Electric. ... If GE and NMPC are unwilling to fully publish their data and their methodology then the NRC should sponsor independent and open research on the materialin question.

Response: The NRC does not agree that measurements and calculations performed by GE  !

and NMPC (or by other reactor vendors and nuclear utilities) should not be relied upon for a safety conclusion unless first repeated by the NRC or its consultants.

As provided in the Atomic Energy Act, as amended, licensed utilities are ultimately responsible for (and have a vested interest in) the safe operation of their nuclear facilities. The NRC's regulatory program is based upon licensees and their contractors, employees, and agents acting with integrity and ,

communicating with candor. The Commission's regulations have appropriate '

provisions and enforcement sanctions to ensure that information provided to the Commission is complete and accurate in all material respects.

The NRC staff also does not agree that all data and methodology, including l some of GE's supporting information, should either be available for public review or the NRC should sponsor independent and open research on that material. In promulgating its regulations regarding availability of official records (see e.g.,

10 CFR 2.790, "Public Inspections, Exemptions, Requests for Withholding"), the Commission provides a balance between the benefits of, public interest in disclosure and the benefits of withholding certain types of information. The information regarding the NMP1 core shroud that has not been made publicly available was submitted to the NRC pursuant to the requirements of 10 CFR 2.790 for withholding certain information.

l Enclosure 2 L

t SPECIFIC COMMENTS AND CONCERNS PRESENTED IN APRIL 15.1999. LETTER Comment: In this letter I would like to enumerate the shortcomings of that inspection plan and to present an addition to the inspection plan which would greatly increase our understanding of the extent of cracking in the shroud. ... The current inspection plan will measure the present size of the cracks along all the accessible vertical welds.

Assuming that the crack depth measurements are performed with the same level of accuracy as in the last measurement performed in 1997, then the data should be able to establish a crack growth rate with an accuracy much better than 10*

inches / hour. This accuracy should be sufficient to determine whether the core shroud cracking poses a serious safety risk during the next fuel cycle. I would lil<e to point out that even if the data indicates an acceptable crack growth rate, I strongly recommend that NMPC measure the crack size during subsequent RFO's to ensure that that safety levelis maintained.

Response:The licensee has inspected the core shroud vertical welds and will provide the results of these inspections on the docket. If the analysis of the crack growth shows that the crack would exceed the threshold previously established, the licensee has stated that repairs would be performed on all affected welds. Otherwise, for the established bounding crack growth rate, the licensee could, with appropriate technical justification, opt to operate for another cycle without repairing, then re-inspect and re-analyze to determine what, if any, actions are required. The licensee has determined that a pre-emptive repair will be installed to reinforce the core shroud V9 and V10 welds. It should be noted that inspections are not meant to establish a plant-specific crack growth rate, but to ensure that the bounding crack growth rate is appropriate. The NRC staff will continue to evaluate the licensee's actions to ensure that public health and safety are not compromised.

Comment: In the current inspection plan the horizontal welds in the core shroud will not be inspected. The reason cited in the NMPC inspection plan and the NRC approval letter is that the shroud repair assembly (the tie-rods) willprevent any lateral motion of the core shroud plates and therefore the crack rate in the horizontal welds is_

irrelevant. The assumption is that even if those welds were completely cracked through-wall, the shroud plates would not suffer any lateral motion. However, this idea assumes that the shroud repair assembly is performing as expected. When it was installedin 1995, the shroud repair assembly was thought to replace the structuralfunction of the horizontal welds. As was witnessedin the 1997 inspection, that initial design failed. Only in this inspection will we be able to determine if the second design has failed or not, if that design has failed then the shroud repair assembly has not structurally replaced the horizontal welds and the knowledge of the level of cracking in the horizontal welds is essential to ensure the public safety.

In the current inspection the shroud repair assembly will be thoroughly inspected to ascertain its structuralintegrity. Howeverif the shroud repair assemblyis found to be below specification there is no contingency plan to then inspect the horizontal welds.

The inspection plan assumes the result that the shroud repair essembly will be fully functional. This oversight is a major flaw in the inspection plan.

Enclosure 3

t 2-Response:We do not agree with your conclusion regarding the adequacy of the inspection plan. The NRC has determined that the shroud repair assembly (tie-rods) adequately replaces the function of the horizontal welds. A basic premise in the design of the tie-rods is that the horizontal welds are completely cracked through, 360* around, i.e., the welds are essentially no longer present, leaving separate flat-bottomed cylinders stacked one on top of the other. Even if the horizontal welds were completely cracked through, neither the resistance to lateral movement by friction between the jaggedly cracked surfaces, nor the weight of the shroud and the corresponding resistance to movement, were taken credit for in the design of the tie rods. This scenario of completely cracked horizontal welds is considered unlikely by the NRC staff since residual stresses are relieved by cracking and the operating stresses are small. Further, from a risk perspective, the frequency of the events that provide the major challenges to the repair, i.e., design basis earthquake and large loss-of-coolant accident, LOCA, are remote, with a frequency on the order of 1 x 10 4 events / year. The associated contribution to core damage frequency would be expected to be at least an order of magnitude less. Further, the NRC staff will ensure that the required safety factors will be satisfied at the end of the upcoming operating cycle by reviewing the licensee's inspection results and, depending upon these findings, evaluating the associated fracture analysis, or by the licensee mechanically replacing the vertical core shroud welds. By letter dated April 30, 1999, the NRC staff approved the proposed repair option, and the licensee has repaired the two shroud vertical welds V9 and V10.

We reviewed the licensee's corrective action regarding the tie rod failure and found that it provides reasonable assurance. Further, even with one tie rod failed, it was demonstrated that adequate margin had been maintained for the core shroud.

Continued inspections of the tie rod assemblies will occur to verify they will perform their intended function.

A correction should be noted to your assertion that the "... initial design [of the shroud repair assembly) failed." This is imprecise. While it is accurate to state that the initial placement of one of the shroud repair assemblies was mis-positioned, additional analysis found that the remaining shroud repair assemblies would have performed their intended function. Further, the tie rod was re-positioned to its design location, and, again, will be examined to ensure continueG functionality.

Comment: ...no fundamental safety system should exist without a backup. The shroud repair assembly is a fundamental safety system in that it prevents a lateral shift of the core shroudplates from impinging upon the core. If the shroud repair assembly were to fail then only the horizontal welds wouldprevent such a lateral shift. Without knowledge of the integrity of those welds we do not knowif the backup system exists or what level of safety it provides.

Response: As you acknowledge in your April 15,1999, letter, the tie-rods will be visually inspected to confirm that they are correctly positioned and fully tensioned, thereby ensuring that the tie-rods are performing their intended function. We have previously determined, in the NRC staff's initial review of the NMP1 shroud repair assemblies, that even if one of the four tie-rods were to fail, the remaining three would adequately perform their intended function and allow safe shut-down of the reactor. It is considered highly unlikely that more than one tie-rod could fail simultaneously. Further, the conservatisms in the design assumptions were discussed previously in the NRC staff's SE.

Comment: Finally it is important in fully understanding the mechanism of cracking in BWR con shrouds to gather as complete a data set as is possible. . . You then direct the NMPC to estimate the fluence level distribution through the upcoming operating cycle to assess if the increased crack growth rate might warrant future repairs. I submit to you that this strategy is problematic in that it relies on fluence estimates for the shroud without directly isolating the effect of fluence on the cracking mechanism in the NM1 core shroud.

Response:The NRC staff agrees that understanding the mechanism of cracking is important, and the we are conducting research into the mechanism of both inter-granular and irradiation assisted stress corrosion cracking (IGSCC and 1ASCC) to fully comprehend these phenomena. The results of the research performed to date have been made publically available in the semi-annual research reports on this program.

The latest report was issued in March 1999 (January 1998 - June 1998, NUREG/CR-4667, Vol. 26). We believe that controlled research activities are a much better approach to " isolating the effect of fluence on the cracking mechanism... ." The inspection program for the NMP1 shroud is intended to have sufficient scope and accuracy to provide assurance of core shroud integrity. More basic understanding of the cracking phenomenon is addressed through industry and NRC research programs. This understanding is then used in designing inspection programs and assessing inspection results.