ML20052F905
| ML20052F905 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/29/1982 |
| From: | Dan Collins, Feld S, Murphy E, Rajan J, Rajan J N Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20052F902 | List: |
| References | |
| NUDOCS 8205140128 | |
| Download: ML20052F905 (14) | |
Text
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I-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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SOUTH CAROLINA ELECTRIC & GAS COMPANY
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Docket No. 50-395
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(Virgil C. Summer Nuclear Station,
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Unit 1)
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AFFIDAVIT OF DOUGLAS M. COLLINS, SIDNEY FELD, EMMETT L. MURPHY AND JAI RAJ N. RAJAN REGARDING FAIRFIELD UNITED ACTION APRIL 9, 1982 LATE INTERVENTION PETITION 1.
I, Douglas M. Collins, being duly sworn, state as follows. I am a Section Leader in the Radiological Protection Section of the Radiological Assessment Branch, Division of Systems Integration, NRC Office of Nuclear Reactor Regulation. Sections 23 and 24 of this affidavit were prepared under my direction and supervision. A statement of my professional qualifications is attached.
2.
I, Sidney Feld, being duly sworn, state as follows. I am Regional Environmental Economist in the Antitrust and Economic Analysis Branch of the Division of Engineering, NRC Office of Nuclear Reactor Regulation.
I am responsible for sections 25 thru 27 of this affidavit. A statement of my professional qualifications is attached.
3.
I, Emmett L. Murphy, being duly sworn, state as follows. I am a Materials Engineer in the Materials Engineering Branch, Division of Engineering, NRC Office of Nuclear Reactor Regulation. I am jointly responsible for sections 6 thru 22 of this affidavit. A statement of my professional qualifications is attached.
8205140128 820429 PDR ADOCK 05000395 G
> 4.
I, Jai Raj N. Rajan, being duly sworn, state as follows. I am a Mechanical Engineer in the Mechanical Engineering Branch, Division of Engineering, NRC Office of Nuclear Reactor Regulation. I am jointly responsible for sections 6 thru 22 of this affidavit. A statement of my professional qualifications is attached.
5.
The purpose of this affidavit is to address the matter raised in Fairfield United Action's April 9,1982 late intervention petition concerning the phenomenon of accelerated steam generator tube wear caused by flow-induced vibration in tho Westinghouse Model D-3 preheater type steam generator utilized at the Virgii C. Sumnar Nuclear Station. This affidavit provides a general discussion of the nature of the problem, prospects for an eventual corrective design modification and NRC staff plans and procedures for interim operation. The affidavit contains an estimate of the time required for implementation of the design modification and accompanying occupational exposure. The affidavit also reassesses the cost-benefit analysis performed in the operating license Final Environmental Statement in light of the possibility that full power operation may be restricted for a limited period of time pending implementation of the eventual design modification.
A.
Background
6.
The steam generators for the Virgil C. Summer Nuclear Station are of the Westinghouse Model D-3 series. The staff evaluation of the steam generators is provided in the Safety Evaluation Report (SER) issued in February 1981. The SER contained a confirmatory matter related to the steam generator inspection program and two pote.itial licensing conditions related to tube plugging and inspection ports. A further evaluation of the steam generators 4
4.
was provided in Supplement No. I to the SER issued f r. April 1981. The purpose of that evaluation was to indicate that the confirmatory matter related to the steam generator inspection program had been adequately resolved. A further evaluation of the steam genei itors was provided in Supplement No. 3 to the SER issued in January 1982 to eliminate the proposed condition related to tube plugging.
7.
On January 20, 1982 the staff issued Board Notification BN 82-02 to advise the ASLB of information obtained by the staff relative to the degradation of Model D steam generator tubes which had occurred at two foreign reactors.
The staff also advised the ASLB that we were closely monitoring the inspection and testing program at the McGuire Nuclear Station, Unit I which incorporates Model D-2 preheater type steam generators which are similar to the ones used at the Virgil C. Summer Nuclear Station. The staff stated that we would keep 1
the ASLB informed.
8.
Also, on January 20, 1982, in a letter from D. G. Eisenhut to T. C.
Nichols, Jr., the staff requested that the applicant submit its plans to address this problem. The staff further requested that it address the extent i
to which it planned to rely on the results of the Westinghouse test program
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or testing at operating plants, instrumentation for detection of flow-induced vibrations and the testing and start-up procedures for Virgil C. Summer Nuclear l
Station.
9.
The applicant responded in a letter dated February 19, 1982 from T. C.
Nichols, Jr. to H. R. Denton.
In that letter the applicant described the l
Westinghouse program which included the use of scale model testing, analytical studies, wear testing, periodic eddy current inspection of operating units
. i and the use of diagnostic internal and external instrumentation in selected l
operating plants. The applicant also stated in that letter that it had no plans for installing internal instrumentation in any of the steam generators at Virgil C. Summer Nuclear Station. The applicant described a program for operation of Virgil C. Summer Nucitnr Station pending a permanent design modification to the steam generators by Westinghouse. This included a testing program at power levels up to 50% of full power followed by shutdown and eddy current inspection of the tubes in those rows where problems were experienced on the other plants. The applicant stated that, following this, it would propose an operating power level pending permanent modification to the steam generators.
- 10. Also, on February 19, 1981, the staff held a generic meeting on the Model D steam generators with representatives from Westinghouse. The meeting was also attended by representatives from a nunber of the affected utilities.
At that meeting Westinghouse summarized the results of its initial evaluation of accelerated tube wear at the two foreign plants. Ringhals Unit 3, a three-loop Westinghouse plant in Sweden, was shut down on October 21, 1981 because of l
at 2.6 gpm primary-to-secondary leak. Before the leak, the unit had been operating i
at power levels greater than 50% of full power for approximately five months.
The steam generators, which are the Westinghouse preheat type (Model D), are similar in design to those at McGuire Nuclear Station, Unit 1, the only domestic operating plant with this type of steam generator, and Virgil C. Summer Nuclear Station.
- 11. The leaking tube was located within the preheater section on the cold leg side of the Ringhals Unit 3 steam generator. The eddy current testing l
E.
results revealed numerous tubes with indications localized within the preheater section at baf fle plate locations. The tubes affected are in the peripheral rows (close to the steam generator shell) adjacent to the feeddater inlet.
There are approximately 100 tubes in each steam generator with indications.
Approximately 45 of the tubes with indications have wall reductions of greater than 50%.
- 12. The most recent eddy current testing of the steam generator tubes at Almaraz Unit 1 in Spain also revealed significant tube wall reduction at locations similar to those at Ringhals Unit 3.
Almaraz Unit 1, with steam generators similar to those at Ringhals Unit 3 and Virgil C. Summer Nuclear Station, had been operating at various power levels, including full power, for about four months.
- 13. Westinghouse believes that these vibrations are attributable to vibration of the steam generator tubes from high fluid velocities and that the tube walls are being worn down from rubbing against baffle plates in the preheater sections c7 these steam generators. Westinghouse further believes that a reduction of flow velocity by controlling total feedwater flow should reduce the potential for vibration. Westinghouse stated that the increased turbulence occurred at high feedwater flow rates corresponding to 50 - 70%
of full power operation. Westinghouse stated that it would make recommendations to the various utilities concerning plant operation pending the development of a permanent design modification.
- 14. On April 1,1982 the staff issued a letter to Duke Power Company, a copy of which is attached, to inform them that operation of McGuire Nuclear Station, Unit 1 at 50% of full power was acceptable on an interim basis pending
r receipt of a detailed report describing the basis for continued operation at higher power levels. This was based on the staff evaluation of the testing and inspection that had taken place during the latter part of 1981 and the first two months of 1982.
- 15. Subsequent to the issuance of its letter to the staff dated February 19, 1982, South Carolina Electric & Gas Company reconsidered its position regarding the installation of internal diagnostic instrumentation for purposes of measuring tube vibration in the Virgil C. Summer Nuclear Station steam generators.
In a letter dated April 14. 1981 from T. C. Nichols, Jr. to H. R. Denton, the applicant stated that it planned to install internal instrumentation in the "A" steam generator. The instrumentation will consist primarly of biaxial accelerometers and will be installed in selected tubes. This instrumentation will be in place prior to the issuance of a low-power license. The applicant l
expects that the system will be operational in early May 1982.
16.
In a letter dated April 27, 1982 from T. C. Nichols, Jr. to H. R.
Denton the applicant provided estimates of the schedule for making design modifications to the steam generators, the radiation environment and the total radiation exposure to personnel performing the modification.
(
l B.
Basis for issuance of Low Power License
- 17. As discussed previously, the steam generators used in the Virgil C. Summer Nuclear Station are of the Westinghouse Model 0 design. The generic problem concerning vibration-induced wear in the preheater section of Model D steam generator tubes has been identified at the lead operating facilities with Model D steam generaters, which include McGuire Nuclear Station, Unit 1 and two foreign facilities. The causes of the vibration problem and design
- modifications to correct the problem currently are being evaluated by the indu stry. The available information to date, which includes operating experience at facilities with Model D steam generators, and Westinghouse test data, indicates that the potential for significant wear damage is substantially reduced at power levels less than 50% of full power. Eddy current inspection of the McGuire Nuclear Station, Unit 1 steam generators performed subsequent to low-power physics testing plus 324 hours0.00375 days <br />0.09 hours <br />5.357143e-4 weeks <br />1.23282e-4 months <br /> at or above 50% of full power revealed no evidence of wear degradation. The McGuire Nuclear Station, Unit 1 steam generators were recently reinspected after 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> at or above 50%
of full power and only minor indications of wear degradation were observed.
Significant wear degradation has only been observed to date at the two foreign facilities af ter several hundred hours of operation at or above 75% of full power.
- 18. Based on the above, the staff finds no reason related to steam generator tube integrity to withhold the issuance of an operating license that restricts power to less than or equal to 5% of full power, i.e. a low-power license. The staff will address the applicant's detailed program for interim operation at power levels in excess of 5% of full power at a later date.
In the meantime, the staff is closely monitoring operation of McGuire Nuclear I
Station, Unit 1 and the other foreign operating facilities in addition to following the industry investigations of the cause of the vibration problem and corrective action.
C.
Plans for Licensing Virgil C. Summer Nuclear Station at Power Levels in Excess of 5% of Full Power
- 19. As described in the letter from T. C. Nichols, Jr. to H. R. Denton, l
l
r dated February 19, 1982, the applicant proposes to operate the facility at 50% of full power pending development of a pemanent design modification by Westinghouse. Based on the staff's review of information received to date we believe that a program can be developed and approved to permit safe operation of the facility at or near the 50% of full power level proposed by the applicant pending the development of a permanent design modification.
However, the specific details of such a program need not be developed at this time since the low power test program is expected to take place over the next several months and data are being developed from the operation of other facilities that may be used in the development of the necessary design modifications.
- 20. The staff believes that a program similar to that currently in place at McGuire Nuclear Station, Unit I will be carried out at Virgil C.
Summer Nuclear Station during operation at higher power levels. This program would likely consist of interim operation at power levels below those at which the onset of tube wear vibration was experienced on similar units.
l The diagnostic instrumentation will also serve to indicate the severity of tube vibrations in the Virgil C. Summer Nuclear Station steam generators.
The staff would expect that further confirmation would be obtained by frequent periodic shutdowns of the f acility to perfom eddy current examination of the steam generator tubes.
I
- 21. Following the development and confimation of a permanent design resolution by Westinghouse, and subsequent modification of the Virgil C.
i Summer Nuclear Station steam generators, a program will be implemented to l
confim the design modification specifically for the Virgil C. Summer Nuclear Station. The applicant has stated that the design modification will be potentially
available in the latter part of 1982. The staff expects that the program would make use of the internal diagnostic instrumention and eddy current examination of tubes at escalating power levels until operation at full power can be authorized.
22.
It is the staff's judgement that an acceptable design modification can be developed to permit operation at full power. The staff believes that by means of on-line diagnostic instrumentation and periodic shutdown followed by eddy current examination of the steam generator tubes, the integrity of the tubes can be assured and that plant operation in this manner will not pose an undue risk to the health and safety of the public.
D.
Estimate of Occupational Exposure to Complete Design Modification
- 23. The staff previously reviewed the proposed occupational radiation protection program for Virgil C. Summer Nuclear Station and found it to meet NRC criteria including Regulatory Guide 8.8.
The staff's findings were documented in NUREG-0717, the staff's Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station. The applicant has estimated that repairs to the three steam generators, performed after low power testing, will result in an occupational dose of 450 to 525 person-rems. This estimated dose is within that used to project environmental impacts in NUREG-0719, the staff's Final Environmental Statement related to the operation of Virgil C. Summer Nuclear Station. This dose would not be received by workers if the modification to the steam generators could be made prior to low-power testing.
However, the applicant has stated that it cannot make the design modification prior to startup since it would not be available until the latter part of 1982.
1.
- 24. Because the necessary design modification will not be available until the latter part of 1982, the collective occupational dose is not sufficient to delay startup until the modification is complete, and the applicant has described a radiation protection program to maintain occupational doses As Low As is Reasonably Achievable ( ALARA), the staff finds that the radiation protection aspects of the applicant's approach are acceptable.
E.
Cost / Benefit Balance
- 25. The staff has also reviewed the cost benefit analysis contained in the Operating License Final Environmental Statement (OL-FES) for Virgil C.
Summer Nuclear Station and concludes that there will be no material effect on the cost benefit balance as a result of the tube vibration problem. This conclusion recognizes that until such time as the problem with the Model D-3 steam generator is corrected the Virgil C. Summer Nuclear Station will likely be limited to a maximum performance level of about 50%. The OL-FES cost benefit summary assumed a 60% capacity factor. The applicant has estimated that the design modification potentially will be available by the end of 1982.
However, for the purposes of this review we have conservatively assumed that reduced performance will occur over a several year period.
- 26. The base load electrical energy from the plant was among the benefits cited in the OL-FES from plant operation. The other benefits include improved system reliability, yearly production cost savings, and increased fuel diversity. Possible plant operation at 50% of full power for some limited f
period of time will not significantly diminish the relative value of these l
particular benefits.
I I
c.
- 27. The 60% capacity factor assumed in the OL-FES for Virgil C. Summer Nuclear Station is an expected average value over the life of the facility.
It is based on a statistical analysis of the performance of existing pressurized water reactor f acilities.1 The past performance of these facilities has been affected by down times and deratings due to a wide range of repairs and safety modifications. Therefore, there is no basis to assume that the Virgil C. Summer Nuclear Station cannot absorb a temporary down rating over its initial years and still sustain a lifetime 60% performance level. Furthermore, statistical analyses on nuclear capacity factors show that these reactors experience improvement over the first 4 to 5 years and then level off.2 Thus, the fact that the Virgil C. Summer Nuclear Station may operate at a below average level during its initial years is not inconsistent with past experience of other facilities and does not invalidate the 60% lifetime assumption employed in the OL-FES.
1 Roger G. Easterling, Statistical Analysis of Power Plant Capacity Factors Through 1979, Sandia Laboratories, NUREG/CR-0382, April 1981.
2 1 bid, pp 2-3
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NUCLEAR REGULATORY COMMISSION wassincrow. o. c.aossa krd i b82 Docket Nos: 50-369 and 50-370 Mr. William O. Parker, Jr.
Vice President Steam Production Duke Power Company P.O. Box 2178 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
Subject:
Continued Operation of Unit 1 (McGuire Nuclear Station Units 1 & 2)
By letter of March 12, 1982, we infomed you that operation of McGuire Unit 1, utilizing the Model D steam generator, on an interim basis at 50% power was accep-table pending receipt of a detailed report describing the basis for your continued operation at power level of 50% and above. On March 16, 1982, you filed a report which was to provide the basis for operation of Unit 1 between 50% and 75% over the next several months.
The staff and its consultants have reviewed the contents of the March 16, 1982, report and additionally have had the benefit of telephone discussion with :: embers of your staff and Westinghouse. Based on the infomation provided to us we cannot conclude that it would be prudent to operate Unit 1 above 50% until a demonstrable technical basis is provided to support interim reactor operation at power levels greater than 50%.
In response to your request dated March 16, 1982, we conclude and find acceptable and grant pemission for continued operation of Unit i for a period not to exceed i
1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> as detemined from your mid-March restart at a power level no higher than 50%. However, should you believe that additional data need be obtained at higher power levels, you should submit a proposed testing program and justifica-tion to deal with short tem operation above 50%.
m
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. Mr. William O. Parker, Jr.
We require that the NRC staff be immediately notified in th of severe vibrational behavior or excessive wear during the current operating mode.
Further steam generator eddy current testing shall be perfomed subsequent to this A summary of the results of our period of operation and the results reported to NRR.
review is presented in the enclosure.
Sincerely, EEfE'Lf 4-Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
As stated cc:
See next page l
l l
- _~
Enclosure so REVIEW 0F PROPOSED OPERATIONAL PROGRAM 4
i FOR McGUIRE UNIT NO. I UTILIITE'INEEIF3 TEAK'UNDtATORS Ref: Duke Power Company Letter Report Dated March 16, 1982 INTRODUCTION By letter dated March 16, 1982, Duke Power Company submitted its proposed operating program for McGuire 1 for a period extending to July 4,1982. The proposed program includes operation up to and including 75% power. Operation between 50 and 75%
power would be limited to 60 days from the date of restart following the February 1982 steam generator inspection outage. McGuire would be shutdown for additional inspections at the first practical opportunity following this 60 days above 50%
power, but no later than July 4,1982. The licensee's submittal includes its tech-nical justification for the proposed operating program, which includes what they consider to be an upper bound on the amount of tube wear which will occur during the proposed operating period.
BACKGROUND McGuire 1 had accumulated the following operating history at the time it was shut-down on February 26, 1982:
Power Level Hours at or above this power level 50%
1500 75%
324 90%
72 100%
23 The total number of effective full power hours to date is 1093.
Eddy current (ECT) inspections of the McGuire steam generators during the February outage revealed four tubes with 0.D. indications in steam generator C.
These indications have been attributed to small volume wear defects, less than 20% in depth at the 5th support plate in the preheater section. The licensee estimates that the % through-wall penetrations of these defects is between 5 and 10%, com-pared to the Technical Specification plugging limit of 40%.
It is the staff's understanding that these tubes have not been plugged.
Internal instrumentation was installed in two tubes during the February outage to provide operating data regarding the dynamic response of the tubes. The instru-mentation may also serve to identify the power level associated with the onset of significant vibration activity which may lead to high rate of wear damage to the tubing.
1 I
2 McGuire has operated for approximately 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> since the Febmary outage at a power level less than or equal to 50% with the exception of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during which power was escalated to 75% for purposes of collecting instrumentation data regarding the tube response over this range. Evaluation of this data is in pro-gress, and no preliminary findings have been reported to the staff to date.
The licensee has established its proposed operating program based on the following:
1.
Estimated wear rates as a function of power level. These estimates were detemined on the basis of estimated wear volumes associated with the ECT indications identified in the field during the February 1982 inspection, and on the operating times accumulated at different power levels prior to the February 1982 inspection.
2.
Allowable wear volume.
l The licensee analysis deals directly with wear volumes as opposed to depth of wear consistant with classical models which consider wear rates (in tems of volumetric removal rates) to be a constant for a critical set of parameters.
The licensee has estimated the maximum wear volume associated with the field ECT 1.5 X 10~gs { rom the February 1982 inspection.a be conservatively bound indicatio I
in.
i l
1ength associated with the field indications was.4", and that five flaws observed on the pulled tubes from Almaraz and Ringhals which exhibited lengths rangfy f om f
I
.15 to.5 inches in length exhibited wear volumes ranging from 1 to 7 X 10 in.
However, we have several reservations about the conservatism of this number inclu-ding:
1.
Licensee has provided no estimate of possible error associated with the length estimates of the McGuire indications, nor whether the lengths associated with the pulled tube examinations are based upon ECT estimates or actual measured lengths detemined from the labor-atory. In addition, the licensee has not indicated whether the length estimate were based upon differential or absolute ECT read-ings.
If based on differential signals, taper of the wear flaws could lead to underestimates of the wear length. Even a small underestimation of the flaw length could lead to a large under-estimation of the flaw volume using this methodology. We note that the data base for Ringhals indicates that f g flaw.6 inches in length, the corresponding volume is 4 X 10 n
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2.
The licensee's estimate is based upon a total of five data points from a total of four tubes. We believe the data base to be too small to properly reflect all geometric, mechanical, and flow parameters which may affect the relationship between length and volume of wear defects.
3.
Most importantly, the licensee has performed at least four other wear volume estimates using different methodolojieg which are 4
substantially larger than the assumed 1.5 X 10 in value.
l These estimates range from 3 X 10-4 n3 o 8 X 10-4in3 The 1
t licensee has not provided adequate justification for excluding any of these estimates from its analysis.
Regarding allowable defect volume, the licensee has establised 10-3 n3 as an allow-1 able defect volume on the basis of volumstric wear vs. wear depth data from the Ringhals and Almaraz pulled tube examinations (Figure 14 of submittal). For defect penetrations ranging between 20 and 50% thgough wall, three data points exist indi-i cating flaw volumes from 4.5 to 7 X 10-31nd. A best fit regression line through vglumeof4X10-3nJassociatedwiththeplugging 1
all the data indicates a weagin criterion was selected by the licensee to reflect limit depth of 40%. The 10-I uncertainties in the method and models used and to account for any increase in wear I
rute which might occur due to the onset of wear and the increase in the diametral gap between the tube and the baffle plate.
We {ing that the licensee has not provided adeguate justification to support theWe I
i estimate.
10 ind defect which penetrated 100% through wall at Ringhals in October 1981 was not included among the data plotted in Figure 14 of the licensee's submittal. This data point deviates sharply from the data trends and regression line pictured in this Figure, in the sense that the volume associated with the leaker is substan-tially smaller than what would be estimated on the basis of the infomation in the This data point indicates to us that actual flaw penetrations can vary Figure.
quite substantially from the trend shown in Figure 14 of the submittal, even from a small 4 tube sample. Considering the large number of tubes located in the outer four rows in the preheater section, we would expect numerous instances of such deviations.
Apart from the methods used to estimate allowable operating times as a function of power level, we have concerns regarding the operation at power levels exceeding 50%.
Westinghouse analysis of Ringha1/Almaraz data indicates significant increase in vibration activity and consequent tube wear at power levels above 50% to 75% (Ref.
Westinghouse meeting of February 19, 1982, at Bethesda). Our independent analysis and review of plant operating experience support this conclusion. The RMS spectra
= -.
l 4-of time varying data are used to indicate the characteristics of the transducer out-The increase in tube activity level is indicated by the growth and broadening puts.
of the spectral peaks in descrete frequency intervals. A qualitative assessment of the spectra for one of the accelerometers in tube row 49, column 51 from Almaraz indicates that vibration activity represents a power curve (somewhere between a i
square and cube curve) and activity increases significantly above 505 power levels.
The definition of the tube excitation mechanism in tems of flow rate, tube motion (mode / frequency) and amplitude of motion has not been established. The excitation i
mechanism appears to be of a ' threshold type' - probably a type of fluid-elastic instability. However, turbulent buffeting or some combination of turbulent buffet-ing and fluid-elastic instability cannot be ruled out as a fom of the excitation mechanism.
If turbulent buffeting is the dominant mechanism, wear can be expected at all power levels and there would be no increase in wear at a certain power level or as the report refers ' onset of wear', If the excitation mechanism is predomin-antly of the fluid-elastic instability type, there will be threshold values assoc-isted with it. The objective of the licensee's program should be to identify these threshold values and avoid these instabilities by operating at power levels suffi-ciently removed from these critical values. The subject report does not address the tube excitation phenomena in sufficient detail;instead an attempt has been made to use wear measurement data to detemine threshold power levels (flow rates) for the onset of large amplitude vibration which would be responsible for rapid wear and tube failures.
In our opinion, the threshold values can easily and accurately be detemined by tube vibration measurements from internally mounted accelerometers. This vital informa-tion is unavailable at this time.
i EVALUATION AND CONCLUSIONS The available evidence to date suggest that the potential for significant vibration activity and high wear rates becomes significant for power levels exceeding 50 to 75% power. To justify operation beyond 50% power requires that wear rates and allowable wear volumes be firmly established on a conservative basis. As noted in our earlier discussion, we have a number of questions and concerns about the con-servatism of the licensee estimates as developed in their March 16, 1982, submit-tal. Unless these concerns can be resolved to our satisfaction, operation beyond 50% power can only be approved after the licensee has submitted sufficient data (including data from internal instrumentation) to indicate that significent vibration activity which may lead to high wear rates will not occur, or after r
steam generator modifications have been installed to correct the problem.
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In the interim, we find that McGuire may be operated at power levels not to exceed 50% for a maximum of 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> from the date of restart following the most recent steam generator inspection, witbout undue risk to public health or safety. This finding is based upon the following:
1.
McGuire has operated for 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> at power levels at and above 50%
power, including 324 hours0.00375 days <br />0.09 hours <br />5.357143e-4 weeks <br />1.23282e-4 months <br /> at and above 75% power, while incurring only minor degradation (<20% through-wall penetration) on four tubes.
2.
Preliminary results of Westinghouse analyses and tests, and preliminary data from the internal instrumentation at Almaraz 1 indicate that the onset of significant vibration of the tubing does not occur until power has been escalated beyond 50% power. Thus, we expect any further pro-gression of wear during the current 1500 hour0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> operating period would be small.
3.
Restrictive limits on allowable primary to secondary leakage in the Technical Specifications provide added assurance of adequate tube integrity.
4 The staff will consider an extension to this 1500 hour0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> authorization to July 4,1982, provided the licensee submits its analyses of the internal instrumentation data, which is currently in progress, by April 30, 1982.
Finally, we are requiring that the NRC staff be immediately notified in the event of any information (primary to secondary leakage or internal instrumentation) which may be indicative of severe vibrational behavior or excessive wear during the current operating mode.
1 I
i I
PROFESSIONAL QUALIFICATIONS Douglas M. Collins Radiological Assessment Branch Division of Systems Integration I am presently mployed with the U. S. Nuclear Regulatory Commission, within the Office of Nuclear Reactor Regulation, as Leader of the Radiation Protection Section (RPS) of the Radiological Assessment Branch. My responsibilities include super-vision of, as well as direct participation in the technical review and evaluation of the occupational radiation protection aspects of reactor design features, equip-ment and programs described in license applications and license amendment requests.
l These reviews include evaluation of shielding design, actions to maintain occupational doses as low as is reasonably achievable (ALARA), radiation protect'on administrative controls, radiation protection facilities and equipment, and staffing and i
qualifications of the radiation protection group. RPS also evaluates the occupational ALARA actions associated with major plant modifications such as steam generator replacment and retubing. The section also participates in the development of regulations and guidance in the area of occupational radiation protection. Since i
I joined RPS we have worked in the update of or development of regulatory guides i
for occupational ALARA, radiation protection and risk training, bioassay, radiation protection instrumentation, shielding, and respiratory protection. The section has also developed a program for impleentation of occupational ALARA at operating power reactors.
i My undergraduate study was done at Spring Hill College, in Mobile Alabama, where I received a B.S. degree in Physics in 1969.
I received an M.S. degree in Physics from Georgia Tech in 1970, with a concentration in Health Physics.
I continued l
graduate studies at Georgia Tech in Nuclear Engineering until 1971.
I achieved Certification as a Health Physicist in 1977. After leaving school in 1971, I began sployment with General Electric in San Jose, California, as Supervisor, Radiation Protection for the fuel fabrication and research and development facility. There I directed the radiation protection program including the occupational exposure control, effluent release control, environmental monitoring and mergency planning programs.
I also provided radiation protection training i
to the San Jose plant staff and GE Service engineers.
In 1973 I joined the AEC as a Health Physicist in what is now the Office of Nuclear Material Safety and Safeguards, reviewing the radiation protection programs for applications to use radioactive materials in industry, medicine, and research and develcpnent.
I also evaluated the radiation protection aspects of uranium mill applications.
In 1976 I became an inspector from our Region II office, performing radiation protection inspections of fuel fabrication facilities, power and research reactors, and research and development facilities. These inspections included evaluation of impimentation of external'and' internal dose control; solid, liquid and gaseous processing and control; staffing and training; contamination control; surveys; administrative controls; instruments and equipment; environmental monitoring and mergency planning.
I performed inspections of the Browns Ferry emergency planning upgrades after the fire and the Surry steam generator replace-ment program.
I also assisted in the onsite mergency actions at TMI-2 shortly after the accident. I became Section Leader of the Radiation Protection Section l
in 1979.
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f PROFESSIONAL QUALIFICATIONS SIDNEY E. FELD U. S. NUCLEAR REGULATORY COMMISSION I am Sidney Feld, Regional-Environmental Economist with the Antitrust and Economic Analysis Branch, Division of Engineering of the Regulatory Staff of the Commission.
I served with the Staff from July 1973 to August 1974, and rejoined the Staff in October 1975. I am responsible for reviewing and analyzing Applicants' environ-mental reports and preparing economic input for the Regulatory Staff's Environmental Statements. Over the last several years I have devoted most of my attention to Need for Power Analyses, and I was the principal author of the Staff's Standard Review Plan on Need for Facility. I have prepared testimony on need for power, conservation of energy issues, and various economic cost issues for the hearings on Alvin W. Vogtle Nuclear Power Plant, the Shearon Harris Nuclear Power Plant, the Wolf Creek Generating Station, Midland Plant, the Pilgrim Nuclear Generating Station Unit 2, the Zimmer Nuclear Power Station, and the Lacrosse Boiling Water Reactor.
I received a B.B.A. Degree in Economics from the City College of New York in 1967, an M. A. Degree in Economics from the University of Rhode Island in 1969, l
and a Ph.D. Degree in Resource Economics from the same university in 1973. My graduate degree in resource economics focused on the application of economic theory to public resources. Areas of Study included:
simulation of market economic solutions; consideration of social implications such as environmental impacts; and the application of decision tools such as cost-benefit analysis.
From September 1974 through August 1975, I was an Assistant Professor of Resource Economics at the University of New Hampshir9 at Durham, New Hampshire.
In this capacity, I taught courses in Resource Economics and Statistics.
I also served as co-investigator on a Sea Grant research. project to examine economic activity in the New Hampshire Coastal Zone.
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.c EMMETT L. MURPHY DIVISION OF ENGINEERING _
0FFICE OF NUCLEAR REACTOR REGULATION PROFESSIONAL QUALIFICATIONS My name is Emmett L. Murphy.
I am a Materials Engineer in the Inservice Inspection Section, Materials Engineering Branch, Division of Engineering.
Office of Nuclear Reactor Regulation, of the United States Nt: clear Regulatory Coramission.
In g present position I am responsible for performing technical reviews and evaluations of PWR steam generator tube surveillance and repair programs for NTOL and operating plants.
I hold a Bachelor of Science Degree in Aerospace Engineering and a Master of Science Degree in Civil Engineering, both from the University of Maryland.
I have had a total of ten years of professional experience of which eight years has been in tne nuclear field.
I was employed for almost six years as a structural engineer at the Bettis Atomic Power Laboratory by Westinghouse Corporation.
During my employment at Bettis, I was involved in the structural design and analysis of core and core structurals of naval reactors.
Since joining the NRC in July 1979, I have been involved exclusively in the steam generator review area.
I have been involved in the safety reviews of most of the steam generators which have experienced significant tube cegradation during the past two years, including Point Beach Units 1 and 2 and San Onofre Unit 1.
PROFESSIONAL QUALIFICATIONS JAI RAJ N. RAJAN U. S. NUCLEAR REGULATORY COMMISSION MECHANICAL ENGINEERING BRANCH DIVISION OF ENGINEEP.ING I am a mechanical engineer responsible for reviewing and evaluating safety analysis reports with regard to mechanical engineering aspects of components, the dynamic analyses 'and testing of safety related systems and components and the criteria for protection against the dynamic effects associated with postulated failurcs of fluid systems for nuclear facilities. I am the Mechanical Engineering Branch's principal reviewer on the issue of the structural integrity and plugging criteria of degraded steam generator tubes.
I am also responsible for the review and ev'aluation of kibration problems of a generic nature in the piping systems and components of nuclear facilities.
I received a B.S. degree in 1953 from Lucknow University India majoring in Physics, Mathematics and Chemistry.
In 1956 I received a B.S. in Civil Engineering from Roorkee University, India majoring in Structural and Hydraulic Engineering.
In 1962 I received a M.S. degree from Duke University majoring in Applied Mechtnics and Ph.D. degree in 1966 from the same l
university with majors in Fluid Mechanics. From 1960 to 1962 I was an l
instructor in structural e'ngineering at Duke UniUersity. From 1962 to 1966 I was employed by the U.S. Army Research Office in Durham, N.C. as a research engineer conducting theoretical and experimental research in l
s' high pressure pneumatic and hydraulic shock tubes and investigating wave From 1966 to 1973 I worked as a project propagation phenomenon in pipes.
mechanical engineer and subsequently as a senior project mechanical engineer at the Naval Research and Development Center at Annapolis, Md.
tbjor projects involved design analysis, test and evaluations of fluid piping systems and power fluid systems of advanced nuclear submarines.
Investigations were multidisciplinary in scope utilizing advanced Mathematical models of power plant machinery and piping techniques.
systems of nuclear submarines were developed and analyzed to determine Thermo-system response to flow induced vibrations and hydraulic shock.
dynamic and hydrodynamic analyses of naval boilers and steam plants were conducted including full scale tests.
In April of 1974 I joined the U. S. Atomic Energy Commission prior to the formation of the U. S. Nuclear Regulatory Commission and have remained with the Mechanical Engineering Branch of the Division of Engineering as a mechanical engineer performing the type of work as previously described.
I have taught at the University of Maryland on a part-time basis since l
1967 both at the graduate and undergraduate levels in courses of mechanics of-materials, fluid mechanics and applied mechanics.
Publications include Journals of AIAA and ASME and I am an associate member of Sigma Xi honor society.
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i UNITED STATES OF AFERICA NUCEAR REGUIAIORY COMISSION BEFORE 'UE AlmIC SAFEIY AND LICENSING BOARD In the Matter of
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SOUDI CAROLINA EECTRIC & GAS
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COPANY
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Docket No. 50-395
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(Virgil C. Sumner Nuclear Station,
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Unit 1)
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CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO SECOND LATE INTERVENTION PETITION OF FAIRFIELD UNITED ACTION" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, by deposit in the Nuclear Regulatory Commission's internal mail system, this 29th day of April, 1982:
Micrbert Grossman, Esq., Chairman Richard P. Wilson, Esq.
Administrative Judge Assistant Attorney General Atomic Safety and Licensing Board S. C. Attorney General's Office U.S. Nuclear Regulatory Cormission P. O. Box 11549 Washington, DC 20555 Columbia, South Carolina 29211 Dr. Frank F. Hooper Joseph B. Knotts, Jr.
Administrative Judge Debevoise & Liberman School of Natural Resources 1200 Seventeenth Street, N.W.
University of Michigan Washington, DC 20036 Ann Arbor, Michigan 48109 Randolph R. Mahan, Esq.
Mir. Gustave A. Linenberger S. C. Electric & Gas Company Administrative Judge P. O. Box 764 Atomic Safety and Licensing Board Columbia, SC 29218 U.S. Nuclear Regulatory Comnission Washington, DC 20555
- Atomic Safety and Licensing Board Panel George Fischer, Esq.
U.S. Nuclear Regulatory Cm mission Vice President and General Counsel Washington, DC 20555 South Carolina Electric and Gas Company P. O. Box 764
- Atomic Safety and Licensing Appeal Columbia, South Carolina 29202 Board Panel U.S. Nuclear Regulatory Comnission Brett Allen Bursey Washington, DC 20555 Route 1, Box 93-C Little Mountain, South Carolina 29076
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- Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Comnission Washington, DC 20555 John C. Ruoff P. O. Box 96 Jenkinsville, SC 29065 A
Steven C. Goldberg u Counsel for NRC Staff l
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