ML20028B219
| ML20028B219 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/22/1982 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Clark R, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-48752, NUDOCS 8211300179 | |
| Download: ML20028B219 (12) | |
Text
t nfsconsin Electnc rowen coursur 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 November 22, 1982 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S.
NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention:
Mr. R. A. Clark, Chief Operating Reactor Branch 3 Gentlemen:
DOCKET NO. 50-266 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION STEAM GENERATOR REPAIR POINT BEACH NUCLEAR PLANT, UNIT 1 Attached is our response to the request for additional information, as requested by Mr. T. Colburn of your staff, regarding the Point Beach Nuclear Plant, Unit 1, steam generator repair.
Should you have further questions, please contact us.
Very truly yours, i
/
Assistant Vice President i
C. W.
Fay Attachment Copy to ASLB Service List 00/
NRC Resident Inspector l
8211300179 821122 PDR ADOCK 05000266 I
l p
PDR t
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of
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WISCONSIN ELECTRIC POWER COMPANY
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Docket Nos. 50-266
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50-301 (Point Beach Nuclear Plant,
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(OL Amendment)
Units 1 and 2)
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SERVICE LIST Peter B.
Bloch, Chairman Stuart A. Treby, Esq.
Atomic Safety and Licensing Office of the Executive Board Legal Director U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Wasington, D.C.
20555 Washington, D.C.
20555 Dr. Hugh C. Paxton Richard G.
Bachmann, Esq.
1229 - 41st Street Office of the Executive Los Alamos, New Mexico 87544 Legal Director U.S. Nuclear Regulatory Commission Dr. Jerry R. Kline Wasington, D.C.
20555 Atomic Safety and Licensing Board Kathleen M.
Falk, Esq.
U.S. Nuclear Regulatory Commission Wisconsin's Environmental Decade Washington, D.C.
20555 114 North Carroll Street Suite 208 Atomic Safety and Licensing Madison, Wisconsin 53703 Boa d Panel U.S. Nuclear Regulatory Commission Francis X. Davis, Esq.
Washington, D.C.
20555 Monroeville Nuclear Center Westinghouse Electric Corporation Atomic Safety and Licensing P. O. Box 355 Appeal Board Panel Pittsburgh, PA 15230 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Barton Z. Cowan, Esq.
John R. Kenrick, Esq.
Docketing and Service Section Eckert, Seamans, Cherin & Mellott Office of the Secretary Forty-Second Floor U.S. Nuclear Regulatory Commission 600 Grant Street Washington, D.C.
20555 Pittsburgh, PA 15219
ATTACHMENT 1 LICENSEE'S RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO POINT BEACH NUCLEAR PLANT UNIT 1 STEAM GENERATOR REP /IR QUESTION 1 With regard to heavy load handling during the proposed steam generator replacement, we will nead the following commitment in accordance with ANSI B30.2.0, " Overhead and Gantry Cranes", as referenced in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants".
Maintain complete records on the polar crane of all heavy loads handling a.
operations during the repair phase.
b.
Perform the indicated inspections and load tests on the polar crane after completion of the repair phase and prior to returning the crane to its nor-mal service.
RESPONSE
a.
Special heavy lifts during steam generator repair, as defined by ANSI B30.2.0-1976, will meet the requirements of Section 2-3.2.1 which details the require-ments for the handling of special heavy lifts.
Additionally, normal indus-trial and construction standards for the handling of high value equipment will be observed'.
As stated in Section 3.2.6.2 of the Repair Report, the handling of heavy loads in the containment has no safety impact on plant operation, since all fuel will be removed from the reactor during the steam generator repair.
Therefore, it is considered that the detailed requirements and procedures which have been implemented in response to NUREG-0612 are not applicable to the steam generator repair activities.
Records of special heavy lifts will be maintained in project records as part of the Control Work Packages discussed in Section 3.2.6.3 and Section 6, paragraph 6.4, of the Repair Report.
b.
The inspections indicated in ANSI B30.2.0-1976, Chapter 2-2, will be per-formed after completion of the steam generator repair and prior to return-ing the crane to its normal service.
The existing overhead trolley will not be used for special heavy lifts dur-ing the steam generator repair and will not be modified in any manner.
- Thus, a load test for the overhead trolley will not be required.
The design stres-ses for the polar crane girders will not be exceeded utilizing a center post.
Pre,sently, no structural modifications to the bridge are anticipated.
There-fore, a load test for the bridge will not be required.
In the event that bridge modifications are required, the lifts performed during the repair project will constitute an adequate load test of the modifed bridge in lieu of the' load test required by ANSI B30.2.0-1976, paragraph 2-2.2.2.
QUESTION 2 Verify that the steam generator repair or replacement will not introduce a feedring design which will create a potential for steam generator waterhammer or commit to performance of waterhammer tests as a part of the return-to-service testing program by automatic initiation of auxiliary feedwater in accordance with BTP-ASB-10-2.
RESPONSE
Steam generator waterhammer is believed to be caused by the rapid condensation of steam within the feedwater line.
The presence of steam in the feedwater line can occur during hot standby operation when the steam generator water level drops below the feedring elevation, allowing the feedring to drain and steam to enter the piping.
The subsequent addition of cold auxiliary feedwater can cause the steam to condense rapidly, resulting in waterhammer.
i l
l The Model 44F replacement steam generators utilize a feedring designed to limit drainage during such operations by the use of top discharge J-nozzles and a.
. welded feedwater nozzle thermal liner.
This configuration has been demonstrated by field experience to effectively limit the potential for waterhammer; examples include steam generators backfit with J-nozzles, operating experience with the Model 51F steam generators at Surry 1 and 2, and operating experience with the Model 44F steam generators at Turkey Paint 3 and 4.
In addition, the feedwater piping configuration at Point Beach Nuclear Plant includes feedwater check valves located close to the steam generator feedwater nozzles.
This configuration minimizes the potential for significant quantities of steam in the feedwater piping and thus minimizes the potential for waterhammer to occur.
Based upon the design of the feedring for the replacement steam generators, field experience with steam generators utilizing this design, and experience with the piping configuration at Point Beach, the performance of waterhammer tests fol-lowing repair of the steam generators is not necessary.
QUESTION 3 Loose parts and foreign objects left inside steam generators have been identified as the cause of at least two steam generator tube rupture events.
Recent inspec-tions have found a variety of foreign objects in the secondary side of steam gen-erators.
Plesse describe the precautions and programs that will be implemented to preclude the incorporation of foreign objects into the steam generators during the current repair procedures as well as future surveillance methods to be imple-mented.
RESPONSE
Procedures will be implemented to preclude the introduction of foreign objects into the steam generators during the repair.
There procedures include a combina-tion of physical barriers and administrative controls.
Physical barriers will be specified as part of the Control Work Packages, consistent with the work to be t
performed.
Physical barriers such as herculite, metal plates, and decking will be used in work areas, as appropriate.
Administrative procedures will include procedures for pers'onnel access control, tool control and log-in and log-out procedures.
The administrative cont'rols will include such requirements as lanyards on small equipment and personal items, and design features such as lock wires on equipment to prevent loss of material in the steam generators.
Following the repair, final inspection and search will be performed on steam generator secondary side.
This search will be performed by inserting a fiberscope through the steam generator handholes and conducting a 360 degree search of the annulus at the tubesheet.
Any foreign objects which are judged to have the potential for steam generator tube damage, and which are accessable, will be removed.
Surveillance of the steam generators during subsequent operation could include periodic inspections of the tube bundles using fibre optic or television tech-niques during refueling shutdowns, continuous monitoring via loose parts monitoring systems, or a combination of these.
It is our understanding that recommendations for loose parts surveillance are currently being developed by the NRC staff and will be made available for comment prior to implementation.
Loose parts surveillance programs during subsequent operation will be developed following further definition of these recommendations.
In any event, loose. parts surveil-lance programs during operation, if required, would be implemented whether or not the steam generators had been replaced.
QUESTION 4 The Licensee should provide information demonstrating that decontamination of the steam generators would result in no significant dose reduction to plant personnel, as stated in Table 6-1, page 14 of 42, item 4..
RESPONSE
The ALARA approach to the steam generator replacement at Point Beach Unit 1 has been reviewed with regard to the estimated savings in personnel exposure which might be expected with decontamination of the steam generators prior to removal.
The work sequence has been established such that the work in close contact with the steam generators required to remove and transport them to storage locations has been minimized, thus minimizing the amount of personnel exposure.
The highest source of radiation will result from the top of the steam generator tube bundle, and from the opening to the steam generator channel head.
The removal procedures require welding shielded covers over the reactor coolant nozzles, a cover over the U-bend transition cone, and on all sample lines, instrument penetrations, and the blowdown line.
The intent of these cones is to use the shell as a shielded transport cask for removing the radioactive steam generator internals to the temporary storage area on site.
This technique has been used in the past (Surry, Turkey Point) and is a proven method of removing the steam generators with minimal exposure to repair personnel.
. Based on a review of state-of-the art contamination processes, a chemical process would be required in order to effectively decontaminate the steam generator tubes.
The oxide film on the internal steam generator surfaces is tightly adherent and water washing or flushing processes are not expected to signifi-cantly reduce radiation levels in the steam generators.
Radiation exposures would be received during the chemical decontamination, as well as in processing packaging and handling large volumes of radioactive waste generated by the processes.
Occupational exposure estimates for tasks carried out in preparation for steam generator decontamination are listed in Table A.2 of NUREG/CR-1595, as follows:.
Remove Manway Covers 20 man-rem Clean-up Manway Entries 80 man-rem Remotely place inflatable plugs in coolant inlet and outlet 20 Man-rem Exposures associated with the chemical decontamination process are estimated to be an additional 24 man-rem in Table 8.1 of NUREG/CR-1595.
Thus, a total in the order of 140 to 150 man-rem per steam generator could be expected for chemical decontamination prior to removal.
The exposure savings which could be anticipated from chemical decontamination of the steam generators (including the steam generator tubing) is conservatively estimated to be 286 man-rem, assuming a dose reduction factor of 10 after decon-tamination.
This was calculateo by considering the steps in Table 6-2 of The Point Beach Unit 1 Steam Generator Repair Report which would be affected by the decontamination process.
The NUREG-CR-1595 data indicate that 140 to 150 man-rem per steam generator or 280 to 300 man-rem total would be associated with the decontamination effort.
Thus, no net exposure savings would be expected froi chemical decontamination of the steam generators prior to removal.
The radiation dose to workers associated with transport and long-term storage is estimated to be only 10 man-rem per steam generator in NUREG-CR-1595.
Assuming a decontamination factor of 10, this could be reduced to 1 man-rem per steam generator or a reduction of 18 man-rem total.
This additional reduction in long term radiation exposure is insignificant when compared to the radiation exposure required to perform the decontamination.
. a decontamination factor of 10, this could be reduced to 1 man-rem per steam generator or a reduction of 18 man-rem total.
This additional reduction in long term radiation exposure is insignificant when compared to the radiation exposure required to perform the decontamination.
QUESTION 5 The Licensee should provide a commitment to train personnel in accordance with Regulatory Guides 8.13, 8.27, and 8.29 or submit acceptable alternatives.
RESPONSE
All contractor personnel requiring controlled side entry into the Point Beach facility during steam generator repair will receive radiological protection training prior to working unescorted in Radiation Controlled Areas as outlined in Section 3.3.5.4 of the Repair Report.
The orientation program is conducted by qualified training staff using written training materials, oral presentation, video tape presentation, question-answer session, and a written examination to verify adequate comprehension of the information presented.
The scope of this orientation program is responsive to the recommendations in Regulatory Guides 8.13, 8.27, and 8.29.
Additional radiation protection aspects which are responsive to Regulatory Guide 8.27 recommendations for use of mock-up and work tasks training are out-lined in Sections 3.3.5.5 and 6.3 of the Repair Report.
This training will be performed by qualified personnel experienced in the tasks to be performed to assure that trainees are able to perform the work in a satisfactory and safe l
manner.
Qualification sign-off sheets and records of training attendance will be maintained.
The requirements for training will be included in the Project Procedures Manual.
-y-
- QUESTION b Section 2.1.4 should include a commitment to the following Regulatory Guides or the Licensee should justify their exclusion:
Regulatory Guides 1.39, 1.58, 1.88, 1.94, 1.44, and 1.46.
RESPONSE
These commitments have been made previously.
Licensee's October 27, 1982 letter included commitments to the guidance of Regulatory Guides 1.58, 1.88, 1.144, and 1.146 as applicable to the design and fabrication of the replacement steam generators and Regulatory Guides 1.39, 1.58, 1.88, 1.94, 1.144, and 1.146 as applicable to the installation of the steam generators with the clarifications noted in Attachment 1 to that letter.
QUESTION 7 As principal contractor of WE for the steam generator reoair, Westinghouse should have its QA program descriptions (s) submitted to and found acceptable by the NRC.
WCAP-8370, which describes the QA program for the Westinghouse Nuclear Technology and Nuclear Components Divisions, has been so processed.
It should be referenced in Section 3.6.4 of the WE report as it is already referenced in Section 3.6.3.
WCAP-9245, which describes the QA program for the Westinghouse Nuclear Service Division, should be submitted for NRC Staff review.
RESPONSE
Licensee's October 27, 1982 letter indicated a commitment to reference WCAP-8370 in Section 3.6.4 of the Repair Report.
WCAP-9245 (Revision 6) was also submitted for information as A.ttachment 2 to that letter.
A supplement to WCAP-9245 (Revi-sion 6) is being prepared which will specifically address the Point Beach steam generator repair activities.
This supplement will be submitted for NRC Staff infonnation when it becomes available.
QUESTION 8 Will there be any change in the amount of demineralizer waste discharged to Lake Michigan as a result of replacing the steam generators?.
RESPONSE
Demineralized water from the makeup water treatment system is supplied to various systems at Point Beach Nuclear Plant and is supplied to the secondary systems to replace water removed via steam generator blowdown.
It is expected that steam generator blowdown flow rates following repair of the steam generators will be similar to those for the existing steam generators.
Thus, it is not expected that significant changes will occur in the total quantities of neutralized regenerant waste released to Lake Michigan from the makeup water system.
QUESTION 9 Are any changes to the WPDES permit anticipated as a result of this action?
RESPONSE
The repair of steam generators will not significantly affect waste releases to Lake Michigan, and it is not expected that changes to the WPDES permit will be necessary.
QUESTION 10 Will the parking lot and layout areas be on previously distrubed areas?
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RESPONSE
Parking and laydown areas described in Section "7 of the Repair Report are located in areas which were previously disturbed for parking, laydown and con-crete batch plant operations during construction of Point Beach Nuclear Plant.
As stated in Section 7.2.2 of the Repair Report, these areas are utilized min-imally by species of fauna known to inhabit the site.
QUESTION 11 Will reconstruction of the barge slip involve a Corps of Engineers permit?
If so, please provide a copy.
RESPONSE
It is presently planned to ship the replacement steam generators by rail to Kewaunee, Wisconsin and by road transport from Kewaunee to the site.
- Thus, reconstruction of the barga slip at the site is not necessary and a Corps of Engineers permit is not required.
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