ML20236T434
| ML20236T434 | |
| Person / Time | |
|---|---|
| Issue date: | 10/09/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| TASK-020, TASK-061, TASK-20, TASK-61, TASK-A-03, TASK-A-04, TASK-A-05, TASK-A-17, TASK-A-3, TASK-A-4, TASK-A-46, TASK-A-49, TASK-A-5, TASK-OR ACRS-2526, GL-87-02, GL-87-2, NUDOCS 8712010238 | |
| Download: ML20236T434 (26) | |
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DATE ISSUED:
10/9/87
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SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON I'
THE GENERIC ITEMS SEPTEMBER 30, 1987 WASHINGTON, D.C.
INTRODUCTION
-The ACRS Subcommittee on Generic Items held a meeting on Wednesday, September 30, 1987, at 1717 H Street, N.W., Washington, D.C., to contin-ue the discussion on the effectiveness of the NRC Staff programs that address generic issues and Unresolved Safety Issues (USls). The entire meeting was open to public attendance. Mr. Sam Duraiswamy was the cognizant ACPS Staff Engineer for this meeting. A list of documents submitted to the Subcommittee is included in Attachment A, and a copy of the presentation schedule for the meeting is included in Attachment B.
ATTENDEES ACRS:
C. P. Siess (Subcommittee Chairman)
J. C. Eberscle, C. Michelson, and C. J. Wylie Sam Duraiswamy (Cognizant ACRS Staff Engineer)
Principal NRC Speakers:
T. Y. Chang, F. Rosa, A. Serkiz, R. Frahm, P. Norian, W. Schwink, and R. Hernan EXECUTIVE SESSION Dr. Siess, the Subcommittee Chairman, convened the meeting at 8:30 a.m.
and stated that the purpose of this meeting was to continue the dis-cussion with the NRC Staff on the process dealing with generic issues and USIs and gather information for use by the ACRS in preparing a report to the Commission on the effectiveness of this process. He said that the Subcommittee had received neither written comments nor requests for time to make oral statements from members of the public.
8712010238 871009 DESIGNATED ORIGITAL 6
t Generic Items Meeting Minutes September 30, 1987 Dr. Siess said that on September 8, 1987, Mr. Wylie and him met with the Project Managers associated with the Duke Power Company plants. (0conee Units 1-3, McGuire Units 1 and 2, and Catawba Units 1 and 2) to explore their role in imposing resolved generic issues and USIs, and also to gather information on the coordination between the Project Managers and.
the NRC resident inspectors and licensees. At that meeting, it was decided that it would be helpful to hear a presentation from'representa-tives of the Duke power Company with respect to the steps involved in implementing resolved generic issues and USIs. Arrangements were made with the Duke Power to participate at the Subcommittee meeting. How-ever, due to some miscommunications among the personnel at Duke Power Company, they were not able to participate et this meeting..
Dr. Siess said that the Subcommittee may have to decide whether it still wants to meet with some selected licensees to discuss the implementation process.
If a decision is nade to meet with certain licensees, another Subcommittee meeting will be held sometime in the future. After that meeting, he plans to prepare a draft report on the effectiveness of the NRC Staff programs that deel with ceneric issues and USIs end submit it to the full Committee for consideration.
If the full Committee wants to hear presentations from the Staff on certain items, arrangements will be made for such presentations.
Mr. Michelson suggested that it would be helpful if the Staff, during their presentations, addressed how they plan to apply the resolved generic issues and USIs to future plants.
l PRESENTATION BY THE OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) l The Subcommittee heard presentations from RES representatives on select-ed generic issues and USIs with respect to the following:
- Nature of resolution 1
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Generic Items Meeting Minutes September 30, 1987
- Manpower, time, and budget involved in the resolution I
- Nature of research, if any, conducted for use in the resolution
- Other factors involved in the resolution.
USI A-49, " Pressurized Thermal Shock" - Mr. R. Woods Mr. Woods stated that the Pressurized Thermal Shock (PTS) issue surfaced after the 1978 event at the Rancho Seco nuclear plant. Deterministic calculations of the Rancho Seco event indicated that, under different circumstances, the event could have caused the reactor vessel to fail.
Since there was very limited information, they found it difficult to ouintify the actual risk from such an event. The PTS issue was desig-nated as a USI in December 1981.
In Decerber 1982, they developed a proposed rule containing the possible resolution of the PTS issue.
After going through the cycle involved in issuing a rule, a final' rule containing requirements to prevent or mitigate potential PTS events was published in July 1985 (Attachinent C, Page 1).
Mr. Woods said that it took about 20-30 personnel Staff years to resolve the PTS issue. He is'not sure about the exact amount of money spent in resolving this issue.
However, the probabilistic risk analysis per-formed by the Oak Ridge National Laboratory (ORNL) consumed a funding of q
about $6 million. The cost for the. consequence study performed by the Pacific Northwest Laboratory was about $0.a million.
In addition, results from a certain pertion of the Heavy Section Steel Technology q
(HSST) program were used in the resolution of the PTS issue.
.)
1 Mr. Ebersole commented that, based on the basic characteristics of the pressurized rater reactor (PWR) plants, the Staff should have identified the PTS problem prior to the Rancho Seco event. He does not understand why they have to wait for events to occur to identify potential generic I
problems.
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Generic items Meeting Minutes September 30, 1987
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Dr. Siess and Mr. Mi_chelson asked why the PTS issue was not subsumed into the generic issue associated with the Low Temperature Overpressure Protection (LTOP). Mr. Woods responded that they are two different.
issues. The PTS issue involves a temperature gradient across the vessel because it starts.from a hot operating condition.
LTOP issue starts from a cold shutdown condition end hence does not involve a temperature gradient across the vessel.
Dr. Siess asked what is the reture of the resolution of the PTS issue.
Mr. Woods responded that the PTS rule recuires that all affected plants do the following:
- Establish embrittlement screening limit
- Establish plans to reduce flux
- Report embrittlement rate
- Perform extensive analysis if the embrittlemer.t rate is exceeded certain level and obtain Commission approval to operate the plant at that level.
Mr. Michelson asked, if they identify a new cause for PTS in the future, how would they handle it. Mr. Frahm responded that they would handle it as a new generic issue but would identify it as part of the original PTS
- issue, i
Stating that, although the PTS issue was identified after the 1978 Rancho Seco event, it was not made as a USI until 1981, Dr. Siess asked l
why it took three years to start working on this issue. Mr. Norian i.
responded that since they were involved in analyzing the TMI-2 accident.
l they did not have time and manpnwer to work on the PTS issue.
t Generic Items Meeting Minutes September 30, 1987 Dr. Siess wondered how the work on such an important issue could be delayed for three years.
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Mr. Wylie asked whether any bulletins or information notices were issued after the identification of the PTS issue to make the licensees aware of this problem. Mr. Woods responded that be was not sure.
Dr. Siess asked whether the industry had taken any steps to minimize or prevent the PTS problem when it was being resolved by the Staff. Mr.
Woods responded that during that time the industry took a major effort to avoid the PTS problem. Almost all of tre affected plants were operating at a reduced flux rate. Further, all three PWR Owners Group l
mrt with the Staff periodically to discuss the resolution of the PTS issue. As a result of this extensive interaction, tre industry was aware of the Staff requirements to deal with the PTS issue.
By the time the FTS rule was issued he believes that most of the plants were in compliance with the requirements nf the rule.
Generic Issue 20, " Effects of Electromagnetic Pulses on Nuclear Power Plant Systems", Mr. F. Rosa Mr. Rosa said that the issue related to the Effects of Electromagnetic l
Pulse (EMP) on Nuclear Power Plant Systems was raised by Mr. Basdekas in 1976. Since a study done by ORNL concluded that EPP would not disable the safe shutdown capability of the plant, no work was done on this issue.
It was raised again by Mr. Basdekas in 1979 and the Ccmmission directed the Staff to undertake a study. Sandia National Laboratories (SNL) performed a study between 1980 ard 1983. The results of the SNL study concluded that EMP would not affect the safe shutdown capability of the plant.
Subsequently, the Staff recommended to the Commission that no further action be taken on the EMP issue and the Commission accepted that recommendation in November 1983.
Generic Items Meeting Minutes September 30, 1987
' Mr. Rosa said that the resolution.of this issue involved 1.4 Staff years and about $0.6 million-for the SNL study.
In response to a question ebout the difference between lightning strokes and the EMP, Mr. Rosa said that'the duration of an EMP is on the order I
of one nanosecond and the pulse from a lightning strike is on the order of about two milliseconds.
Mr. Ebersole asked what is the effect of en EMP on solid state devices inside the plant. Mr. Rosa responded that using the results of the SNL study, they examined this particular issue and found that the effect of' an EMP on solid state devices would be insignificant.
Dr. Siess asked whether lightning is still an issue. Mr ' Rosa responded he does not believe that lightning is still an issue. The Staff has taken the position that the protection against lightning provided by the licensees based on conservative industry standards is adeouate.
Mr. Ebersole commented that since the duration of a lightnino pulse is longer than that of an EMP, it seems obvious that lightning will cause more damage to the instrumentation than an EMP. Mr. Rosa responded that there have been instances of lightning strokes that caused spurious scrams and damage tn instrumentation. However, it is not covered within the scope of Generic Issue 20.
Mr. Michelson asked whether internally oererated EMP, such as electric arcing, is included in the scope of Generic Issue 20. Mr. Rosa said no.
Mr. Michelsnn asked whether internally generated EMP has ever been brought up as a generic issue. Mr. Rosa responded that they are aware of this issue, but it has never heen raised es e separate generic issue.
Generic Items Meeting Minutes September 30, 1987 Mr. Michelson commented that he does not believe that the SNL study looked at the effects of internally generated EMP on instrumentation and other systems. He believes that the scope of Generic Issue 20 should be made explicit.
Mr. Michelson stated that the conclusion reached by the Staff, that EMP would not disable the safe shutdown capability of a plant, is based on the characteristics of the present systems in the existing plants. The same conclusion right not apply to a new system with different charac-teristics that might be installed in a new plant. He asked how do they handle the EMP issue under such circumstance and, is there any require-ment that the reviewer should go back and check the assumptions used in the resolution of Generic Issue 207 Mr. Wylie and Dr. Siess asked whether there is any requirement that the Staff should reevaluate the resolution of all generic issues and USIs, especially those that resulted in no requirements, to determine their applicability to new plant designs.
Mr. King responded that the Severe Accident Policy Statement requires that all future plants show how they comply with all generic issues and USIs. The Staff's present position is that future plants should go back and look at all resolved generic issues and USIs to decide whether they apply to those plants. They plan to write several Regulatory Guides to provide guidance for implementing the Severe Accident Policy.
In those Guides, they plan to include guidance for dealing with resolved IISIs and generic issues.
Dr. Siess asked whether they have to go back and look at the applicabil-ity of those items that are included in the LOW and DROP categories.
Mr. King responded that the program being developed by the Electric Power Research Institute (EPRI) for future plants deals with all generic issues, incl'uding those in the LOW and DROP categories.
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l Generic Items Meeting Minutes September 30, 1987 1
USI A-46, " Seismic Qualification of Equipment in Operatino Plants" -
Mr. T. Y. Chang Mr. Chang said that the issue on Seismic Qualification of Equipment in Operating Plants was designated as a USI in December 1980. After three years of study, they concluded that the use of seismic and test experi-ence is the most reasonable and cost-effective way to address this issue.
The study performed by the Seismic Qualification Utility Group (SQUG) confirmed this conclusion. He said that the industry has participated extensively in the resolution of this issue.
Industry also has assumed the responsibility for developing procedures to implement the NRC requirements concerning this issue. This issue was resolved in February 19, 1987. The Staff requirements and the details of the resolution are contained in Generic Letter 87-02 and NUREG-1211 and NUREG-1030.
Mr. Chang discussed briefly the resources used in the resolution of this issue ( Attachment C, Page 2) and the research performed for use in the resolution (Attachment C, Page 3).
Mr. Michelson asked if a licensee made a commitment to seismically qualify a particular equipment but failed to do so, does the USI A-46 resolution require that he should go back and qualify that equipment?
The Staff responded that if a licensee made a comitment to qualify an equipment in accordance with the provisions of the IEEE Standard j
344-1971 or 1975 but failed to do so, USI A-46 resolution does not apply. Such a deficiency would probably be identified by the Staff 4
during the onsite review.
USI A-46 resolution applies to only those older plants that did not make the commitment to comply with the pro-visions of the IEEE-344 Standard.
Mr. Michelson commented that the scope of the USI A-46 does not make it clear how it will be applied.
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l Generic Items Meeting Minutes September 30, 1987 l
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. Mr. Michelson asked how they plan to handle the seismically induced
. system interactions. Mr. Chang responded that seismically induced fire, flood, or inadvertent actuation of protection systems due to earthquakes are not within the scope of USI A-46.
These. issues are being lonked at
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separately. USI A-46 covers only the physical interaction between equipment, fir. Michelson commented that in the scope of the resolution of USI-A-46 it should be made explicit that the USI A-46 resolution is limited to only physical interaction between equipment and does not include system interactions resultir.g from a seismic event. Mr. Chang responded he believes that 'the scope of USI A 46 makes this point very clear.
Mr. Michelson asked whether the USI A-46 resolution applies to new plants. Mr. Chang said no.
Stating that the USI A-46 resolution document states that during plant walkdown to determine compliance with the seisr11c qualification require-ments, effort will be made to identify interactions between equipment, Dr. Siess commented that this is not a good idea.
It seems that they are trying to mix up two different issues.
If they want to perform a seismic interactions review, it should be done separately; doing it as part of the equipment qualification walkdown is e mistake.
Mr. Michelson commented that during the ACRS review of USI A-17. "Sys-tems Interactions in Nuclear Power Plants," the Staff stated that the seismically induced systems interaction issue will be covered by USI A-46.
Since USI A-46 covers only the physical interact..on between equipment, the Subcommittee may want to find out from the Staff which USI is going to address this issue.
3:
Generic Iters Meeting Minutes September 30, 1987 Generic Issue 61, "SRV Discharge Line Break Inside Wetrell Airspace of BWR Mark I and II Containments" - Mr. A. Serki7 Mr. Serkiz said that this issue was identified by the ACPS in 1982 as a potential gereric issue. The safety concerns associated with this issue were studied by Brookhaven National Laboratory (BNL) and the results were documented in BNL/NUREG-31940, dated October 1982. Based on the comments received from the NRC Staff and the BWR Owners Group on BNL/NUREG-31940, BNL reevaluated Generic Issue 61 and the results were documented in NUREG/CR-4591, dated June 1986. The BNL reevaluation indicated that the core-melt frequency associated with different kinds of SRV discharge line breaks would be about 10 10-10 per reactor year; and the estinated radiation release would be between 0.2 and 5 person-rem per reactor year.- Based on the BNL studies, Generic Issue 61 was declared resolved in August 1986..
Mr. Serkiz discussed briefly the resnurces expended in resolving this issue (Attachment C, Page 4).
Mr. Serkiz mentioned that it took about four years to resolve Generic Issue 61. He believes that the reason for the delay is that no dedicated task manager was assigned to this issue in the beginning.
After assigning a dedicated task manacer, this issue was resolved quickly.
Stating that there are 17 generic issues being resolved by NRR, Dr.
Siess asked whether NRR has a dedicated task manager for each of those issues. Mr. Hernan responded that some of those issues have dedicated task managers and some don't. Mr. Schwink added that NRR is in the i
process of assioning a lead technical person and a lead task manager for each generic issue.
Mr. Michelson asked whether the resolutien of Generic Issue 61 applies to new Advanced Boiling Water Reactors (ABWRs). Mr. King responded that
Generic Items Meeting Minutes September 30,'1987 EPRI has been looking at the applicability of all generic issues to future plants.
If EPRI recommends that Ieneric Issue 61 be looked at' I
with regard to its applicability to new ABWRs, he expects that GE will' follow that' recommendation.
Stating that certain generic issues that had been resolved with no requirements for existing plants may become valid for future plants as a result of some new concerns, Mr. Michelson asked whether the Staff will go back and reevaluate the associated old generic issues to address the j
new concerns or' designate a new generic issue. Mr. Serkiz responded that.they probably would identify that as a new generic issue and go back and look at the list of old resolved issues.to see whether.any of them are similar to the new issue.
Mr. Michelson suggested that it would be helpful to develop a computer data bank that includes key informatico on all resolved issues. When a new generic issue is identified, the Staff should be able to go to the computer data bark and find out whether any similar issues had already been resolved, nature of resolution, and assumptions used in the resolu-tion, etc.
NRR PRESENTATION Effectiveness of the Overall Process - Mr. W. Schwink l
i Mr. Schwink stated that, in general, NRR believes that the overall process dealing with generic issues and USIs is effective. However, there are some weaknesses that NRR believes could be improved.
- There should be a global approach in dealing with generic topics, including generic issues, USIs, etc.
- All generic topics of similar nature should be consolidated and integrated.
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Generic Items Meeting Minutes September 30, 1987
- The project managers and technical reviewers in NRR should get involved in the early stages of identification, prioritization, and resolution of generic issues and USIs.
- Efforts should he made to ensure that fixes resulting from the resolution of generic issues and USIs are implemented promptly at the applicable plants.
- The tracking and record keeping of the status of all generic actions should be improved.
Althcugh the Safety Issues Management System (SIMS) is intended for this purpose, the data put in SIMS should be validated.
- Institutional procedures to deal with generic issues and other generic topics should be developed..
Dr. Siess asked whether NRR believes that the generic issue process has improved plant safety in a cost-beneficial manner.
Mr. Schwink respond-ed that, based on his discussion with various licensees, he believes that net safety at plants has improved.
Dr. Siess asked whether NRR thinks that the overall process should be speeded up to take care of all backlog issues; if all backlog issues are taken care of, will it improve the overall process and make it more effective. Mr. Schwink resporded that the NRR recommendation to research personnel is that any generic issue that is identified be resolved within two years.
Mr. King stated that everyone seems to agree that the overall process should be improved. To do so would need adeouate personnel and re-sources. He believes that unless higher level management gets involved in iirproving the process, it will be difficult to do so.
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Generic Items Meeting Minutes September 30, 1987-Dr. Siess comented that unless the Commission and the EDO decide that.
the overall process should be improved, it would be very difficult to do I
so.
Mr. Michelson commented that sometimes moving too fast to resolve a particular issue may not be a good idea. Understanding the safety concerns of an issue is very important and may take some time.
Trying to resolve an issue in haste without clearly under.ctanding the technical i
concerns associated with that issue may not be advantageous.
Mr. Michelson asked whether there are any efforts under way to computer-ize the regulatory comitments made by the applicants in the FSARs and in other licensee documents. Mr. Schwink responded that there is a project under way to try to do that. However, he does not believe that it will look at each and every FSAR and pick out the commitments made by the applicants.
i Dr. Siess asked at what stage of the process do they bring the licensees into the picture. Mr. Schwink responded that industry is made aware of what is being done by the NRC on certain generic issues through the NUREG-0933, "Prioritization of Gercric Safety Issues," document.
j Dr. Siess stated that probably it would he helpful to find cut from a licensee at the next Subcommittee rneeting whether they follow NUREG-0933 and how much it helps them understand what is being done by the NRC.
Effectiver.ess of the Interaction Between Licensees and NPR project I
Managers - Mr. W. Schwink Mr. Schwink stated that the interaction between licensees and NRR project managers in establishing schedules for implementing resolved generic issues and USIs, and reviewing licensee proposals, etc., is generally effective. However, he believes that additional emphasis
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Generic Items Peeting Minutes September 30, 1987; j
should be placed on ensuring that all items imposed are implemented promptly.
Mr. Michelson asked how they make sure that a preposed fix for a
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specific plant meets the intent of the resolution. Mr. Schwink respond-ed that, et the present time, based on the review of'the resolution, NRR i
technical revit.wers propose what needs to be done to meet the intent of the resolution.
If the licensee accepts that fix, he would go ahead and implement it.
If the licensee proposes an alternate way of taking care of the problem, he will submit a proposal and it will be reviewed by the NRR technical reviewers; based on that review, they would make a deci-
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sion.
In the future, they plan to involve the person who was involved I
in the resolution, the project manager, and the technical reviewer in j
the implementation process.. The person responsible for the resolution would work with the NRR technical reviewers and project managers until the last affected plant implemented the necessary modification.
Mr. Wylie suggested that it would be helpful if the Staff provided a flow chart identifying all the steps and personnel involved in a generic issue from the time it was identified until it got implemented.
l Steps Involved In Implementation Process - Mr. R. Hernan j
Mr. Hernan discussed briefly the outstanding generic issues that are yet to be implemented at Oconee Units 1-3 (Attachment C, Page 5). With the following two exemples, Mr. Hernan discussed the various steps involved from the time they are imposed until they are implemented (Attachment C, Pages 6-19).
- Generic Issue related to " Seismic Qualification of Auxiliery Feedwater Systems"
- USIs A-3, A-4, A-5 related to Steam Generator Tube Integrity.
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Generic Itens Meeting Minutes September 30, 1987 Stating that a sprcific fix resulting from the resolution of a generic issue sometimes cannot be applied to a particular plant and an alternate fix may have to be applied to meet the resolution requirements, Mr.
Michelson asked, ur. der such circumstances, does NRR have to discuss the alternate fix with the persons involved in the resolu' ion and get his t
concurrence. Mr. Hernan responded that there is no uniform way of handlino such situation.
The meeting was adjourned at 1:57 p.m t'OTE:
Additional reeting details can be obtained from a transcript of this meeting avaflable in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., or can be purchased from Heritage Reporting Corporation,1220 L Street, N.W.,
Washington, D.C.
20555,(202)628-4288.
l LIST OF DOCUMENTS SUBMITTED-T0 THE SUBCOMMITTEE GENERIC ITEMS MEETING SEPTEMBER 30, 1987 1.
Presentation Schedule.
2.
Memorandum from Chairman Zech to D. Ward, dated September 18, 1986.
.3.
Minutes of'the May 27, 1987 Generic Items Subcommittee meeting.
4..
Summary of the Informal meeting between Dr. Siess and representa-tives of RES and NRR held on August 7, 1987.
5.
Summary of the Informal meeting between Dr. Siess and NRR Project Managers associated with Duke Power Plants held on September 8,.
1987.
6.
Infonnation Provided by the Staff on Selected Generic Issues and USIs during the August 7,1987 Informal meeting.
7.
Information on Generic Issues 20, 61 and USIs A-46 and A-49.
i l
l ATTACHMENT A
REVISED: 9/29/87
- TENTATIVE PRESENTATION SCHEDULE -
ACRS SUBCOMMITTEE MEETING ON GENERIC ITEMS SEPTEMBER 30, 1987 R00M 1046, 1717 H ST., N.W.
WASHINGTON, D.C.
ACRS CONTACT:
Sam Duraiswamy 202-634-3267 NOTE:
- Presentation Time should not exceed 50% of the Total Time allocated for a specific item. The remainfng 50% of the time H
is reserved for Subcommittee questions and answers by the Staff.
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' Number of Copies of the Presentation Materials to be submitted to the Subcommittee:
25 copies.
TOTAL PRESENTATION ITEM PRESENTER TIME ACTUAL-TIME i
1.
EXECUTIVE SESSION 15 min 8:30 - 8:45am 2.
RES PRESENTATION RES 150 min 8:45 - 11:15am Discuss the following USIs and Generic Issues with respect to:
- Nature of resolution
- Manpower, Time, and Budget involved in the resolution
' Nature of research, if any, conducted for use in the resolution
)
- Other factors involved in the resolution u
b77AcM8Enr b
B-t i
Generic Items Schedule
- 2..
September 30, 1987 4
- Tentative Presentation Schedule.-
TOTAL-PRESENTATION ITEM
- PRESENTER TIME ACTUAL TIME'
'2.
RES PRESENTATION (Cont'd) a.
USI A-46, " Seismic RES Qualification of
's Equipment in Operat-
- 7. y, c 3,"O ing Plants" b.
US1 A-49, "Pressur-ized Thermal Shock" R. Woods c.
Generic Issue 20,
" Effects of Electro-magnetic Pulses.on I ' R * '6 a Nuclear Power Plant Systems" d.
Generic ~ Issue 61, "SRV Discharge A.Serkt)
Line Break Inside Wetwell Airspace of-BWR Mark I and 11 Containments"
- BREAK ***
15 min 11:15 - 11:30am 3.
NRR PRESENTATION a.
- Do you think that NRR 75 min 11:30 - 12:45pm i
the overall process i
y,,, 3,y,, <
(prioritization re-solution, imposi-tion, and imple-mentation)is effective? If not what could be done to improve it?
d Bz
.a Generic Items Schedule September 30, 1987
- Tentative Presentation Schedule -
TOTAL
-PRESENTATION ITEM PRESENTER TIME ACTUAL TIME-I 3.
NRR PRESENTATION (Cont'd)
- Do you think that NRR the interaction is pj, g g g g,.,g witn Licensees in establishing schedule for implementation, reviewing licensee proposal, etc. is effective?
b.
Using some typical examples of USIs and/or Generic R. Hernan Issues that re-sulted in major i
design changes, Discuss all the steps. involved from the time they are-imposed until they are imple-mented.
4 SUBCOMMITTEE REMARKS 45 min 12:45 - 1:30pm
- ADJOURN ***
1:30pm
)
i 25 -:5
FORMAL " CLOSURE" ELAPSED MONTHS
.TOOK UNTIL JULY,1985 T0 00 IT!
I
'12 DECEMBER 9, 1982, WE STARTED PROCESS, WITH BROAD STAFF SUPPORT WROTE " PROP 0' SED RULE" PACKAGE (TOOK s l' MONTH)-
REVIEWED,. CHANGED, REVIEWED... (" DISASTER ON BUS")
- CONCURRENCE BY NRC STAFF, STAFF LAWYERS CONCURRENCE SY ACRS, CRGR 19 JULY 15, 1983, PROPOSED RULE (SECY-83-288) TO COMMISSION REVIEWED BY COMMISSION STAFF AND LAWYERS 21 SEPTEMBER 5, 1983, CHAIRMAN PALLADING APPROVAL (FIRST) 23
- NOVEMBER 16, 1983, COMMISSIONER GILINSKY' APPROVAL (LAST).
25' JANUARY.13, 1984, FORMAL COMMISSION APPROVAL (N0 SIGNIFICANT CHANGES) 26 FEBRUARY 7, 1984, PUBLISHED " PROPOSED RULE" 29 MAY 7' 1984, PUBLIC COMMENT PERIOD ENDED WROTE " FINAL RULE" PACKAGE (NO SIGNIFICANT CHANGES) 38 FEBRUARY 20, 1985, FINAL RULE (SECY-85-60) TO COMMISSION 42 JUNE 20, 1985, FORMAL COMMISSION APPROVAL (h0 SIGNIFICANT CHANGES) 43*
JULY 23, 1985, FINAL RULE PUBLISHED Was 62 months including Regulatory Guide.
Would take longer now.
(Average USI time 54 months, range 45 to 81 months).
b TTACHMtHT'
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RESOURCES-USED FOR RESOLUTION i
TECHfilCAL' ASSISTANCE TIME EXPEllSE WORK
-LLNL COMPLETED 83 75K TASK 4
-IllEL COMPLETED 84 300K TASK 3
-BNL COMPLETED 84 320K TASK 1, 5
-SNL 84 - DATE 200K SUPPORT 1 895K SSRAP MEMBER SOUG COMMITTED EXTENSIVE RESOURCES FOR EXPERIENCE DATA COLLECT 10ll, SSRAP, GIP, CABLE TRAYS, TRAINING SEMlflARS
- TO DATE APPROX. 200K PER MEMBER EPRI DEVELOPED AllCH0 RAGE REVIEW GUIDELINES, RELAY EVALUAT10ll PROCEDURES, GERS, SEISMIC DEMAliD i
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G?-;L
.RESEARCH PROGRAMS SWRI - COMPARIS0N OF OLDER QUALIFICATION METHODS WITH CURRENT CRITERIA (TASK 2)
- ON-GOING RESEARCll PROGRAM
- MODIFIED TO ACCOMMODATE A fl6 NEEDS (CHANGES IH SCHEDULE. AllD TASK ORDER)
- NOT THE PREFERRED. METHOD FOR A-4G Bill /LLNL - EQUIPMENT FRAGILITY PROGRAM
- PROVIDES FRAGILITY INFORMAT10ll FOR USE IN'A-46 l
IMPLEMEtlTATION-
- 0f1-G01NG RESEARCH PROGRAM 1
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GSI-6L RESOLRCES EXPENDED
FY 86 5 75.21(
$ 174.81( Total
- 2) Prior Costs (before 1985):
Estimated to be less that 5 S01(
- 3) Concluding Phase:
Jan 85 to May 86 (Dedicated PM assigned)
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NATLRE OF' RELATED RESEARCH No special research required to resolve GSI-61.
1)
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- 2) Resolution was based on re-assessment based j
use of actual plant data, pipe break probability
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I Ocoree 1. ? and Unit 1 Unit P Unit 3 Total Active Actions 33 32 32 97 Actions Cor.pleted FY87 50 50 50 150 Active i:PA Actions 12 12 12 36 1.
RV and SV Testing (F-14) 2.
III.A.2.2. Meteorological Data Upgrade (F-68) 3.
II.K.3.31 Compliance with 10 CFR 50.46 (F-58) 4 Reactor Coolant Pump Trip (G-01) 5.
Instrumentation to follove the course of an accident (R.G. 1.97) (A-17) 6.
Item 2.1 (B-77) 7.
Item 2.2 (B-86)
E.
Items 4.2.3 and 4.2.4 (B-S9) 9, Iters 4.5.2 and 4.5.3 (B-93) 10.
RCS vents TS II.E.1 (P-83) 11.
GL 83-37 TS (B-83) 12, 10 CFR 50.62 (A-20) l l
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GENERIC ISSUE: SElSMIC QUALIFICATION OF AFW SYSTEMS (NULTI-PLANT ACTION C-14) 1 lhPOSED BY:
GENERIC LETTER 81-14 (FEBRUARY 10, 1981)-
PLANT:
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CHR0h0 LOGY:
10/80 AD HOC TASK GROUP ESTABLISHED BY NRR 2/81 G,L 81-14 ISSUED TO ALL OPERATING PWR'S 1/82 DUKE POWER RESPONSE TO G.L. 81-14 (SUPPLEMENTED 5/82) 8/82 EVALUAT10N REPORT FROM NRC CONSULTANT (LLL) 2/85 PLANT-SPECIFIC BACKFITS IDENTIFIED FOR OCONEE 3/86 BACKFIT ANALYSIS COMPLETED 1/87 SAFETY EVALUATION REPORT ISSUED TO DUKE POWER 3/87 DUKE ACKNOWLEDGED COMMITMENTS, COMPLET10N DATE - 3/90 c-6
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GENERIC --ISSUE: STAFF RECOMMENDATIONS CONCERNING STEAM GENERATORS b
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GENERIC LETTER 85-02 (APRIL 17, 1985)
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~ Q - s f $',m Y &T9/&~[ a^ ) UNITED $TATES Y, Jg 'j[1 + NUCLE AR REGULATORY COMMISSION V 1 g l)- ~ ~ "'"" &. R m * ) f FEB 101981 } r ~ / TO ALL OPERATING PRESSURIZED WATER REACTOR LICENSEES m 4,a /
SUBJECT:
SEISMIC QUALIFICATION OF AUXILIARY FEEDWATER SYSTEMS (Gene 6e Letter No. 81-14) j car letter to you dated October 21, 1980, identified concerns regarding (l' d pe seismic qualification of Auxiliary Feedwater (AFW) systems. That (i Wter outlined the continuing evaluation program being conducted by the staff with regard to this issue to assure conformance of each plant s with General Design Criteria 2 and 34, of Appendix A to Part 50. A'; a result of the NRC's continuing review of this issue, including the f.spetito of site vi0its described in our previous letter, we have } cMirnined that it is riecenary to request certain information from PWR licensees and to request that certain actions be performed by PWR g licensees, as described below. The purpose of our information request 4 9 it'to obtain sufficient information that identifies the extent to which We are also requesting that PWR /f! systems are seismically qualified..icensees perform a walk-down cf the non-seism >h t 1/ X of their AFW systems to identify apparent and practically correctable deVcycies that may exist. s,, Fg,r pMts with AFW systems that are net seismically qualified either in whuir c-in part, our plan involves increasing the seismic resistance of the t{estems in a timely, systematic manner to ultimately provide reah,nable assurance, where necessary, that they are able to function folgdown Eatthouake (SSE) for the plant. ira the occurrence of earthquakes up This plan is a result of a S @. - J study of the seismic requirements which should be applied to AFW systems for those not presently seismically qualified, as discussed in NUREG-b 0667, " Transient Response of Babcock and Wilcox-Designed Rchetors." j 7 / k Encicere 1 to this letter contains a request for information from all opehting PWRs concerning AFW system seismic design. We have determined tha t docketed information from licensees is not sufficient to allow us p to conduct a detailed review of this aspect of AFW systems. In addition, 'b @r a number of older plants, this information is likely not to be '/ 4' current. Furthermore, since the safety significance of the system may k,h not have been defined fnr all plants, the AFW system may not have been + adequately maintained and considered to be included within the scope @A. N> \\, - of IE Bulletins 79-02., 79 04, 79-07, 79-14, and 80-11, and IE Information 6 \\ Notice 80-21 Therefore, the existing AFW systems, either in total or in part.) may have as-built characteristics which result in uncertain -{ spismic cesign characteristics. For plants with AFW systems, or portions N I thereof, which are not seismically qualified, Enclosure 1 also requests ir.formatim concerning systems which provide an alternate decay heat ,y / f ( ( ,e removal path. 0 I t' / Dh 7c0 C [ @ h
L y .n a, ,v f + s p_ p w, ' FEB 101981 o w We are also requesting that you conduct a walk-down by personnel experienced in the analysis, design and evaluation of such structures, systems and - c'omponents, of. the non-seismically qualified portions of the AFW system for the purpose of ider.tifyinij-more readily recognized deficiencies in seismic resistance. These walk-downs are requested for only those portions of the AFWesystem vhich have not been designed, constructed, . and maintained as seismically qualified systems in accordance with the criteria for safety grade systems at the facility. The scope of the L, walk-down should jnclude the types of equipment, components, and piping described in Enclosure 2. Enclosur& 1 describes what we consider to comprise the bounds of the AFV systsm, and any alternate decay heat . removal paths g y For plants with AFW systems tnat tire # not seismically qualified, we consi-der that actions should be taken'soon ^o ensure a reasonable level of earthquake resistance. This applies to both the AFW system and the alternate system used for decay heat removal if portions of it are not seismically qualf fie'd. Based upon the consideration.of the tast perfor-n- 1, 'mance of nuclear and fossil poeer plants, and ctherao6-nuclear facilities ~ subject to large earthquakes, we note that well engineered structures, ~ equipment, components and piping possess a substantial amount of inherent i i i t ca, even without the rigorous seismic qualification > O se sm c res s an performed for safety-grade portions of nuclear f&cilities. Of the failures of structures, piping, equipnent and cumponenfs noted in these past earthquakes, a large fraction have been due to brittle failure, lack of restraint, large displacements, or some other obvious deficiency which would have been easily identified before the failure caused by the seismic event. Such identified deficiencies could have been corrected to significantly enhance reliability without detailed seismic analyses but by eWrcising careful engineering judgement. These considerations were factored into the development' of Enclosure 2. In addition, certain of these deficiencies were noted as existing at the several facilities for which we conducted AFW system walk-downs (see for details of the visits), Accordingly, your walk-down of the non-seismically qualified portions of the AFW system and other alternate decay heat removal systems should identify any appropriate modifications in the context of the above discussion. identifies in nere detail the actions we consider appropriate for pi:rts with AFW systems, or portions i.heree" that Are not seismically Although we are not at this time revesting that the AFW qualified. systen, be rooMied to be in conformance with the facility design seismic requireants, we have stated that our plan is to increase the seismic resistance, where necessary, to ultimately provide reasonable assurance that the sysbm wiD function after the occurrence of earthquakes up to and including tni 95F. 1 i e-9 )
M 10 2 ,3, Accordingly, the following actions are requested by this letter: 1. In accordance with 10 CFR 50.54(f) of the Comission's regulations, all PWR licensees are requested to provide the information contained ' in Enclosure I within 120 days of receipt of this letter; and 2. The results of any walk-downs are requested within 120 days of receipt of this letter. These results should include all identi-fied deficiencies and all corrective actions taken, or planned along with the schedules for such. Such modifications, if any, I shall be handled in the customary manner consistent with the provisions of your license and the Comission's regulations. Responses should be submitted to enable us to determine whether or not your. license should be modified, s spended, or revoked. a re G.gisenhut, Director Division of Licensing
Enclosures:
As stated "This request for information was approved by GAO under a blanket clearance number R0072 which expires November 30,1983. Comments on burden and duplication may be directed to the U.S. General Accounting Office, Regulatory Reports Review, Room 5106, 441 G Street, N.W., Washington, D. C. 20548." G e e C-lO
REQUEST FOR INFORMATION AUXILIARY FEEDWATER SEISMIC DESIGN 4 In responding to this letter, the AFW system boundary from suction to discharge (including the water source and heat sink) shall include those portions of the system required to accomplish the AFW system function and connected branch piping up to and including the second valve which is nomally closed or capable of automatic closure when the safety function is required. The AFW system boundary shall also include any portion of branch piping that .is structurally coupled to the AFW system boundary such that the seismic response of the branch piping transmits loads to the AFW system. As a minimum, this includes the branch lines outside the AFW system boundary to a point of three orthogonal restraints. All mechanical and electrical equip-ment, piping (e.g., instrument air), conduits and cable trays, which are necessary or contain items which are necessary, for the operation of the AFV system shall also oe considered. In addition, the structures housing these systems and components shall be included. Similar considerations shall be applied when considering alternate means of decay heat removal. Specify whether your AFW system is (a) designed, constructed, A. and maintained (and included within the scope of seismic related Bulletins 79-02, 79-04, 79-07, 79-14, and 80-11, and IE Informa-tion Notice 80-21), in accordance with Seismic Category I require-C ments (e.g., conformance to Regulatory Guides 1.29 and the applicable portions of the Standard Review Plan or comparable criteria) or (b) designed, constructed and maintained (and included within the scope of seismic related Bulletins 79-02, 79-04, 79-07, 79-14, and 80-11, and IE Infomation Notice 80-21) to withstand a Safe Shutdown Earthquake (SSE) utilizing the analytical, testing, evaluation methods and acceptable criteria consistent with other safety-grade systems in your plant. To assist the staff in an expedious assessment of your plant, if your AFW systenor portions thereof, is not qualified to with-stand an SSE utilizing the analytical, testing and evaluation criteria consistent with other safety-grade systems in your plant, we request that you identify those components and structures not seismically qualified in the appropriate row of the attached Table 1. Where seismic qualification is indicated by leaving Table 1 blank, B. provide a description of the methodologies and acceptance criteria used to support your conclusion of seismic qualification, including: Seismic analyses methods emnloyed, seismic input, load combinations which include the SSE, allowable stresses, qualification testing and engineering evaluations perfortned. + In addition, where seismic qualification of a secondary water supply or path is relied upon, provide a summary of the proce- } dures which would be followed to enable you to switch from the ( primary to secondary source. ( c-l)
I l . s C. If a' lack of seismic qualification is indicated for items' 1, 2, 3, 4, 5 and 6, 7, or B in Table 1, provide additional information which specifies the level of seismic qualification afforded in the original design for each of these areas. D. If substantial lack of seismic qualification is indicated for items 1, 2, 3, 4, 5 and 6, 7, or 8'in Table 1, provide the same information requested in A through C for any alternate decay heat removal system. The bounds of these systems shall be considered to a similar extent as that described for the AFW system. Provide a sumnary of.the procedures by which operation of these alternate heat removal systems will be a ccompli s hed. 0 .e e a t 4:-125
.L TABLE 1 4 AUXILIARY FEEDWATER SEISMIC QUALIFICATION (1) Pumps / Motors - (2) Piping (3) Valves /Actyators (4) Power Supplies (5). Primary Water and Supply Path (6) Secondary Water and Supply Path * (7) Initiation and Control System (8) Structures Supporting or Housing l these AFW System items l 0
- Applicable only to those plants where the primary water supply or path is not provided, however, a seismically qualified alternate path exists.
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4' g. .s , ENCLOSURE 2 ACTIONS REQUESTED OF PRESSURIZED WATER REACTOR LICENSEES WITHOUT A SEISMICALLY OUALIFIED 1 AUXILIARY FEEDWATER SYSTEM 1. For all mechanical and electrical equipment and components including battery racks, controls, instruments, motors, compressors, tanks electrical supplies and the cabinets containing such items, note all items which are not securely attached to their supporting structures such that there is substantial resistance to movement caused by seismically induced forces. 2. For piping, note cases where relatively large deflections cannot be accommodated without impairing system function. Where such displacements will lead to loss of system function, note where sufficient amounts of restraint could be afforded,' thus signifi-cantly reducing stresses that would be imparted to such items as pump nozzles and branch connections, as appropriate.
- Further, note eccentric valve operators that are not sufficiently supported and act to severly overload the pipe. Where such support is lacking, you should note where substantial restraints could be added to the extent practical.
Also, where pipes are resting j on existing supports, note where substantial resistance to the pipe moving off these supports could be added where it does not already exist. 3. For cable trays and conduits, assure that relatively large displacements can be accommodated without i@ airing system function where seismic restraint is substantially less than required for these which are seismically qualified. Focus particular attention on preventing the breakage of the electrical and control cables they contain at such places as points of attachment of the cables to equipment or other relatively fixed points. Note where any deficiencies exist. Given the time frame we are recommending for the completion of these actions, no explicit analyses are requested to demonstrate system qualifi-4 cation unless deemed necessary by you. However, sound engineering judgement should be applied considering the level'of seismicity specified for your site and the design requirements for other seismically qualified systems in the facility when judging the necessity for and adequacy of any modifications (e.g., piping, cable trays, conduit, equipment and component restraints,and estimations of displacement levels).
- Further, these actions shall be accomplished using personnel who are experienced l
in the analysis, design and evaluation of such structures, systems and components. 1 l 1 d-lk
Where you have determined that it is prudent to institute modifications, no modifications should be instituted which will detrimentally affect the function of the piping, equipment and components of the system, considering all 6ther loads in addition to seismic. For example, when - providing additional restraint to piping systens, assure that they do not have a detrimental impact on the system considering all loads, in addition to seismic, including thermal loads and support displacement induced loads. Similar considerations as described above should be given to other non-seismically qualified piping, equipment and components in the vicinity of the non-seismically qualified portions of the AFW and the alternate decay heat removal systems to provide for a substantial decrease in their susceptibility to failure if such failure c "'d impact the function of the AFW and alternate decay heat remot 0 systems. 0 l 0 1 ~ C-Ib w--___--_____
4 3 PLANT: OCONEE 1, 2, 3 CHRONOLOGY: 1978 IDENTIFIED AS A USI 4/85 GENERIC LETTER 85-02 ISSUED TO ALL PWR'S 7/85 DUKE POWER RESPONSE TO G.L. 85-02 IMPROVEMENTS TO BE ACCOMPLISHED: PRIMARY - SECONDARY LEAK RATE TECHNICAL SPECIFICATION LIMIT TO BE RED' CED FRGM 10 GPM' TO 1.0 GPM J e l~ 4 ti-lb L
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UNITED $TATES /..._'. ;, NUCLE AR REGULATORY COMMISSION 1.- . j WASHINGTON D. C.20$55 ( yf April 17, 198.S. N..v p TO ALL PWR LICENSEES OF OPERATING REACTORS, APPLICANTS FOR OPERATING ~ LICENSES, AND HOLDERS OF CONSTRUCTION PERMITS, AND FT. ST. VRAIN 8 Gentlemen:
SUBJECT:
STAFF RECOMMENDED ACTIONS STEMMING FROM NRC INTEGRATED PROGRAM FOR THE RESOLUTION OF UNRESOLVED SAFETY ISSUES REGARDING STEAM GENEPATOR TUBE INTEGRITY (GENERIC LETTER 85-02) The Corr'ission has recently' approved issuance of this generic letter to all nuclear power plants utilizing steam generators, to obtain infomation on their-overall progran for steam generator tube integrity and steam generator tube rupture mitioation. This information will allow the staff to assess the areas of concern addressed by the staff's recommended actions (see Enclosure 1) which were developed as 'part o' the integrated program for the resolution of Unresolved-Safety Issues A-3, A-4 and A-S regarding steam generator tube integrity. 'The staff's prroram repnrt, NUREG-0844 (draft report for comment)..is provided.as'. NUREG-0844 will be issued in final form following a 90-day period for public comment. (- Stean generator tube integrity was designated an unresolved safety issue (USI) in 1978 and Task Action Plans (TAP) A-3 A 4 and A-5 were established to evaluate the safety significance of degradation in Westinghouse, Combustion Engineering and Babcock & Wilcox steam generators, respectively. These studies were later combined into one effort due to the similarity of many problems among the PWR vendnrs. Steff concerns relative to steam generator tube degradation stem from the fact that the steam generator tubes are a part of the reactor coolant system (RCS) bounAry and that tube ruptures allow primary coolant into the secondary system where its isolation from the environment is not fully ensured.. The leakage of primary coolant into the secondary system has two potential safety implications which were considered. The first is the direct release of radioactive fission prnducts to the environment; and the second is the loss of primary coolant water which is needed to prevent core damage. An extended, uncontrolled loss of coolant outside of containment could result in the depletion of the initial RCS water inventory and ECCS water without the capability to recirculate.the water. .o An integrated procran was initiated by the staff in May 1982 to consider initial recommendations from the US1 effort, and to assess the lessons which could be , f' learned from the four domestic SGTR events; Point Beach 1 in 1975; Surry ? in 1976; Prairie Island 1 in 1979; and Ginna in 1982. A number of potential require-ments for industry were identified and subjected to a value impact evaluation. ( a,wan - -~ v r.u r v.,_arL,( 3(( c r1 rr r - //I
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- -d 3e e' b -These analyses indicate that: the proba' bility of core melt from events involving
. steam aenerator tube ruptures is not a major contributor to total core melt ris'k; tbst steam generator tube ruptures are an important contributor to-the probability of.significant non-core nelt releases; and that steam generator. tube degradation <is a ma.ior contributor to occupational radiation exposure at PWP's. Based'upon the results of the staff's intecrated program, the staff has developed reconnended actions in the following areas: ~1. Prevention and Detection of Loose Parts and Foreign Objects 2. Steen Generator Tube Inservice Inspection 3. Secondary Water CFenistry Program 4. Condenser Inservice Inspection Program 5. .Prir.ary to Secondary Leakage Limit 6. Coolant Iodice Activity Limit 7.. Safety injection Signal Reset The staff's recomended actions have been found to be effective' measures on a plant specific basis for significantly reducing (1) the incidence of tube degradation. (2) the frequency of tube ruptures and the corresponding potential for significant non-core melt releases, and (3) occupational exposures, and are consistent with good operating and engineering practices. Accordingly, operating reactor licensees and applicants for an operating license (this letter is for information only for those utilities that have not applied for an operating license) are requested to. furnish to the Director, Division of Licensing, Office of Nuclear Reactor Regulation, no later than 60 days from the date of this letter, a description of their overall programs for assuring steam generator tube integrity and for steam generator tube rupture mitigation. The description of the plant specific programs should be sufficiently detailed to allow the staff to compare these actions with the staff recommended actions as presented in Enclosure 1. -The staff recommended actions above do not addrets supplemental tube sample inspections 'or the case where Categnry C-2 results are obtained during initial sample inspections. The sta'f initially considered a proposed upgrading of existing Technical Specification requirements in this area (see Section ?.9.1 of the enclosed draft NUREG-0844), and this proposal was commented upon extensively.by. industry. The staff has concluded that the particular proposal considered was not warranted as e neneric staff position or reenmendation. However, as part nf the information reauested by tFis letter, licensees and applicants are requested tn describe practices thev employ to ensure adequate I inspection samples are teken in tbt rvent that Cateoory C-2 results are obtained during initial sampling. The information requested is described in additional detail in Enclnsure 2. C-18
, g,. The staff will review each response from licensees and applicants, and evaluate the overall effectiveness of plant-specific programs to prevent and mitFgate the occurrence of steam generator tube ruptures. The staff recognizes, however, that plants specific programs may differ from the staff recommended actions, and still be adequately effective. The results of the staff review will be reported directly to the Commission. The Commission has specifically requested that the staff include in its report the number and quality of the responses, noting in particular-any utilities delinquent in providing the requested information and any recommended corrective actions. The staff will continue to monitor licensees' commitments and programs relative to steam generator integrity and SGTP m'tigation to determine if they are being effectively implemented. As has been true in the past, additional actions may. become necessary in plant specific instances of extensive or severe degradation. This request fnr infortnation was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985. Comments on burden and duplication may be directed to the Office of Management and Budget. Report Manaaenent Room 3208, New Executive Office Building, Washington, D. C. 20503. Mr. Em-ett Murphy, Operatina Reactors Assessment Branch, will be the point V contact. If you have ouestions or desire additional information, he can be I reached on (301.) 492-7457. I Sincerely, /1?? ugh L. Thompson, Jr. Dire 'r Div s on of Licensin Office of Nuclear Reactor Regulation
Enclosures:
1. Staff Reconrended Actions Stemming from NRC Integrated Program for the Resolution of Unresolved Safety issues Regarding Stean Generator Tube Integrity P. Request for information 4 Concerning Category C-2 Steam Generator Tube Inspections 3. NUPEG-0844 (Draft Report For Comnent), NRC Integrated Program for the Resolution ( of Unresolved Safety issues 4 List of Generic Letters e-I9 _.}}