ML20151P938
| ML20151P938 | |
| Person / Time | |
|---|---|
| Issue date: | 06/24/1988 |
| From: | Zech L NRC COMMISSION (OCM) |
| To: | Sharp P HOUSE OF REP., ENERGY & COMMERCE |
| References | |
| CCS, NUDOCS 8808100224 | |
| Download: ML20151P938 (62) | |
Text
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NUCLEAR REGULATORY COMMISSION l
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June 24, 1988 CHAIRMAN The Honorable Philip R. Sharp, Chairman Subcomittee on Energy and Power Comittee on Energy and Comerce United States House of Representatives Washington, D.C.
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Dear Mr. Chaiman:
I am responding to your letter of May 19, 1988, concerning the status of NRC actions on improvements to Mark I containments, the Babcock and Wilcox reassessment, and the pipe wall thinning issue. Our detailed responses to your specific questions are enclosed. Your letter also expressed reservations about the Nuclear Regulatory Comission's ability to resolve important safety problems in a manner that will enhance the public's confidence in the NRC.
Because your perceptions of our perfomance raise the fundamental issue of the role of the regulator in protecting the public health and safety, I would like to clarify the Comission's approach to safety issues such as the three that are the subject of your letter.
The NRC perfoms its regulatory function in two ways.
First, it promulgates and reviews regulations and requirements on the basis of new infomation as it becomes available. Before a decision is made to impose new requirements, however, the Commission must ensure that the action proposed will enhance rather than detract from the safe perfomance of our licensed facilities.
While we agree that unwarranted delays in implementing safety requirements will reduce public confidence, we also believe unnecessary actions, counter-productive actions, or actions which fail to achieve the expected results also create the perception of inefficient or ineffective regulation. The issues associated with improvements in Mark I containments are an example of this process.
In view of the wide differences of opinion within the technical comunity regarding the best approach to achieving real improvement, the NRC, as noted in the enclosure, has broadened its examination of this issue in the context of its approach to severe accidents. The staff will brief the Comission on the prograss of this examination in a public meeting presently scheduled for July 22, 1988.
Secuad, the NRC maintains close oversight and, when appropriate, takes tough enforcement action to ensure compliance with existing regulations and requirements. We also strongly encourage the industry to aggressively pursue self-improvements in safety perfomance where such measures are desirable and are likely to enhance public safety. Both the B&W reassessuent and the pipe wall thinning issue are examples of areas where effective licensee actions should result in desirable improvements. NRC oversight will assure that industry self-improvement activities are effective and timely.
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2 NRC encoerages changes to facilities that enhance safety and uses the fortnal process established by the backfit rule to better manage and control imposition of new requirements to ensure the objective of safety enhancement is achieved.
Therefore, the backfit rule provides stability to the regulatory process in establishing a reasonable method for controlling and accounting for necessary regulatory activity.
I want to assure you that the Commission intends to pursue safety concerns and enhanced protection of the public by whatever means is appropriate and effective.
Although we insist that individual licensees are ultimately responsible for the safe operations of their plants and encourage industry self-improsements in their safety performance, we will not delegate the responsibilities assigned to us by the Congress.
In carrying out these responsibilities, we intend to be the tough, but fair regulators the American people expect and deserve.
Sincerely, W-A.
Lando W. Zech Jr.
Enclosures:
As Stated cc: Rep. Carlos J. Moorhead
O QUESTION 1(A).
Please explain the five-element improvement program for MARK I's proposed by the staff in 1986.
ANSWER.
A five-element program proposed in 1986 was intended to enhance the performance of the BWR MARK I containment in response to severe accident challenges. The elements of the program were as follows:
1)
Hydrogen Control - to reduce the likelihood of hydrogen combustion challenges by reducing the allowable time that an inerted containment is permitted in a de-inerted mode.
2)
Containment Drywell Spray - to provide for backup water sources and pumping capability that are not dependent on AC power; 3)
Containment Ver.
o provide a vent path through the suppression pool for the removal of decoy heat and non-gaseous radioactivity during accidents including station blackout; i
4)
Core Debris Control - to provide cooling O' molten core debris to delay or prevent liner failure, and to retain water in the torus room to help quench molten debris and to scrub and remove particulate radioactivity; i
and
O QUESTION 1(A).
(Continued) 5)
Emergency Procedures and Training - to incorporate the latest version (Revision 4) of the BWR emergency procedures guidelines into emergency operating procedures and training.
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O QUESTION 1(B).
Why did the NRC choose not to order implementation of this i
program?
ANSWER.
After the initial proposal had been drafted, the staff held two separate meetings in February and March 1987, respectively, with key scientists representing NRC contractors, and with those representing the industry.
In these meetings many questions relating to the MARK I issue were reviewed in order to assess containment challenges and failure modes as well as to establish the efficacy and justification for the different proposed improvement strategies. The range of technical opinions expressed in these meetings was very wide. The industry spokesmen indicated a fairly substantial confidence in MARK I perfonnance, and especially that molten core debris would not melt through the containment drywell shell.
There was a wide range of views regarding the effica' various improvements expressed by NRC contractor-supported scientistt 1, with some viewing drywell liner melt-through as quite likely.
In view of the lack of technical consensus on fundamental questions of accident phenomenology and contairment response, the staff had no assurance that the five-element program would represent a reasonable set of improvements without additional study.
In July 1987, the staff informed the Commission that it intended to examine the MARK I issue in the context of an integrated approach to the closure of severe accident issues. This integrated approach includes design and procedural improvements to further reduce the likelihood of severe accidents 7--,,
QUESTION 1(B).
(Continued) as well as to improve the capability to mitigate the consequences of these low probability accidents.
As indicated in response to Question 3, below, the staff's study of the MARK I containment issue is in its final stages, with a status report expected in July 1988 and with final recomendations er.pected by the fall of 1988. A copy of the staff's final recontendations will be provided to the Subcomittee staff upon request, when the recontendations become available.
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QUESTION 2 (A).
What is the Commission's best estimate of the likelihood of failure of Mark I containments during a severe accident?
ANSWER.
The likelihood of early containment failure of the MARK I containment under severe accident conditions is a matter of considerable controversy, and there is no best estimate at this time. A group of containment experts provided a pre-liminary assessment of the Peach Bottom containment, which was published in Draft NUREG-1150 in February 1987.
Because of the uncertainties in accident phenomenology and progression as well as MARK I containment performance, their estimates of containment failure produced no consensus but showed a large variation ranging from almost zero to almost ninety per cent probability of failure. At a February 1988 workshop at which representatives of the staff, NRC research contractors, industry representatives, and the public participated, there was a range of views as to the likelihood of containment failure; however, there was general agreement that water in the drywell is useful in preventing / delaying liner failure and especially in reducing fission product release.
l It should also be noted that failure of a MARK I containment would not necessarily lead to a large radioactivity release due to the significant fission product retention expected by scrubbing in the suppression pool, plateout on cooler surfaces within the reactor coolant system and containment, and settling and deposition within the secondary reactor building surrounding the containment,
0 QUESTION 2(A).
(Continued) l Rihk analyses performed for BWR MARK I plants have generally shown that the likelihood of occurrence of severe accident conditions (i.e., severe core damage and core melting) can realistically be 3xpected to be quite low since there are many ways in which water can be provided to maintain or restore adequate core cooling. Overall safety of each plant is well served by placing considerable emphasis on the prevention of severe accidents to preclude the kind of challenges to containment discussed above.
It is for this reason that the staff is developing a balanced approach to the resolution of the MARK 1 concerns that places considerable errphasis on prevention as well as mitigation of severe accidents, i
QUESTION 2(B).
Please explain the liner melt-through issue. How muen does this issue contribute to the probability of containment failure?
ANSWER.
During postulated severe accidents, molten core debris may heat the reactor vessel bottom head until it fails, and the molten core debris is then released onto the drywell floor.
Molten debris could then flow and spread to the drywell steel shell. As the debris reaches the shell, it heats the steel.
If the local shell temperature were to rise sufficiently high, the shell would be likely to melt locally.
There is a consensus of opinion among experts that if the core debris at the shell is less than about 2 inches deep, the shell will not fail. However, if the temperature of the core debris is several hundred degrees higher than its liquefaction temperature and the debris is more than approximately 8 inches deep, there is a consensus of opinion that the shell will melt through.
There is a large degret et uncertainty in the actual depth of the core debris reaching the liner as the depth is dominated by the following factors: 1) the amount of the core that melts; 2) the location of the failure of the reactor vessel; 3) the rate of flow of core debris from the vessel; 4) the temperature and composition of the core debris; 5) the type of concrete used in the drywell; and 6) the presence of. water in the drywell.
There is no consensus among experts on these issues, and this results in a range of views as to the survivability of the steel shell.
EESTION2(B).
(Continued) 2-1 Draft NUREG-1150 has investigated the importance of shell melt-through to early containment failure. A wide range of early containment failure estimates from almost zero to close to ninety percent resulted when shell melt-through was assumed to occur. When shell melt-through was assumed not to occur, the upper range of early containment failure estimates was reduced to about forty percent. This suggests that shell melt-through could contribute significantly to the probability of early containment failure for MARK I containments.
There is a consensus that an additional source of water in the containment drywell could significantly reduce the amount of radioactivity released to the environment, and might delay or reduce the likelihood of liner melt-through as well.
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QUESTION 3(A).
What is the NRC's schedule for final resolution for the PARK I containment issue?
ANSWER.
NRC's study of the MARK I containment issue is in its final stages. A status report by the staff to the Consnission is expected to be completed in July 1988. A final report with recomendations is expected by the fall of 1988.
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OVESTION 3(B)._
When will the NRC staff make recomendations to the 4
Comission?
ANSWER.
As noted in response to Question 3(a), the NRC staff expects to transmit a final report with recomendations to the Comission in the fall of 1988.
The report will discuss the preventive a:!d mitigative effectiveness of proposed improvements, and will make recomendations.
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QUESTION 3(C). When will the Comission act on these recomendations?
ANSWER.
The Comission intends to act on these recomendations in a timely fashion when it concludes that the staff recomendations are complete, consistent.
technically sound, and will result in valid improvements in MARK I safety.
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QUESTION 3(D).
How will any new requirements be implemented? By rule? By order? By other means?
ANSWER.
No decision has been made at this time regarding the form the requirements may take.
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l QUESTI0fi3(E).
When would you expect that actual changes to hardware or procedures would be implemented by the utilities?
ANSWER.
The schedule for implementing improvements would be highly dependent on the nature of the changes.
Some changes could be implemented almost immediately, while others are likely to be completed within 2 years after approval of recuirements by the Comission.
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OUESTION 4(A).
What role will the backfit rule play in deciding what new requirements to impose?
ANSWER.
All potential improvements and requirements will be evaluated by the staff in accordance with 10 CFR 50.109, the backfit rule, as required by the Commission's regulations. However, the Commission recognizes that there are unique situations where the staff may identify an improvement that could significantly reduce the risk of a severe accident, but which, in the staff's judgment, may not satisfy all the tests of the backfit rule. The staff is being directed to bring these issues to the attention of the Commission.
The resolution of these issues then would rest with the Conr.ission, which has the ultimate authority and responsibility to make regulatory policy decisions.
QUESTION 4(B).
What is the basis for Mr. Speis' statement that improvements to containment performance may not be able to meet the cost-benefit requirements of backfit rule?
ANSWER.
The statement that containment performance improvements needed to meet potential new requirements might not be justified by using the cost-benefit requirements of the backfit rule was based on a judgment of past regulatory analysis results and was also considered applicable to ome, but not all, of the proposed irprovements.
Cost-benefit analyses have estimated the benefit of a proposed change by calculating the difference in risk (before and after the proposed change) expressed in units of population dose, or person-rem.
The benefit is then monetized by usino a value of 1000 dollars per person-rem, and can be compared to the erst to see if it is justified.
Past analyses have shown that only thos equirements that balanced dose reductions with implementation costs c,
. be justified.
Changes were likely not justified where the proposed change resulted in small dose reductions, or where the initial risk value was very low to begin with. However, as noted in response to Question 4(a), the staff will identify significant improvements and bring these forward to the Commission, even if they do not meet the cost-benefit requirements of the backfit rule.
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QUESTION 4(C). Does the Comission agree with that statement?
Please explain in detail.
ANSWER.
All potential containment improvements and requirements will be evaluated by 1
the staff in accordance with 10 CFR 50.109, the backfit rule.
There is always the possibility that proposed improvements to nuclear power plants may not meet the standard for imposition under the backfit rule.
The detemination for containment performance improvements cannot be made until the NRC staff completes its study of this issue. However, as noted in the response to Question 4(A), the Comission is directing that the staff bring to its attention those unique situations where the staff may identify an improvement that could significantly reduce the risk of a severe accident, but which, in the staff's judgement, may not satisfy all the tests of the backfit rule. The resolution of these issues then would rest with the Comission, which has the ultimate authority and responsibility to make regulatory policy decisions, t
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QUESTION 4(D).
Does this apply to the five-element program proposed by the staff in 19867 il ANSWER.
A detailed cost-benefit analysis of the five-element program proposed by the staff in 1986 was not perforr.ed. This is being incorporated into the analysis supporting the final report to the Comission in the fall of 1988.
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QUESTION (5). What is the Commission's schedule for deciding whether plants with other containment types are in the need of changes to improve containment performance?
ANSWER.
Issues related to ice condenser, MARK II, and MARK III plants will be discussed with industry in a workshop scheduled for February 1989. A resolution of the issues on these containments is expected to be submitted in a report to the Commission, with staff recommendations, in August 1989.
Issues related to dry containments will be discussed with industry in a workshop tentatively scheduled for April 1989. A resolution of the issues on these containments is expected to be submitted in a report to the Commission, with staff recommendations, in August 1989.
For each of the above phases, the Commission will be requested to approve implementation of the staff recommendations.
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QUESTION f(A).
Please provide the subcommittee with whctever infomation yo have on how foreign countries have approached the issue (ainment performance.
ANSWER.
Foreign countries hue taken a wide range of approaches with regard to the issue of containment performance.
Several European countries have decided to require the installation of filtered vents on containments.
Many other countries are following U. S. efforts with some interest, but have issued no new requirements in this regard.
For the Subcommittee's infomation, Enclosure I to this question provides a draft report prepared by several members of the NRC research staff on the subject of filtered vented containments. This report discusses the bases for selection and implementation of filtered vented desions in France, Sweden, and the Federal Republic of Germany. The report also discusses U. S. reser
-fforts, areas of incomplete infonnation, and regulatory issues as Obviously, it is important that filtered vented containment designs not be mandated in the U.S. until a sound technical basis has been developed which assures that public safety will not be reduced under any of the various severe accident sequences which have been hypothesized.
The International Atomic Energy Agency (IAEA) is sponsoring a conference to be held in October 1988 in Moscow entitled "Technical Committee Meeting on Severe Accident Containment Design Bases." This conference is expected to discuss efforts underway in member nations of the IAEA with regard to assessment of
RESTION6(A).
(Continued) 2-containment loads and capabilities for severe accidents beyond the design basis and to examine possible improvements. The U. S. expects to have a strong presence at this conference to present detailed infonnation on U. S. research on prevention and mitigation of severe accidents as well as to learn more about efforts in foreign countries.
Enclosure:
As stated
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e QUESTION 6(B).
Some European countries have decided to require filtered vents on all of their nuclear power plants. Why has the NRC decided not to require filtered vents?
ANSWER.
A final decision has not been reached concerning the need for and overall benefits versus risks of filtered vents. The NRC has been actively participating in international conferences on this subject end is currently working to resolve this issue.
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QUESTION 7.
The owners of two PARK I plants - Vennont Yankee and Pilgrim-expressed an intention to implement the five element program proposed by the staff in 1986. What is the status of implemertation of these two plants?
ANSWER.
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Vermont Yankee Nuclear Power Corporation (Vennont Yankee), by letter dated June 25, 1987, apprised the NRC of its plans to voluntarily investigate two major containment safety initiatives -- improved drywell spray /RHR injection capability and containment venting. More recently, in a letter dated March 1, 1988, Vermont Yankee informed the NRC that, as a result of their study of and ongoing research on severe accident issues, they plan to install additional spray system capability in their 1989 refueling outage. However, referring to uncertainties over the associated benefits and risks, Vermont Yankee has concluded that no further work on venting is currently warranted.
Boston Edison, by letter dated July 8,1987, submitted the description for a comprehensive voluntary program to address safety improvements at the Pilgrim plant. This effort, identified as the Pilgrim Safety Enhancement Program (SEP), encompasses all elements of the staff's fonner "five element program" except provisions for torus room debris coolability. Additionally, the Pilgrim SEP has addressed plant specific safety improvements that extend beyond the five element program.
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QUESTION 7.
(Continued) The staff has completed an initial review of the Pilgrim plant modifications and generally finds them to be acceptable.
It is anticipated that all the safety improvements will be installed prior to restart from the current outage with the exception of the direct torus vent system (DTVS), which is installed but not currently operational. The staff is not able at this time to provide a schedule for completion of the DTVS operahility review because it is awaiting the licensee's response to a series of questions regarding operation of the system.
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QUESTION 8.
Why did the NRC decide to let the B&W Owners Group conduct the l
design review of B&W reactors rather than the NRC7 Please provide the subcommittee with all documents relevant to this decision.
ANSWER.
Since utilities are respensible for the safety of their plants and can benefit the most from such in-depth review, it was appropriate for the B&W Owners Group (B&WOG), in conjunction with the B&W utilities, to perform this review with oversight from the NRC staff. These groups had the resources, knowledge, and familiarity to provide the needed broad-based, in-depth review. The NRC staff resources were appropriately focused on review and oversight of the B&WOG efforts.
In the summary and conclusions section of the staff's Safety Esaluation Report (NUREG-1231, page 3-2, Enclosure 1 to this cuestion), the staff stated that the detailed review that was obtained was a direct result of each utility's familiarity with its own plants and equipment performance and that the B&WOG lead role, in conjunction with the oversight provided by the staff, was appropriate and made this effort succeed. to this question contains a copy of a memorandum from V. Stello, Acting EDO, a the Conmissioners and related correspondence which provided the basis for this decision.
Enclosures:
As stated
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O' ESTION 9.
Why has the NRC not fomally required that the BWOG recomen-J dations be implemented at B&W plants? Please provide all documents relevant to this decision.
ANSWER.
The B&WOG "Safety and Performance Improvement Program" (SPIP) is primarily based on the evaluation of complex transients and reactor trips at B&W plants between 1980 to 1986.
The predominant cause of these events was due to failures in balance of plant (BOP) systems and components.
Since BOP equipment is not cenerally cor.sidered to be "safety related", the stati did not usually scrutiniza this equipment in detail during the course of the review for initial licensing.
In addition there are currently few regulatory requirements specifically directed toward BOP systems.
However, the NRC is in the process of assessing the need for expansion of its regulatory policy v -h regard to these systems. Equally important, the utilities are in fat
-tarily undertaking the recomended changes. The staff intends te audit their implementation of the significant recomendations (those that will primarily enhance safety rather than operational perfomance) to ensure they are properly accomplished. As part of these audits, the staff also intends to ensure that these recommendations do not introduce undesirable interactions with safety-related equioment.
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QUESTION 9,.
(Continued) l The utilities have committed through the B&WOG to implement all the recommendations that are applicable to their plants or provide justification for rejecting any recommendations. As can be seen in the enclosed Table on page 12-2 of Supplement I to the Safety Evaluation Report (NUREG-1231), the utilities are in the process of implementing the recommendations approved by the B&WOG. Also note that only about 1% of these recommendations have been rejected by the utilities.
In short, NRC is closely following these voluntary actions by B&W utili-ties and has thus far found no reason to change its approach relative to this issue.
Enclosure:
As Stated i
QUESTION 10.
Why has the NRC not formally required that schedules for the implementation be submitted te the NRC for approval?
ANSWER.
The staff has not formally required through regulatory actions that schedules for implementation be submitted because these recommendations, as previously mentioned, are primarily related to balance of plant systems.
The staff requested in its Safety Evaluation Report (SER) (page 12-5) that the utilities provide a schedule for implementation of the recommendations. This request was somewhat modified in the SER Supplement (page 12-2) based on the B&WOG comments provided on the staff's SER (page 12-1 of the SSER).
Based on discussions with the B&WOG, this information will be provided in July 1988.
We believe this is a reasonable time frame for providing schedular information to NRC.
If the information received is not adequate, the staff will consider additionift steps to a'c that the NRC obtains this desired information.
The utilities have cou....ed through the B&WOG to implement all the recommendations that are applicable to their plants or provide justification for rejecting any recommendations. As can be seen in the Table on page 12-2 of Supplement I to the SER (NUREG-1231 which is enclosed), the utilities are in the process of implementing the recommendations approved by the B&WOG. Also note that only about 1% of these recommendations have been rejected by the utilities.
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OVESTION 10.
(Continued) In short, NRC is closely following these voluntary actions by B&W utilities and has thus far found no reason to change its approach relative to this issue.
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i QUESTION 11(A).
When does the NRC expect that all of the BWOG reconnendations will be implemented at all plants?
ANSWER.
As previously mentioned, the staff is awaiting schedule infonnation from the B&WOG, which is expected in July 1988. The B&WOG schedule will be provided to the Subcommittee staff upon request, once it is received by the NRC.
QUESTION 11(B).
What is the NRC doing to ensure that they are implemented in a timely manner?
ANSWER.
The NRC will evaluate the acceptability of the implementation dates for the selected recommendations.
It has been proposed to the B&WOG that the individual utilities provide an integrated schedule for all licensing actions, including the BAWOG recommendations, for NRC approval.
Once this schedule has been agreed upon, the utilities would be responsible to notify the staff of schedule slippages.
This integrated schedule would only include the recommendations selected by the staff.
Implementation of the recommendations will also be monitored by the NRC.
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QUESTION 11(C).
If a utility decides not to implement the recommendations or only to implement some of them, what action would the NRC take?
ANSWER.
The action the NRC would take would depend on which recommendations the utilities decided not to implement. However, it is our opinion that all the significant applicable recommendations, based on the B&WOG and the utilities commitments to this program, will be implemented.
Should they elect not to implement some, the NRC would consider requiring implementation in accordance with the provisions of 10 CFR 50.109.
QUESTION 12(A).
What role did the NRC's backfit rule play in the Comission's decisions on whether to require changes to plants and procedures?
ANSWER.
The NRC's backfit rule did not play any role in the decision to initiate a B&W plant reassessment program. As mentioned in the January 24, 1986 letter from V. Stello, acting EDO, to H. Tucker, Chainnan of the B&WOG, this program was initiated because of a safety concern resulting from a number of events that occurred at B&W-designed reactorc that reinforced our concerns regarding these designs. Since the program that resulted was a voluntary program initiated by the B&WOG and the recomendations are being implemented by the utilities on a voluntary basis, the backfit rule does not apply.
The staff has monitored the program and performed independent analyses and audits to determine if any additional actions were required. The staff would have evaluated any additional actions under the backfit rule if additional actions were detemined necessary; however, none were identified.
QUESTION 12(B).
Did the NRC conduct a backfit review or analysis, either fomally or infomally? Please provide whatever informa-tion you have relevant to this.
ANSWER.
As discussed in the response to question 12a, there was no need for a backfit review since this was a voluntary program, l
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i QUESTION 12(C).
If the B&W owners do not make changes to their plants as recomended by the B&WOG, will the B&W plants meet the adequate protection standard over the long term?
ANSWER.
Yes, the B&W plants meet the NRC requirements with regard to overall safety.
The recomendations resulting from this program are considered to be safety enhancements.
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QUESTION 12(E).
If the Comission were to decide to make implementation of the B&WOG recomendations a new regulatory requirement, would they meet the requirements of the backfit rule?
ANSWER.
The NRC has not analyzed the recomendations to determine whether they would meet the backfit rule.
However, the staff believes that an overall enhancement in safety in the long tem will result from the implementation o# the most significant recommendations.
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,e QUESTION 13.
The B&WOG recomendations have been criticized as being aimed more at improving economic efficiency of plant operations than at improving safety.
Does the Comission agree? P M se explain.
ANSWER.
The Comission participation in this program is aimed at plant safety; however, the utilities efforts may well be based on both safety and economic improvements.
This concern was brought up by the Advisory Comittee on Reactor Safeguards (ACPS) in a memorandum dated July 16, 1986, following an early program briefing. The NRC staff responded to the ACRS regarding this concern in a memorandum dated August 14, 1986.
(Enclosure 1 to this question.)
In a subsequent meeting, the ACRS concurred that the program, which had been expanded since the ACRS memorandum, would result in safety benefits.
In addition the staff has had almost continuous interaction with the B&WOG during the development of the program and throughout the program. The B&WOG program is directed towards the reduction of reactor trips and complex transients that have a direct impact on reactor safety.
The B&WOG program relied heavily on past operational experience to deterinine which systems and components caused the occurrence of these events, and the B&WOG recomendations are consistent with the findings of the program.
Enclosure:
As Stated
QUESTION 14(A).
In NUREG-1231, the NRC staff stated that it had not yet completed review of certain issues such as the integrated control systems. See for example pp. 3-1, 10-1, 11-4, and j
l 12-5.
Has that review now been completed? Please provide j
l the relevant documentation.
ANSWER.
i The staff has completed its review of the B&WOG plant reassessment program, and
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the results are contained in Supplement I to the Safety Evaluation Report (NUREG-1231), which was issued in March 1988. As previously mentioned, a copy of this report has been provided as an enclosure.
OVESTION 14(B).
Since the staff had not yet reviewed certain issues, how was the staff able to conclude that continued operation of the plants was acceptable?
ANSWER.
The determination that BAW-designed plants were safe for continued operation was discussed by the staff in the staff's response to a Union of Concerned Scientists petition dated October 19, 1987 (enclosed). The SER (NUREG-1231) and its supplement documents the staff's basis for this determination.
One of the objectives of the B&WOG reassessnent program was to determine whether B&W plants have adequate safety margins for the longer term.
Enclosure:
As stated
QUESTION 15.
Has the NRC taken any action to resolve the pipe thinning issue beyond that described in the March 1988 GA0 Report? Please explain.
ANSWER.
Our response to the GA0 Report was issued on May 31, 1988 and is enclosed.
As indicated in our response, the GAO Report does not find fault with the technical or schedular resolution of the NRC actions. However, the GA0 Report recomends that we fonnalize the actions as regulatory requirements. We believe that the current course of action has resulted in the expeditious implementation of plant specific programs to resolve the pipe wall thinning issue. Thus, it has not been necessary for the NRC to issue further specific requirements. The industry program results will be closely monitored by NRC.
If the program is found not to be effective, the NRC will promptly issue requirements covering needed industry actions.
The GA0 Report accur reflects staff actions through January 1988.
- However, they are scattered throughout the report. The attached chronology provides a concise listing of staff milestone actions, including those taken subsequent to January 1988 or planned, to achieve closecut of this issue.
These actions demonstrate that we have not deferred to industry the resolution of the pipe wall thinning issue.
Instead, we have taken a positive course of action which is resulting in the rapid implementation of programs to address pipe wall thinning at all plants.
j QUESTION 15.
(Continued) In summary, NRC initially responded to the Surry event on the day it occurred.
Within 3 months, the cause of the problem had been determined, all licensees had been informed of the findings, and preliminary inspections and repairs were conducted at many plants. Within six months of the event, a pipe wall thinning program, including inspection procedures, replacement criteria, and long tem monitoring criteria, was developed by NUMARC/EPRI and found acceptable by the staff. This program, or an equivalent one, is currently being implemented at all plants. As of June 1, 1988, 81 cut of 113 plants have completed inspections and repairs, as necessary. The remaining inspections are being completed en the schedule discussed in our response to the GA0 report.
Additionally, the May 2, 1988 Subcommittee Staff memorandum attached to your May 19, 1988 letter incorrectly notes in paragraph No. 4, on page 5, that the discovery of pipe wall thinning at the Trojan Plant "was particularly I
significant because it found thinning of safety-related piping in the primary system." To date, there has been no reported pipe wall thinning occurring in l
the primary system of operating nuclear power plants. The staff stated in the NRC Infomation Notice No. 87-36 issued on August 4,1987, that the safety-
)
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related portion of the feedwater system (secondary system) at the Trojan Plant has experienced pipe wall thinning. Although this piping is safety-related, it is not part of the reactor coolant system pressure boundary (primary system).
Enclosure:
As stated
\\
i LIST OF STAFF MILESTONE ACTIONS FOR RESOLUTION OF PIPE WALL THINNING ISSUE Date Action Dec. 9, 1986-Jan. 14, 1987 Special, announced augmented inspection was conducted for the Surry Unit 2 feedwater pipe rupture event of Dec. 9,1986.
Dec. 16, 1986 IE Information Notice IN 86-106, "Feedwater Line Break," was issued to alert the industry to the Surry Unit 2 feedwater pipe rupture event.
Jan. 15, 1987 Panel discussion with industry experts on the generic implications of the Surry Unit 2 feedwater pipe rupture.
Feb. 1986 Telephone survey of 91 plants regarding their erosion / corrosion monitoring program.
Feb. 13, 1987 IE Information Notice IN 86-106 Supplement 1 was issued to provide updated information on the cause and effect of the pipe rupture, i
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Date Action l
Feb. 17, 1987 Letter to the Subcommittee on Nuclear Inservice Inspection (S/C XI) of the ASME Boiler and Pressure Vessel Committee and the ASME Council on Codes and Standards requesting these groups to address the issue of pipe wall thinning.
l Feb. 25, 1987 Commission briefing on the Surry pipe rupture event.
Mar. 18, 1987 IE Information Notice IN 86-106 Supplement 2 was issued providing additional information on the analysis of the pipe rupture event.
Apr. 7, 1987 Meeting with NUMARC to discuss industry efforts to address the Surry 2 event. Provide verbal staff comments.
i May 5, 1987 Transmittal of formal staff comments to NUMARC regarding industry proposed plan to address pipe wall thinning.
May 28, 1987 Meeting with NUMARC to discuss proposed guidelines for addressing wall thinning in single phase piping.
June 2, 1987 Receipt of NUMARC's proposed progran to address pipe wall thinning.
Date
_ Action June 12, 1987 Transmittal of staff acceptance to NUMARC regarding its program for single-phase pipe inspections and provided connents.
July 9, 1987 NRC Bulletin 87-01, "Pipe Wall Thinning in Nuclear Power Plants," was issued to request industry provide infomation on its pipe wall thinning history and inspection programs for single-phase and two-phase lines.
This coincided with completion of inspections at the spring refueling outage for many plants.
July 22-23, 1987 The Task Force on Pipe Wall Thinning made a site visit to the Trojan plant.
Aug. 4, 1987 NRC Infomation Notice, "Significant Unexpected Erosion of Feedwater Lines, was issued to alert industry about pipe wall thinning in the safety-related portions of the feedwater lines at the Trojan plant.
Aug. 13, 1987 The Task Force on Pipe Wall Thinning issued its evaluation of the Trojen event.
I
Date Action Aug. 20, 1987 Letter to the ASME Section XI Comittee requesting development of requirements and procedures to predict, detect and prevent pipe wall thinning due to erosion / corrosion in nuclear power plants.
Dec. 10, 1987 Meeting with NUMARC to discuss its status of implementation of its program on pipe wall thinning in single-phase lines.
Staff raised concern that not all utilities are comitted to its program.
Dec. 15, 1987 Letter to NUMARC regarding staff's concern abcut the implementation of its single-phase program at all utilities.
March 17, 1988 Meeting with NUMARC to review implementation status of single-phase inspection programs.
All but 1 plant had comitted to implement the program in the near future. This plant subsequently conducted an inspection.
April 22, 1988 NRC Infomation Notice No. 88-17 was issued, sumary of responses to NRC Bulletin 87-01 Thinning of Pipe Walls in Nuclear Power Plants.
l l
Date Action April-October 1988 Visits to 10 selected plants to review programs and results for single-and two-phase inspections.
Dec. 1988 Closecut, either industry actions are adequate or NRC will establish new requirernents.
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.i QUESTION 16.
What is the Commission's schedule for resolving this issue?
ANSWER.
As indicated in the response to Question 15, the NRC is scheduled to close out this issue by December 1988.
Additionally, the May 2, 1988 Subconnittee Staff Memorandum attached to your May 19, 1988 letter suggests, in part, that the NRC may be reluctant to impose new requirements for pipe wall thinning because of the chilling effect of the backfit rule end associated cost-benefit evaluation.
That is not the case.
If a determination is made by the NRC through the scheduled plant inspections that licensees have not implemented the available programs to resolve the pipe wall thinning issue, appropriate new regulatory requirements will be imposed pursuant to 10 CFR 50.109.
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SUBCOMMITTEE ON ENERGY AND POWER
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"'JlyiL"'!?if0 Watjington BC 20515 May 19, 1988 The Honorable Lando W.
Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Chairman:
As we have discussed previously, I believe it is critical that the NRC be a tough regulator that is not perceived in any way to be an advocate of or deferential to the nuclear industry.
I am deeply concerned about the NRC's actions--or inaction--on several safety matters detailed in the enclosed subcommitteo staff memorandum.
The memorandum highlights three cases of NRC failure to take regulatory action in matters identified by Commission staff as significant safety problems.
In each case, the Commission has either studied the matter to death or allowed utilities to decide for themselves the appropriate course of action.
In cne case, the Commission denied two utilities the right to make voluntary safety improvements.
The Commission was informed in November of 1986 that GE Mark I containments at 24 nuclear power plants across the nation would be as likely to fail as succeed in containing radioactive emissions in the case of a severe accident.
Yet the Commission has taken no action, and instead has prevented two utilities from taking voluntary action.
While the Commission continues to i
I study this matter, other countries have already decided to improve the safety of their containmento, i
1 In another case, as long ago as 1985 your staff identified safety problems at nuclear reactors designed by Babcock fr Wilcox.
Not only did the Commission defer to the owners of i
these plants to determine what safety lu.provements were required, but when that owners group recommended actions, the NRC left it to each utility to decide whether and, when to comply.
j i
In the thitd case, despite reports of the dangers of pipe thinning and 34 plants with identified problems, the Commission has adopted no regulations and has added no new inspection 1
requirements.
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The public will continue to lose trust in the nation's j
nuclear industry if it believes that the NRC is too timid in cracking down on safety problems. It is hard to imagine engendering public confidence in nuclear safety when such i
significant decisions are left to the industry itself.
The Commission can hardly be seen as a tough regulator if it simply i
refuses to regulate'.
Until the Commission can establish a clear regulatory agenda with a focus on safety, the rising criticism of the NRC will continue.
I am enclosing a number of questions on these thrse safety areas.
I would appreciate your answers by June 20, 1988.
Your response should also include a clear timetable for resolving each of these matters.
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QUESTIONS Mark I containments:
RES coordinate w/NRR (1) (a)
Please explain the five-element improvement program for Mark I's proposed by the staff in 1986.
(b)
Why did the NRC choose not to order implementation of this program?
RES (2) (a)
What is the Commission's best estimate of the likelihood of failure of Mark I containments during a severe accident?
RES (b)
Please explain the liner melt-through issue.
How much does this issue contribute to-the probability of containment failure?
RES (3) (a)
What is the NRC's schedule for final resolution of the Mark I containment issue?
RES (b)
When will the NRC staff make recommendations to the Commission?
RES/COMM (c)
When will the Ccmmission act on those recommendations?
RES (d)
How will any new requirements be implemented?
by rule?
by order? by other means?
RES coordinate (e)
When would you expect that actual chat:ges to w/NRR hardware or procedures would be implemented by the utilities?
RES (4)(a)
What role will the backfit rule play in deciding what new requirements to impose?
RES (b)
What is the basis for Mr. Speis' statement that improvements to containment performance may not be able to meet the cost-benefit requirements of backfit rule?
RES/C0th (c)
Does the Commission agree with that statement?
Please explain in detail.
RES coordinate (d)
Does this apply to the five-element program w/NRR proposed by the staff in 1986?
RES (5)
What is the Commission's schedule for deciding whether plants with other containment types are in need of changes to improve containment performance?
RES coordinate (6) (a)
Please provide the subcommittee with w/ NRR & GPA whatever information you have on how foreign countries have approached the issue of containment performance.
R{Scoordinate (b)
Soms Europgan countrias hava dscidad to e/ NRR & GPA require filtersd vanto on all of their nuclear power plants.
Why has the NRC decided not to require filtered vents?
NRR coordinate (7)
The owners of two Mark I plants - Vermont e/RES Yankee and Pilgrim - expressed an intention to implement the five element program proposed by the staff in 1986.
What is the status of implementation at those two plants?
B&W Reactors:
NRR (1)
Why did the NRC decide to let the B&W Owners Group conduct the design review of B&W reactors rather than the NRC?
Please provide the subcommittee with all documents relevant to this decision.
NRR (2)
Why has the NRC not formally required that the BWOG recommendations be implemented at B&W plants?
Please provide all documents relevant to this decision.
NRR (3)
Why has the NRC not formally required that schedules for the implementation be submitted to the NRC for approval?
NRR (4) (a)
When does the NRC expect that all of the BWOG recommendations will be implemented at all plants?
NRR (b)
What is the NRC doing to ensure that they are implemented in a timely manner?
NqR (c)
If a utility decides not to implement the recommendations or only to implement some of them, what action would the NRC take?
NRR coordinate (5) (a)
What role did the NRC's backfit rule play w/RES & OGC in the Commission's decisions on whether to require changes to plants and procedures?
NRR (b)
Did the NRC conduct a backfit review or analysis, either formally or informally?
Please provide whatever information you have relevant to this.
NRR (c)
If the B&W owners dc not make changes to their plants as recommended by the BWOG, will the B&W plants meet the adequate protectien standard over the long term?
NRR coordinate (e)
If the Commission were to decide to make w/0GC & RES implementation of the BWOG recommendations a new regulatory requirement, would they meet.he requirements of the
backfit rule?
NRR/C0mi (6)
The BWOG recommendations have been criticized as being aimed more at improving economic efficiency of plant operations than at improving safety.
Does the commission agree?
Please explain.
NRR coordinate (7) (a)
In NUREG-1231, the NRC staff stated that it w/RES had not yet completed review of certain issues such as the (A-47) integrated control systems.
See for example pp.
3-1, 10-1, 11-4, and 12-5.
Has that review now been completed?
Please provide the relevant documentation.
NRR (b)
Since staff had not yet reviewed certain issues, how was the staff able to conclude that continued operation of the plants was acceptable?
Pipe Thinning NRR (1)
Has the NRC taken any action to resolve the pipe thinning issue beyond that described in the March 1988 GAO report Please explain.
fMR (2)
What is the Commission's schedule for resolving this issue?
o May 2, 1988 MEMORANDUM TO:
Hon. Phil Sharp, Chairman FROM:
Subcommittee Staff
SUBJECT:
Unresolved Safety Issues:
Delay and Deference at the NRC Summary Several recent actions by the NRC on nafety issues raise questions about the adequacy of the NRC's regulation of nuclear power plants and seem to illustrate two tendencies in agency decision making which may be cause for concern.
The first is the NRC's seeming inability to resolve safety issues in a timely manner.
The agency seems to study issues without ever reaching any sort of conclusion about whether procedural or hardware changes are necessary at nuclear power plants.
For example, a recent GAO report pointed out that some unresolved safety issues have been pending for as long as 10 years.
Nuclear safety issues are complex, but other countries are far ahead of the U.S. in resolving some safety problems that we in the U.S.
are still studying.
The second concern is the Commission's pattern of deferring to the nuclear industry's opinions on safety issues.
There arc several instances in which the Commission has chosen not to impose safety requirements but rather simply to endorse voluntary industry programs -
training for power plant personnel, fitness for duty, maintenance (although this may be changing), and access authorization, to name a few.
While cooperative ventures between the NRC and industry are valuable in addressing safety issues, they are not substitutes for independent regulatory action.
Something which may be playing a role in the NRC's decisions not to issue new requirements is the NRC's backfit rule.
The backfit rule is the NRC's self-imposed limitation on what changes the Commission can require to licensed nuclear power plants.
Basically, the rule states 1
1
O that the NRC may only require changes to procedures or hardware after conducting a cost-benefit evaluation of the proposal and finding that as a result of the change there will be a substantial increase in the overall protection of the public health and safety.
It is difficult to be certain because the Commission's decisions not to impose new requirements do not always include a discussion of the requirements of the backfit rule, but it is possible that the rule may be having a chilling effect on the NRC staff's willingness to propose or require safety improvements.
A recent memo from an NRC staffer on the containment performance issue discussed below indicates that this may be the case.
There are several safety issues which illustrate the patterns we mentioned above.
These are the Mark I containment issue, the design review of Babcock & Wilcox (B&W) reactors, and the thinning of pipes in nuclear power plants.
Mark I containments An example of delay in the resolution of a safety issue, which has long been before the Commission, is the question of whether to require improvements to plants with General Electric Mark I containments.
Following several years of study by the NRC and its contractors, the NRC staff informed the Commission in November of 1986 that there is a 50-50 chance that a Mark I containment will fail during a severe accident.
T.f the containment fails, there is nothing to prevent the escape of radioactive materials to the environment.
The staff recommended that the Commission order immediate hardware and procedural changes to the Mark I plants in an attempt to reduce the likelihood of failure.
The staff originally concluded that the five-element program proposed by the staff was justifiable under the backfit rule.
The changes were opposed by industry, and the Commission did not adopt the recommendations.
The NRC staff met with the Commission again in July of 1987 and withdrew their recommendation.
In tabling this issue, the NRC precluded two utilities - Public Service of Vermont and Boston Edison - from voluntarily making safety improvements to the Vermont Yankee and pilgrim plants.
Those two companies expressed an intention to implement the Commission staff's five-element program.
The NRC has so
l e
far refused permission to make certain changes pending further study.
Whether to impose new requirements to improve the performance of Mark I containments is now being considered as a part of NRC's overall research on severe accidents.
After additional study, the NRC staff has again concluded that the Mark I containment has a high likelihood of failure, possibly as high as 90%, and that the containment could fail in only a few minutes.
("BWR Task Group Report," February 10, 1988.)
Yet, the NRC still has not decided whether to require changes to Mark I's, and if so, what to require.
In addition, recent press reports suggest that the deputy director of the NRC's research office has concluded that changes to plants to improve containment performance may be precluded by the Commission's backfit rule.
(Inside NRC, March 14, 1988, p.
13.)
This may contradict the staff's original statement to the Commission in 1986 that improvements could meet the requirements of the backfit rule.
In contrast to the NRC's inaction, the Europeans have already begun to implement improvements to containment performance.
For example, the Swedes, the West Germans, and the French have all decided to install filtered vents on their containments, even on their large dry containments which are generally thought to be less likely to fail in the case of a severe accident than are Mark I's.
The NRC has concluded that the containments on 24 nuclear powc-
' ants have only a 50-50 chance of actually performing function during a severe accident.
These plants are :
at the end of this memo.
This appears to be a signifn safety problem which should be resolved expeditiously.
However, even if the Commission does decide that improvements are necessary, they may be barred by the backfit rule.
B & W Reactors The NRC's handling of the design review of Babcock &
Wilcox (B & W) reactors is an example of the Commission deferring to industry judgment in addressing safety issues.
In 1986, the NRC staff became concerned about the design of B & W reactors.
Serious operating events at two of the reactors - Rancho Seco in California and Davis-Besse in Ohio - prompted the staff to decide that a reexamination of the "basfc design requirements" of all B&W reactors was
4 e
necessary.
(Letter from Victor Stello, Jr., EDO, to Hal Tucker, Chairman BWOG, January 24, 1986.)
Rather than initiating a staff review, however, the NRC decided to allow the owners of the B & W plants to conduct the design review for their own reactors.
Questions about the adequacy of this approach were raised by the Advisory Committee on Reactor Safeguards (ACRS), on the ground that the B & W Owner's Group (BWOG) study seemed intended more to improve the economic performance of the plants than to improve safety.
(Letter from David A. Ward, Chairman ACRS, to Victor Stello, Jr., EDO, July 16, 1986.)
The BWOG completed its work in late 1987 and recommended improvements to the plants.
The NRC staff concluded upon its review of part of the BWOG study that B&W plants are safe enough to continue operating in the short term, but recommended "aggressive implementation of applicable recommendations by the utilities, so that acceptable safety margins in the long term will be assured."
(NUREG-1231, "Safety Evaluation Report Related to Babcock & Wilcox Owners Group Plant Reassessment Program," November 1987, p. 3-3.)
Despite the NRC staff's obvious support for implementation of the BWOG recommendations to ensure the long term safety of these plants, the NRC has decided not to require utilities to implement the recommendations.
Instead, the NRC has embarked on a voluntary program, requesting that each plant owner submit an implementation plan and recommending that the utility plans be submitted by June 1, 1988.
(NUREG-1231, p. 3-3.)
The NRC has imposed no new requirements.
It is not clear what role the backfit rule played in the commission's decision not make the BWOG recommendations mandatory.
A list of B&W reactors is attached to this memorandum.
l Pipe Thinning l
A third area of concern is the Commission's handling of newly discovered thinning of both safety-related and non-safety-related piping in nuclear power plants.
The Commission has not yet decided whether to require that the utilitiea adopt inspection programs designed to discover thinning before a pipe break occurs.
The thinning of pipes in nuclear power plants can have significant safety implications.
If thinned pipes burst
4 Plants Reporting Evidence of Erosion / Corrosion of Pipes
- PLANT STATE San Onofre Unit 1 California Haddam Neck Connecticut oyster Creek New Jersey Dresden Unit 2 Illinois H.B. Robinson Unit 2 South Carolina Pilgrim Unit 1 Massachusetts Surry Unit i Virginia Turkey Point Unit 3 Florida Surry Unit 2 Virginia Fort Calhoun Nebraska Fort St. Vrain Colorado Duane Arnold Iowa Arkansas Unit 1 Arkansas Rancho Seco California Calvert Cliffs Unit 1 Maryland Milestone Unit 2 Connecticut Trojan Oregon Calvert Cliffs Unit 2 Maryland Salem Unie. 1 New Jersey D.C.
Cook Unit 2 Michigan North Anna Unit i Virginia Arkansas Unit 2 Arkansas North Anna Unit 2 Virginia Sequoyah Unit 1 Tennessee Salem Unit 2 New Jersey Sequoyah Unit 2 Tennessee San Onofro
't 2 California San Onofr.
3 California Diablo Ca:
nit 1 California callaway Missouri Diablo Canyca Unit 2 California River Bend Unit 1 Louisiana Perry Ohio Shearon Harris North Carolina
- From "Nuclear Regulation:
Action Needed to Ensure that Utilities Monitor Repair Pipe Damage,"
GAO/RCED 73, March 1988, p.33.
o Babcock & Wilcox (B & W) Reactors Arkansas-1 Arkansas Crystal River-3 Florida Davis-Besse Ohio Oconee 1,2,3 South Carolina Rancho Seco California TMI-1 Pennsylvania Under construction:
Bellefonte 1 & 2 Alabama WNP-1 Washington 8
j e
GE Mark I Reactors Browns Ferry 1,2 & 3 Alabama Brunswick 1 & 2 North Carolina Cooper Nebraska Dresden 2 & 3 Illinois Duane Arnold Iowa Fermi 2 Michigan Fitzpatrick New York Hatch 1 & 2 Georgia Hope Creek 1 New Jersey Millstone 1 Connecticut Monticello Minnesota Nine Mile Point 1 New York oyster Creek New Jersey Peach Bottom 2 &3 Pennsylvania Pilgrim Massachusetts Quad Cities 1 & 2 Illinois Vernent Yankee Vermont l
e o
submit reports on the history of their pipe inspection programs.
According to a GAO report published in March of 1988, the NRC identified 34 plants with erosion / corrosion damage.
These plants are listed at the end of this memorandum.
The industry trade and research organizations have developed criteria and a computer program which can be used by utilities in developing inspection programs.
There is no requirement that they do so, however.
GAO concluded that the NRC needs a mechanism to "ensure that utilities periodically assess the integrity of pipe systems in their plants to reduce the risk of future injury to plant personnel or damage to equipment caused by erosion / corrosion."
GAO recommended that NRC require licensees to inspect plants, replace pipe which doesn't meet industry standards, and periodically monitor pipe systems.
The NRC has not yet decided whether to impose such requirements.
The issue of pipe thinning is directly related to the concerns discussed at the subcommittee's hearing on the aging of nuclear power plants.
Inspection of equipment important to safe operation in order to discover the effects of aging is an important element of an effective preventative maintenance program.
The NRC Chairman testified at that hearing that not all utilities voluntarily do as much as they should in adequately maintaining and inspecting their plants.
The NRC's response to the discoveries at Trojan and Surry, however, has been primarily to gather information and to monitor industry programs.
While the NRC may yet decide that pipe inspection programs should be required, the subcommittee should at a minimum keep an eye on what the Commission does and on how the backfit rule affects Commission decisionmaking.
Conclusion In recent years, the NRC has failed to adequately i
address safety issues.
The Commission has deferred to voluntary industry programs rather than exercising its own regulatory authority.
Important safety issues remain unresolved for years.
Some safety actions appear to have been shelved because of the NRC's self-imposed backfit i
rule.
As a result, significant safety issues at many of the nation's nuclear power plants may be left unresolved.
o
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while the plant is operating, the plant can be placed in an unsafe condition.
Coolant for the reactor may be affected if safety-related piping breaks, and water and steam from non-safety-related pipes may adversely affect important plant safety systems.
The thinning of pipes became an important new safety consideration in December 1986 when a pipe ruptured at the Surry plant in Virginia.
Four plant workers died as a result of the break.
The pipe which ruptured was not safety-related, and the cause was erosion / corrosion of the metal from which the pipe was made.
The rupture of this particular pipe was considered to be significant because until the Surry accident the industry and the NRC had not expected erosion to occur in pipes like that which burst.
4 originally, pipe thinning was only expected in pipes with two-phase flow (both steam and water).
The Surry accident demonstrated that thinning could occur in pipes with single-phase flow (only steam or water).
However, the problem was thought to be limited to areas of pipes where there were bonds or obstructions.
The industry changed its inspection criteria to include pipe sections like that involved in the Surry event.
In 1987, however, it became apparent that thinning was not necessarily limited to areas of pipes with bends or obstructions or to non-safety-related piping.
The owners of the Trojan plant in Oregon discovered significant thinning of straight sections of safety-related piping.
In some cases, the pipe had eroded to thicknesses below industry standards and had to be replaced.
Once again, this was unexpected, and inspection programs were not designed to discover this type of erosion damage.
This discovery was particularly significant because it found thinning of safety-related piping in the primary system.
On August 4, 1987, the NRC issued a notice to inform the industry of the events at Trojan.
The notice requested that utilities review the information and consider taking action, if appropriate, at their facilities.
The NRC imposed no new requirements and did not even require licensees to respond to the notice.
The NRC staff did conduct an informal survey of 91 plants during February of 1987 and concluded that the information they had gathered raised questions about the adequacy of pipe inspection programs.
The NRC issued a bulletin on July 9, 1987 which required all licensees to 1
--