ML17193B178
| ML17193B178 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/09/1980 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17193A968 | List: |
| References | |
| TASK-CA, TASK-SE SECY-80-325, NUDOCS 8103180932 | |
| Download: ML17193B178 (97) | |
Text
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July 9, 1980
{)_ELD FlLE COP1 SECY-80-325 COftliMISS~Of\\JER ACTION
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For:
From:
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Subject:
Purpose:
Discussion:
Contact:
Hank George, NRR 49-27136
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The Corr:iiissioners Harold R. Denton, Director
~l L Office of Nuclear Reactor Regulation.;
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. Iv Executive Director for Operations l
SPECIAL REPORT TO CONGRESS IDENTIFYING NEW UNRESOLVED SAFETY ISSUES To obtain Commission approval of new Unresolved Safety Issues for reporting to the Congress in a special report in accor-dance with a com~itment made in the 1979 NRC Annual Report.
The 1979 NRC Annual Report includes a progress report on Unresolved Safety Issues as required by Section 210 of the Energy Reorganization Act of 1974 as amended.
This progress report does not identify any new Unresolved Safety Issues, however, because an in-depth review consid-ering all candidate issues was not possible in late 1979 when the Annual Report was prepared.
Instead, the Annual Report includes a ccrranitment to provide a special report to the Congress i.n July 1980 identifying new Unresolved Safety Iss~es following a systematic review of all candi-date issues from the Three Mile lsland investigations and other sources.
A systematic review of candidate issues has been performed by NRR. This Commission Paper provides the results of that review and requests the Commission's approval of the six issues listed below for designation as Unresolved Safety Issues and reporting to the Congress. These issues are described in Enclosure 1.
The process for selecting the issues is described in. It is based on the definition of an
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The Colllilissioners Unresolved Safety Issue and the explanatory statements 1/
in NUREG-0510,- regarding what constitutes an issue that "poses important questions regarding the adequacy of existing safety requirements." The issues subjected to the selection process came from three principal sources--
the TMI Action Plan, ACRS letters and Reports since January 1979 and the NRC staff. Many are derived from operating experience. Listings of all of the issues considered are provided in Enclosure 3.
The initial step in the selection process was a screening of issues based on a set of initial screening criteria that considered attributes of an Unresolved Safety Issue other than the level of safety significance, e.g., an Unresolved Safety Issue must apply to a number of plants; it must relate to nuclear power plant safety, etc. Forty-four candidate issues were identifiec:J-.as a result of the initial screening.
The results of the initial screening are provided in Enclosure 3.
Each of the forty-four candidate issues was then subjected to a systematic review to judge its safety importance on the basis of whether the issue involved e_qui pment, opera-tions or emergency response and whether it was a potentially significant safety deficiency or a potentially significant safety improvement.
The judgment as to which issues should be designated and reported as Unresolved Safety Issues.was of necessity based principally on qualitative information provided as answers to these questions.
The answers were' developed through discussions with staff experts in the areas being considered.
More quantitative information would have been preferable as an aid in judging many of the issues. This was not possible, however, because of the limited time available to consider the large number of issues and the unavailability of key personnel from the Probabilistic Analysis Staff. However, an individual from the Probabilistic Analysis Staff did review and provide comments-on the information developed and the judgments made; these cormients were incorporated into the NRR.evaluation.
The development of more quantitative information to aid in such decisions by NRR in the future is intended as the new Reliability and Risk Assessment Branch in NRR develops its capabilities in these areas.
As a result of the application of the process described 17 NUREG-0510, "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants, 11 January 1979.
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The Com~issioners in Enclosure 2, the six issues listed below are recommended for designation as new Unresolved Safety Issues.
~~ 1. Long-Term Upgrading of Training and Qualifications of Operating Personnel
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Operating Procedures
~Control Room Design
~- Consideration of Degraded or Melted Cores in Safety
~ Reviews
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Shutdown Oecay Heat Removal Requirements
- 6. Seismic Qualification of Equipment in Operating Plants Four related candidate issues were combined into one of the above recommended Unresolved Safety Issues, Control Room Design.
A brief discussion of each of the thirty-five candidate issues that were considered but not *reco!T'GTlended for reporting is provided in Encl osur'e 4.
Seven of these issues have been designated as requiring "further study" to determine whether or not they should be reported as Unresolved Safety Issues. Further investiga-tion of these issues will be conducted over the next several months and a decision made as to whether any should be reported as Unresolved Safety Issues in the 1980 NRC Annual Report.
In preparation of the 1980 Annual Report, the NRR staff will also review all generic safety issues identified in NUREG-0510, but not designated as Unresolved Safety Issues at that time, to determine whether any of these warrant designation as USI's in the light of today's know-1 edge.
The Advisory Corrrnittee on Reactor Safeguards has been briefed on the review process described in Enclosure 2.
The Corrmittee, however, has not had the opportunity to provide its advice on the results of the Staff's review because of the limited time available to conduct the review and achieve NRR management and Corr.mission approval before the July deadline. The Committee's advice will be solicited as our efforts continue to select additional
<J The Corrrnissioners Unresolved Safety Issues for reporting in the 1980 NRC Annual Report.
The NRR staff has begun to prepare the NUREG report that will be transmitted to Congress in accordance with the commitment in the 1979 NRC Annual Report. The informa-tion contained in that report will include:
- 1. a description of the review process as provided in ;
- 2. a discussion of each issue selected as provided in Enclosure l; and,
- 3. a discussion of our plans to develop detailed Task Action Plans and allocate resources for the resolu-tion of the selected issues.
In connection with Item "3," NRR fiscal planning is based on the addition of about six new Unresolved Safety Issues in FY-81 and six more in FY-82 after which it is estimated that new Unresolved Safety Issues will be added at a slower rate of about three per year.
The special report is due to the Congress in July. For this reason, it is important that we obtain Commission comments and approval by no later than mid-July to allow sufficient time to revise the text of the report to reflect Coliiillission comments and have it printed before the end of July.
Reco~mendation: That the Corrmission:
(1) approve the six issues recorrrnended for designation as new Unresolved Safety Issues by NRR; and, (2) note that NRR will prepare a NUREG report for transmittal to the Congress as outlined above.
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,*. Coordination:
The Office of the Executive Legal Director has no legal objection.
Enclosures:
- 1. Description of Proposed New Unresolved Safety Issues
- 2. Description of Unresolved Safety Issue Selection Process
- 3. Results of Initial Screening
- 4.
Summary Discussion of Candidate Issues Not Recormiended as Unresolved Safety Issues Harold R. Denton, Director Office of Nuclear Reactor Regulation NOTE:
Corrrnissioners' comments should be provided directly to the Office of the Secretary by c. o. b. Wednesday, Ju1y 23, 1980.
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT July 16, 1980, with an information copy to the Office of the Secretary.
If the paper is of such a nature that it requires
- additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
Distribution Commissioners Commission Staff Offices Executive Dir. for Operations ACRS ASL BP AS LAP Secretariat
DESCRIPTION OF PROPOSED NEW UNRESOLVED SAFETY ISSUES This enclosure includes descriptions of the 6 issues proposed for designation as new Unresolved Safety Issues. Five of the issues are derived from items included in the TMI Action Plan. The sixth issue regards a concern related to the seismic qualification of equipment at older plants.
Long Term Upgrading of Training and Qualifications of Operating Personnel The ability of operator*s and technicians of nuclear power plants to correctly respond to abnormal conditions and to avoid errors which could lead to abnormal conditions is principally dependent upon the individual's training, experie!'lce, and education.
A number of varied incidents t*hat have occurred throughout the history of commercial nuclear power, and in particular the TMI-2 accident, have involved errors of omission or commission by the operations I
personnel. Consequently, the risks associated with human errors could be signi-ficantly reduced by improving the scope and content of the training programs
\\.*. for reactor opera tors, specifying min_imum training requirements for other operations personnel, and imposing stricter qualifications requirements for all operations personnel.
10 CFR Part 55 currently specifies the requirements for reactor operator qualification and requalification. As a result of the short-term recorr:nenda-tions of the TMI-2 Lessons Learned Task Force, the staff issued ~ Jetter to all power reactor licensees and applicants, dated March 28, 1980, which speci-fied revised criteria for reactor operator qualifications that could be im-plemented under the existing regulation, and provided guidance for operator training.
Related short-term actions have been taken to expand the scope
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of the licensee training programs, to emphasize the team aspect of the operations personnel and utility management, and to provide broad~r based training for mitigating core damage.
These short ter.n actions are not part of this proposed Unresolved Safety Issue.
A related short-ter.n activity involved a study performed by the Basic Energy Technology Associates (NUREG/CR 1280). This study compared the selection, training, and qualification practices of the nuclear industry to those used in the Naval Nuclear Propulsion Program.
The results of this study presented a number of recommendations which will be considered in this proposed Unresolved Safety Issue in developing training and qualification require-
'ments for nuclear power plant operations personnel.
The principal objective of the proposed Unresolved Safety Is.sue is the longer-tenn development of new regulations and regulatory guides which will provide improved requirements for the training and qualification of reactor operators, senior operators, shift supervisors, auxiliary operators, technicians, and possibly other operations personnel. These requirements will be developed from studies of selection, training, and qualification programs by the staff and contracted consultants in the field.
In a related program (NUREG 0660, Task I.A.4} that is not part of this proposed Unresolved Safety Issue, the staff will review improvements in reactor*
simulators that could enhance the training of operations personnel.
This activity relates specifically to Task I.A.2.6 in the TMI-2 Action Plan, NUREG 0660.
The revised requirements and the subsequent rulemaking activities are expected to be completed in approximately two years.
r Operating Procedures The actions perfonr~d by plant operators for both nonnal plant operation and off-nonnal plant conditions are described in a set of written procedures.
These instructions reduce the reliance on the operator's memory in order to assure the proper sequence of manual actions. A number of reported events have been directly related to deficiencies in the written procedures. This experience has suggested that the procedures are not sufficiently explicit and may not contain sufficient diagnostic information to assist the operator to quickly identify and readily cope with abnormal.conditions.
In addition, the inter-relationship between the administrative, operation, test, surveillance, and* maintenance procedures may contribute to events, when the required actions are not clearly defined.
Consequently, the potential for procedural errors can be significantly reduced by providing consistent format and content to the procedures and improving the delineation of symptoms, events, and plant conditions that identify abnonnal situations.
For the short-te'r.n, the staff has required {letters dated September 13 and 27, Octobe~ 10 and 30, and November 9, 1979) that licensees and applicants per-form analyses of several accidents and transients and from the results of these analyses, develop improved operational procedures for off-noTiilal plant conditions.
The short-term actions include clarifying the delineation of authority, shift change practices, and ~ontrol room access. These short term actions are not part of this proposed Unresolved Safety Issue.
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Although the short-ter:n actions are considered adequate to ensure the health and safety of the public, this proposed Unresolved Safety Issue involves development of a detailed program plan for the upgrading of plant procedures. This plan will include guidelines on procedure content and for:nat review procedures, and auditing techniques that could provide significant additional improvements in the operating procedures and further reduce the potential for procedural-related errors.
The plan development will include consideration of related criteria resulting from other activities, such as the system response analyses, reliability anaiyses, human factors engineering, crisis management, and operator training.
In addition, specific emphasis will be placed on guidelines to assure that procedures identifying symptoms of accident and transient
- scenarios not presently being investigated, with selected consideration of current accident and transient scenarios (e.g., small break loss-of-coolant accidents, steam generator tube rupture, loss of feedwater, and uncovering the reactor core). Tne resultant plan will foTill the basis for reviewing operating procedures for off-nonnal plant conditions for their quality and diagnostic capabilities.
This proposed Unresolved Safety Issue relates specifically to Task I.C.9 in the TMI-2 Action Plan, NUREG 0660.
The plan is scheduled to be completed in 1981.
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Control Room Design ihe design and layout of a nuclear power plant control room can significantly affect the operators' ability to deal with abnonnal plant conditions.
This conclusion was drawn by several of the studies of the lMI-2 accident.
The operators' effectiveness in periods of high stress, following an accident or severe transient, is dependent upon both the type of information provided and the manner in which it is displayed.
By improving these aspects of the control room design, the potential for human errors can be significantly reduced.
Consequently, the objective of the proposed Unresolved Safety Issue is to establi~h improved design requirements and standards for the control room instrumentation and arrangement and identification of important controls.
As a first step, the staff will establish guidelines and requirements for control room design reviews, including site visits to establish existing control room design capabilities. In addition, the staff will establish requirements for a plant safety paramefer display console. These activities are expected to be completed in approximately a year.
As a second step, the staff will develop final control room design require-ments, rel a te.d standards and regulatory guides, and improved control room instrumentation research.
The res~arch activities will investigate audio-visual alanns, plant surveillance instrumentation, and post-accident monitoring instrumentation. Revised regulatory requirements and implemen-tatjon schedules are expected to be issued ir. mid-1982.
This proposed Unresolved Safety Issue re1ates to Tasks I.Q.l. 2, 4, and 5 in the n!I-2 Action P1an, NUREG 0660.
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Ccnsiderat~on of Degraded or Melted Cores in Safety Review Historically, the design basis for nuclear power plants has been predica:ed on preventing core damage.
Consequently, very little investigation has been done regarding provisions to deal with degraded or melted cores.
As a result of the TMI-2 accident, the staff is now considering the extent to which the plant design basis should include damaged core conditions.
Provid-ing equipment and procedures to de~l with damaged core conditions can signi-ficantly reduce the risks to the public from events that go beyond the original design basis.
For the short-term, the staff has issued criteria (letters dated Septe~cer 13 and 27, October 10 and 30, and November 9, 1979) which require that licensees and applicants develop design provisions for reactor coolant syste~ ve~ts, access shielding, and post-accident sampling.
Requirements and guidelines are currently being developed to train operations personnel to deal with core damage events.
In additi-0n, the staff is currently reviewing design changes and additional measures that could reduce the consequences of a severe accident for plants located in high population densities (Zior. Ur.its l and 2 and Indian Point Units 2 and 3). These interim ~easures are intended to provide an immediate increase in the capability to deal with degraded core conditions.
The principal objective of the proposed Unresolved Safety Issue* will be the longer-term rulemaking that will formalize the design requirements for degraded core conditions. The rulemaking activity will specifically consider the use of filtered-vented containment systems, the use of core-retention devices, design criteria for decay heat removal, radwaste, and ventilation-filtration systems, provisions for post-accident recovery, criteria for locating highly radioactive systems, and the effects of multiple reactors on a given site. This activity relates specifically to Task II.B.S in the TMI-2 Action Plan, NUREG 0660.
A number of related activities have bearing and may potentially impact this rulemaking.
These include core melt research studies (NUREG 0660r Task II.8.5), siting policy (NUREG 0660, iask II.A.1 and 2), emergency preparedness (NUREG 0660, Task II.A.2), and the licensing actions for Zion and Indian Point (NUREG 0660, Task II.8.6). This last activity will have to be closely coupled with the proposed Unresolved Safety Issu~.
The rulemaking is expected to occur in approximately*two years, depending on the extent of public corrrnent, the progress on research* and design studies, and the possible need for a hearing.
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Shutdown Decay Heat Removal Reouirements Following a reactor shutdown, the radioactive decay of fission products continues to produce heat {decay heat) which must be removed from the pri~ary system.
The principal means for removing this heat in a pressurized water reactor (PWR), in the absence of a large loss-of-coolant accident, is through the steam generators to the secondary side of the plant.
The Reactor Safety Study (WASH 1400), later reliability studies, and related experience from the TMI-2 accident have shown that the loss of the capability to remove heat through the steam generators is a significant contributor to the probability of a core melt event. Although many improvements to the steam generator auxiliary feedwater system were required by the NRC following the TMI-2 accident, providing an alternative weans of heat removal would substantially increase the plants' capability to deal with a broader spectrum of transients and accidents and potentially could, therefore, significantly reduce the overall risk to the public.
Consequently, the proposed Unresolved Safety Issue will investigate aiternative means of decay heat removal in PWR plants, using existing equip~ent where possible. This study will consist of a generic systems evaluation and will result in recorrunendations regarding the desirability of and possible design requirements for an alternative decay heat removal method, other than that normally associated with the steam generator and secondary system.
This activity relates to Task II.E.3.3 in the TMI-2 Action Plan, NUREG 0560.
The study is scheduled to be completed in 1981.
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Seismic Qualification of Equipment in Operating Plants The design criteria and methods for the seismic qualific~tion of mechanical and electrical equipment in nuclear power plants have undergone si~nificant change during the history of the commercial nuclear power program.
Conse-quently, the margins of safety provided in. existing equipment to resist seismically induced loads and perfonn the intended safety functions may vary considerably.
The seismic qualification of the equipment in operating plants must, therefore, be reassessed to assure the ability to bring the plant to a safe shutdown condition when subjected to a seismic event.
The need for such a reassessment was identified as a result of experience with the Systematic Evaluation Program (SEP) for eleven older operating plants.and the staff's Seismic Qualification Review Team (SQRT) reviews of recent operating license applications. During the course of the SEP and SQRT activities, the staff identified a concern with the anchoring and supports used for electrical equipment in the SEP plants.
- These plants have been required to resolve this issue by September 1980.
An Information Notice concerning this issue was sent to all other operating plants. The staff has concluded that if sufficient anchoring is provided for equipment, it should function properly in the event of an earthquake, during the interim period until the overall seismic qualification issue can be resolved.
The objective of the proposed Unresolved Safety Issue fs to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants. Tnese guidelines are expected to be completed in 1981.
On a longer term basis, a research study is currently undeJioiay that will establish probabilistic methods to estimate the margin of failure for the seismic design of structures, systems, and components.
The results of the research studies will be used to confinn the guidelines developed under this proposed Unresolved Safety Issue.
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DESCRIPTION OF THE UNRESOLVED SAFETY ISSUES SELECTION PROCESS In order to evaluate safety concerns, recommendations or general safety issues and determine if these should be designated "Unresolved Safety Issues" and reported to Congress as such, the process described below was developed.
This process was intended to provide a systematic and consistent approach to evaluating these issues and judgin~*their impact on risk to public health and safety.
By amendment of December 13, 1977 to the Energy Reorganization Act of 1974, Section 210 required that a plan be developed for the identification and resol ufion of Unresolved s*afet.v issues and that progress reports be provided in the annual report to Congress.
The following definition of an unresolved safety issue was developed to satisfy the intent of Section 210; this definition was used in identifying the Unresolved Safety Issues previously reported to Congress by NUREG-0510 in January 1979:
11An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves condi-tions not likely to be acceptable over the lifetime of the plants affected."
In applying this definition, matters that pose "important questions concerning the adequacy of existing safety requirements" were judged to be those for which*
resolution is necessary to (1) co~pensate for a possible major reduction in the deqree Qf protection of the public health and safety, or (2) provide a potentially siqnificant decrease in the risk to the public health anrl safety.
Those issues that satisfy (1) above are basicall.v "backward" 1ookin9; that is, they brinq the degree of protection back up to the assumed level. Those issues that satisfy (2) are "forward" lookino; that is, they are improve-mer.ts that decrease the risk significantly below the assumed leve1.
The process that was developed includes a*set of initial screeninQ criteria to screen out those issues that do not satisfy certain elements of the definition for an Unresolved Safety Issue. This initial screenino was done without. addressing the issue's importance to safety. A screening using roughly the same criteria was used in identifying Unresolved Safety Issues for the 1978 NRC Annual Report.
The following criteria were used for the initial screening:
Initial Screenina Criteria (an issue or recorrrnendation has been screened out from further considerations as an Unresolved Safety Issue if it meets one or more of the following)
- l.
The issue or recommendation is not related to nuclear power plant safety, e.g., transportation of radioactive materials.
- 2.
A staff position on the issue or recorrmendation has been develooed or could be developed within 6 months.
The purpose of this criterion is to eliminate those issues that are near resolution and, therefore, do not constitute truly "unresolved" issues. Such issues do not warrant the attention and resources nonnal ly associated wi.th an Unresolved Safety Issue.
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The issue is not generic.
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The *issue or recommendation is only indirectly related to nuclear power plant safety, e.g., reconmended changes in the licensing process. NRC organization, etc.
- 5.
Definition of the issue requires long tenn confirmatory or.exploratory research.
The basis for this criterion is that investigative studies of matters for which no clearly defined safety deficiency or improvement has been identified, although an appropriate regulatory activity, do not warrant designation as Unresolved Safety Issues.
~. The issue or recommendation is related to one already being addressed as a USI and can reasonably be or already is included in the current program.
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The issue or recoli?Tlendation requires a policy decision rather than a techni cal solution. The purpose of this criterion is to elimjnate those is~ues t 1 only require a management decision and do not represent potential deficienc*
in existing safety requirements that require development of a resolution.
In some cases, the results of these policy decisions may require designatio:
of new Unresolved Safety Issues.
The results of this initial screening were a set of candidate Unresolved Safety Issues tb be evaluaterl for importance to safety. Enclosure 3 provides the surna~y list of candi1ates and the results o7 the initial screenino.
"-,.... To assess the importance to safety of each of these issues, a set of questions was develooed to assist in evaluating the issue's general impact on various factors affectinq safety. Attachment 1 to this Enclosure ?rovides an overview of the process. After the initial s~reeninq, the issue is: (1)-identified as either a deficiency or an improvement; (2) deter~ined to be an issue related to either operations, equipment, or emergency response; and, (3) evaluated in teTiils of potential for siqnificantly affe~ting risk.
We evaluated the impact of each candidate issue on:
probability of an accident or transient; pr9bability of losing rnitigatin~ functions, given the event; and consequences, Qiven the event and lo~s o& mitioatino functions, by answerinq various questions as shown i~ Attachment 1, and then arrivin~ at an overall conclusion based on these answers as to the potential for siqnificantly affectina fission product barrier inteqrity, frequency of transients or accidents, safety functions, or emergency response capability. For example, if the issue beina evaluated passes the initial screening criteria, is deter~i~ed to be a deficiency (i.e., possibl~
major reduction in 'the assumed deqree of protection), is prinarily an equipment concern, and impacts on the capability to perform safety functions, the foll0\\'1-ing questions would be asked:
. (1) What is the ootential deficiency?
(This *question is intended to obtain a clear and concise description of the deficiency, including identification of plants affected and the cause of the deficiency).
(2)
What is the likelihood that the potential deficiency exists?
(The answer should be "Low", "Low to Medium", "Medium," "!1edium to High," or 11Hiqh.
11 If insufficient in.formation is available, the ans\\"er
r.,, could be "Further Study" to assess the likelihood. Quantitative information could be used if avai1ab1e).
(3)
What equip~ent/systems could be affected by the potential deficiency?
(Should identify those safety and non-safety systems that could be affected) *
. (4)
What is the likelihood that, given the above deficiency, the affected equipment will fail as a result?
(Use similar likelihood estimates as described for ouestion (2) above).
(5)
What safety functions could be affected by failure of the equip~ent/
systems?
(Identify safety functions that may be performed usinq any of the equipment/systems identified in response to question (3) above).
(6)
~/hat is the likelihood of loss of the affected safety functions when needed, if the affected equipment/systems fail?
(Use similar likelihood estimates as described for question (2) above).
(7) Based on the above, is it likely that a safety function will be lost due to this deficiency?
(Based on the answers to questions (1) through (6), a conclusion was made as to the potential for losing a safety function due to this deficiency.
If potential is judged to be significan~, the answer is
. i "yes", and the deficiency is desi9nated as an Unreso.lved Safety Issue.
If insufficient infonnation is available to answer any of these question, the answer is that "Further Study" is required to obtain the information necessary in order to determone if the deficiency should be an Unresolved Safety Issue.)
A sample question fonn that was used for an operations concern (i.e., possible
~ajor reduction in the assumed degree of protection due to a potential ooeraticns' deficiciency) is shown in Attachment 2 to this Enclosure.
Where possible, quantitative infor~ation was used to answer the questions and arrive at conclusi:r.s on potential impact. However, in most cases relevant quantitative information was not available so.that qualitative estimates were developed for the likelihoods and the conclusions were based on these. The quali-tative estimates were based on the engineering judgement of individuals knowledgeable of the issue, with input and review by a group of staff personnel from various technical disciplines. These. included an individual from the Probabilistic Analysis Staff and an individual with plant operations experience.
The result of this evaluation is: (1) a set of new Unresolved Sa7ety Issues; (2) a set of issues requiring further study to detel"Tiline if they shouid be Unresolved Safety Issues; and (3) a set of issues that are not Unresolved Safety Issues.
The ita~s that are identified as Unresoived Safety Issues will be evaluated and resolutions established in the sa~e manner !$ the previously identified Unresolved Safety Issues. This will include development of Task Action Plans with a high staff priority placed on resolution of the issue. Progress on these issues wil 1 be provided
~imonthly in the 11Aqua 11 Sook, and in the annual report to Congress
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f.:.':,,* Items requiring further study will be evaluated to determine if they should be classified as Unresolved Safety Issues. Results o~ this further study will be included in the next annual report.
Items not classified as Unresolved Safety Issues may still have some benefits in terms of safety, and accordingly will have priorities assigned to them by the Safety Program Evaluation Branch and will have their resolu-tion monitored by the Generic Issues Branch.
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POSSIBLE MAJOR REDUCilON rn THE ASSUMED DEGREE OF PROTECTION OF THE PUBLIC HEALTH AND SAFETY Operations Concern -
Title:
Long Term Program for Upgrading of Procedures (Task I.C.9)
What is the potential deficiency?
Descriotion:
Present procedures, as evidenced by the TMI-2 accident, may not be expEcit e~ough and may not present enough of a diagnostic and syste~atic set of instructior.s to assist the operator in his efforts to quickly identify and readily cope with a.~
abno~mal situation.
Also, the interrelationship of ad:Dinistrative, ope~atir.5, test, surveillance, and maintenance procedures may actually contribute to trar.sie~ts and accidents, if ~he required actions are not clearly defined.
What Plants are Affected?
All
\\./hat Tv~e of Def.iciencv? Operator Qualifications Training -(Procedures)
Equipment Aides - Technical Support What is the likelihood that the potential deficiency exists?
Likelihood High Remarks The review of several procecures has shown them to be ambiguous, hard to follow, replete with refere"-ces to other procedures, etc.
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Title:
I.C.9
~hat is the 1ikelihood that the potentia1 deficiency will result in operator errors by plant personnel?
Likelihood Medium - High I:!P.O.CT ON FREQUENCY OF TRANSIENTS/ ACCIDENTS Reinarks Errors in procedures lead the operators to mistrust the procedures.
Although the operators understand the pla.,t a.,d syste=s extremely well and don't usually need to rely on the procedures for i!D.I:lediate actions, this underlying mistrust could lead to errors of commission or omissio*n in tices of high stress.
Wha! equipr.;ent/systerr:s could be affected by these errors:'
All systems are potentially affected, but the greatest ecphasis should be on those areas in which errors during maintenance or testing could result in a transient and in which errors during an accident could increase the severity.
What is likelihood these equipment/systems will fail or maloperate; given these errors'?
~1edium, depending on the system involved and, more important, on the evolution being performed with a faulty or a!!lbiguous procedure.
Title:
I.C.9 What transients/accidents could be initiated as a result of these equipment/
syste~ failures or ma1o~erations?
A.,y number of transients is possible, such as that which occured at Oyster Creek when the iQproper operation of a valve caused a high pressure signal that ultfoately isolated the reactor.
The only accidents postulated are those in which operation of the components affecting reactor coola.,t pressure integrity (e.g., relief valves) are performed improperly.
What is the likelihood that these transients/accidents will result, given the
../
f.,
1 equi~~enw sys~ern ai ures or r.a opera~ions.
- Like 1 i hood High Remarks Icproper operation of the syste~ in the mann~r postulated above will result in a substantially increased likelihood of the transients described.
3ased on the above, is the frequency of occurrence of the transient/accident sisni-ficantly increased by the potential deficiency?
Yes.
Title:
I.C.9 H~?ACT 0;1 SAFETY ru~:CTIC~!S,iECE'SSARY* TO *mT!GATE EVENTS AND CONTAIN STORED RADIOACTIVITY what equi?rnent/syste~s could be affected by the resulting operator errors?
All systems and equip~e~t.
What is the likelihood that the affected equipment/systems will fail, given that these errors occur?
Likelihood Medium Remarks Alarms, indicators, and the other operators could cause an error to be corrected when that error is the result of an a.cbigudus procedure.
What safety functions could be affected by failu~ of the equipment/systems?
The ability to citigate the accident and retain stored radioactivity, and all paths for acco~plishing safety functions th.at rely on operator actions i~
accorda.'ice with the established procedur'es.
Title:
I.C.9 What is the likelihood of loss of the affected safety functions when needed, if affected equipment/systems fail?
Likelihood Medium Remarks As before, alarms, indicators, and the other operators serve to mitigate the errors in system operation caused by ambiguous procedu..-es.
In addition, the safety-related syste:s typically have backup syste~ available.
Based on the above, is it likely that a safety function will be lest When needed due to this potential deficiency?
There is a "medium" probability that.a safety function could be lost when needed due to misuse or misapplication of the procedural controls.
Conclusions/Basis This task should be designated an Unresolved Safety Issue.
Events at ~I-2 and the review of other facility have shown the procedures to be less than satisfactory and in most cases aobiguous.
They require operator evaluation and shifting to other procedures at a time (accident conditions) when such a luxury does not exist.
... (
Sl\\FETY FUNCTION PURPOSE Sl\\FETY FUNCTIONS TO MITIGl\\TE EVEtlTS AND CONTAIN STORED Rl\\DIOl\\CTIVITY rrnronrt/\\tlCE OF SAFETY FUflCT IOrtS
/\\REAS OF POTEtlTll\\L DEFICIENCY Tlll\\T COULO AfFFf.T PEnFOrH11\\tlCE OF EQUIP~ENT on OP[tt~TORS
============1=============================================~====================
React tv1ty Control Reactor Coolant System Inventory Control Shut reactor down to reduce heat production Maintain a coolant medium around core Reactor Coolant System Maintain the coolant in the pro-Pressure Control per state Core lleat Removal lransfor heat from core to a coolant Reactor Coolant System*
Transfer heat from the llcal Removal core coolant Con ta inmcn t I so la t ion Containment Tempera-ture and Pressure Control Combustible Gas Control Ma tntenance of Vital Aux11 tarics Indirect Radioactivity He 1 t!il se Con t.rn 1
- Close openings tn containment to prevent radiation releases Keep from dama91n!) containment and equipment Remove and redistribute hydro9cn to prevent explosion inside con-tainment Maintain operability of systems needed to support safety systems Con tat n m I see 11 aneous s to reel racl to.1c t Iv I ty to 1>ro Lee l puh 11 c
- 11111 ii VO Id II I sl.rilC t I 11~1 oper.l t.nr~;
Equipment Operators Oesi9n Construction, fabrication, Installation Materials Testing Inspection Maintenance Quality Assurance*
Qualification Verification Qua l 1fica t tons Training Operating Procedures Equipment Aides Technical Support
ENCLOSURE 3 RESULTS OF. fNIT!AL SCREENiNl.i This enclosure includes the results of the initial screening of all issues considered for reporting as Unresolved Safety Issues. The screening criteria used are described in Enclosure 2.
The issues are grouped in tnis enclosure by source.
The three principal source groups are:
{l) The TMI Action Plan, (2) ACRS Letters/Reports and (3) Other Potential Issues. The screening results for all issues are presented.
Each issue was determined to be either (1) screened on the basis of one or more of the screening criteria, (2) included within the definition of another issue being considered (e.g., many of the ACRS recomrner.dations were included in TMI Action Plan items), or (3) a candidate issue for further consideration for reporting as an Unresolved Safety Issue.
The candidate issues were subjected to further review as described in Enclo-sure 2.
A list of all candidate issues preceeds the screening results for t.he three proups of issues.
y y
TM! ACTION PLAN ITEMS I t e:n LIST OF CANDIDATE UNRESOLVED SAFETY ISSUES AFTER INITIAL SCREENING
- 1.
Section I.A.2.2 - Training and Qualification of Operations Personnel
- 2.
Section I.A.2.6 - Long-tenn Upgrading of Training and Qualificatior.s
- 3.Section I.A.4.2 - Long-tenn Training Simulator Upgrade
- 4.
Section I.A.3.3 -
Require~ents for Operator Fitness
- 5.Section I.B.1.1 - Organization and Management Long-Tenn Improvene~ts
- 6.
Section I.C.9 - Long-Tenn Program Plan for Upgrading of Operating P~oce~~~es
- 7.
Section I.D.l - Control Room Design Reviews
- 8.
Section I.D.2 - Control Room Design - Plant Safety Parameter Display Cons'.)ie
- 9.
Section I.D.4 - Control Room Design - Control Room Design Standard
- 10.Section I.D.5 - Control Room Design - Improved Control Room Instru0::ntati~n*
Research
- 11.
Section I.F.l - Expand QA list
- 12.
Section I.F.2 - Develop More Detailed QA Criteria
- 13.
Section I.G.l - Preoperational and Low-Power Testing - Training Require;:;e;.ts
- 14.
Section II.B.8 - Consideration of Degraded or Melted Cores in Safety Rev1:*,... -
Rulernaking Proceeding
- 15.
Section II.E.2.1 - Reliance ~ ECCS
- 16.Section II.E.2.3 - Emergency Core Cooling System - Uncertainties in Performance Predictions
- 17.Section II.£.3.3 - Decay Heat Removal - Coordinated Study of Shutco:-:n He::
Removal Requirements.
- 18.Section II.E.4.3 - Containment Design - Integrity Check
- 19.Section II.E.5.1 - Design Sensitivity of B&W Reactors - Design Evaluation
- 20.
Section II.E.6.1 - In Situ Testing of Valves
- 21.
Section II.F.4 - Study of Control and Protective Action Design Requirenents
- 22.
Section II.F.5 - Classification of Instrumentation, Control and Electrical Equipment
- 23.Section III.A.3.5 - Improving NRC Emergency Preparedness - Training, Drills, and Tests
- 24.
Section III.D.1.3 - Radiation Source Control - Ventilation System and Radioiodine Adsorber Criteria
- 25.Section III.D.3.3 - Inplant Radiation Monitoring
- 26.
Section II.K.3.33 - Action Plan Recorrmendations - Evaluate elimination of PORV function ACRS Ite!ns
- 27.
Reliability of Ventilation Monitoring Equipment ZS.
Prdtective Device Reliability
- 29.
Instrumentation Set-Point Drift
- 30.
End-Of-Life and Maintenance Criteria
- 31. Design Check and Audit of Balance-of-Plant Equipment
- 32.
BWR Control Rod Worth
- 33.
- 34.
Inadvertent Actuation of Safety Injection
- 35.
~eevaluation of Reactor Coolant Pump Trip Criteria Other !tens
- 36. Turbine Disk Cracking
- 37.
D.C. Power System "Reliability
- 38.
BWR Jet Pump Integrity
- 39. Seismic Qualification of Equipment at Operating Plants
- 40.
Small Break LOCA from Extended Overheating of Pressurizer Heaters
y
-~-
- 41.
PWR Pipe Cracks
- 42.
BHR Main Steam Isolation Valve Leakage Control Systems
- 43.
Radiation Effects on Reactor Vessel Supports
- 44.
Loss of Offsite Power Subsequent to a LOCA r,.. :..
r _,..
- 1.
Kr?':
Tnltra~ ~creenlng frlterfa (an issue or recommendation has been screened out from further considera-tions as an Unresolved Safety Issue if it meets one*or more of the following)
The issue or recom~endation is not related to nuclear power plant safety, e.g., transportation of radioactive materials.
- 2.
A staff position on the issue or recorrrnendation has been developed or could be developed within 6 months*.
The purpose of this criterion is to*
eliminate those issues that are near resolution and, therefore, do not
- constitute truly "unresolved" issues. Such issues do not warrant the atten-tion and resources nonnally associated with an Unresolved Safety Issue.
- 3.
The issue is not generic.
- 4.
The issue or recorrr.iendation is only indirectly related to nuclear power plant safety, e.g., recoJTTnended changes in the licensing process, NRC organization, etc.
- 5.
Definition of the issue requires long terin confirmatory or exploratory research.
The basis for this criterion is that investigative studies of matters for which no clearly defined safety deficiency or improvement has been identified, although an appropriate regulatory activity, do not warrant designation as Unresolved Safety Issues.
- 6.
The issue or recommendation is related to one already being addressed as a US! and can reasonably be or already is included in the current program.
- 7.
The issue or recorrmendation requires ~policy decision rather than a techni-cal solution.
The purpose of this criterion is to eliminate those issues that only require a management decision and do not represent potential deficiencies in existing safety requirements that require development of a resolution.
In some cases, the results of these policy decisions may require designation of new Unresolved Safety Issues.
INITIAL SCREENING RESULTS TM I ACT ION PLAN ITEMS
USI INITIAL SCREENING TMI ACTION PLAN ITEMS ACT ION PLAN ITEM I. A. 1 Operating Personnel and Staffing
- 1. Shift Techni ca 1 Advisor
- 2.
Shift Supervisor Admin. Duties
- 3.
Shift Manning
- 4.
Long-term Upgrading I.A. 2 Training and Qualification of Operating Personnel
- 1.
lrrrnediate Upgrading of Operator and Senior Operator Training and Qualifications
- 2.
Training and Qualifications of Operations Personnel
- 3. Administration of Training Programs for Licensed Operators
- 4.
NRR Participaticin in IE Inspector Training
- 5.
Plant Drills 6 *. Long-Term Upgrading of Train mg and Qua l ificat1ons
- 7. Accreditation of Training Institutions l.A.3 Licensing and Requalification of Operating Personnel
- 1. Revise Scope and Criteria for Licensing Exams
- 2.
Operator Licensing Program Changes
- 3.
Requ1rements for Ope~ator Fitness
- 4.
Licensing of Additional Operations Personnel
- 5.
Establish Statement of Understanding wi~h INPO and DOE INITIAL USI CANDIDATE SCREENING
- 2 2
2 7
2 Candidate 2
2 2
Candidate 4
2 2
Candidate 7
4 ACTION PLAN ITEM INITIAL USI CANDIDATE SCREENING 1.1\\.4 Simulator Use.and Development
- 1. Initial Simulator Improvement 2
- 2.
long-Term Training Simulator Upgrade Candidate
- 3. feasibility Study of Procurement of NRC Training Simula tor 4
- 4.
feasibility Study Of HRC Engineering Computer 4
)or*.:
l.B.1 *Management for Opcrat1ons
- l. Organization and Management long-Tenn Improvements Candidate c
- 2.
Evaluation of Organization and Management Improvements of NTOL Applicants 2
- 3.
Loss of Safety function 4
- 1. n. 2 Inspection of Operating Reactors
- 1.
Rev1se IE Inspection Program 4
- 2.
Rcsiclcnt Inspector at Operating Reactors 2.4
- 3.
Regional Evaluations 2,4
- 4.
Overview of l iccnsee Perfonnance 2.,,
l.C Operating Procedures
- 1. Short-Term J\\ccidcn t Analysis and Procedure Revis ion 2
- 2.
Shift and Relief Turnover Procedures 2
- 3. Shift Supervisor Responsibilities 2
II.
Control Room J\\cccss 2
- 5.
Procedures for Feedback of Operating Experience 2
- 6.
Procedures for Verification of Correct Perfonnance of Operating Activities 2
- 7.
NSSS Vendor Review of Procedures 2
i ACTION PLAN ITEM
- 0.
Pilot Monitoring of Selected Emergency Procedures for NTOL Applicants
- 9.
Long-Tenn Program Plan for Upgradrng of Procedures I.D Control Room Design
- 1.
Control Room Oes1gn Reviews
- 2.
Plant Safety Parameter Display Console INITIAL USI SCREENING 2
CANOIOJ\\TE Candidate Candidate Candidate
~~:--~~~~~~~~~~~~~~~~~~~~~:;:---::--~~~~~~~~~~-<
- 3.
Safety System Status Monitoring 2,7 tl.
Control Room Design Standard
- 5.
Improved Control Room Instrumentation Research
- 6.
Technology Transfer Conference I.E Analysis and Disse11lnation of Operating Experience
- l. Office of Analysis and Evaluation of Operational Data
- 2.
Program Office Operational Data Activities
- 3.
Operational Safety Data Analysis
- 4.
Coordiniltion of Licensee, Industry, and Regulatory Pro9rams 5.- Nuclear Plant RelialJility Data System G.
lleportin9 Requirements
- 7.
Foreign Sources O.
lluman Error Rate Analysis 4
2,4 2,4 2,4*
2 ',,
2,4 4
Candidate Candidate Included under issues more directly related to plant safe(
C. IJ.' IllEP.
'. *~
ACTION PLAN ITEM I.F Quality Assurance
- 1.
Expand QJ\\ l st
- 2.
Develop More Detailed QI\\ Criteria 1.L_Preo_P-erational and Low-Power Testing
- 1.
Training Requirements 2
Scope of Test Proqram
- 11.
Siting and Oesign I 1./\\ Siting
- 1.
Siting Pol icy Reformulation
- 2.
Site Evaluation of Existing Facilities II.n Consideration of Degraded or Melted Cores in Safety Review
- 1.
Reactor Coolant System Vents
- 2.
Plant Shielding to P-rovide /\\ccess to Vital /\\reas and Protect Safety
- 3.
Post-accident Sampling INITIAL USI SCREENHm 2
7 7
2 2
2 CANO IO/\\ TE Candidate Candidate Candidate c:
II.
Training for fllt1gating Core Oamage 2
- 5.
nesearch on Phenomena /\\ssocialed with Core Degradation and Fuel Melting 5
- 6.
Risk Reduction for Operating neactors at Sites with lli!]h Population Densities 2,J
- 7.
Ana 1 ys f s of llydro!Jen Cont ro 1 J
- -------n-. -,lilYClihif111~rrl>cP.cli_1_ri~-
11.C nellilhility Engineering and IUsk /\\ssessmcnt
- l.
Interim Reliability Evaluation Program (TnEP)
?..
Contin11alion of WEP
).
Systems l11Leractio11
~
11, 11,.1i.1hili1v r1111i11l'l'r*i1111 Cand 1 cla te InEP and fol low on programs arc viewed as major programs that 111.1y identify new issues, but are not is~11es in thcmsel ves.
(i I 1. r. 11 i *; vi 1*vwd,,.,
.111 1 lU 1'
-!j-ACTIOH PLAN JT[M INITIAL USI CANDIDATE SCllEEIHNG I I. D lleac tor Coolant System Ile l i ef and Sa Fe ty Valves
- l. Testing Requirements 2
- 2.
Research on Relief and Safety Valve Test Requirements 2,5
- 3.
Relief and Safety Valve Position Indication 2
I I. E System Design 11.E.l Aux il i a ry F eedwa ter System I;.*
- l. Auxiliary Fcedwater System Evaluation 2
- 2.
Auxiliary Feeclwater System Automatic Initiation and Flow Indication 2
- 3.
Update Standard Review Plan and Develop Regulatory Guide 4
11.[.2 Emergency Core Cooling System
- 1.
Reliance on ECCS i--
Candidate
- 2.
Research on SmiJ 11 Break LOC/\\s and Anomalous Transients 5-
- 3.
Uncertainties in Performance Predictions Candidate 11.E.3 Decay lleat Removal
- 1.
Reliability of Power Supplies for NiJtural Circul~tinn 2
- 2.
Systems Relia~ility 1.c.1. I.C.2
- 3.
Coordinated Study of Shutdown Ilea t llcmova 1 lleq u*i remen ts Candidate
~
,,. /\\ltcrnat.e Concepts Research 5
- 5.
Regulatory Gui de I --
- -. t'\\
ACTION PL/\\N ITEM II.E.'1 Containment Desi9n
- 1.
Oedicated Penetrations
- 2.
Isolation Oependability
- 3.
integrity Check
'1.
Purging 11.E.5 Design Sensitivity of O&W Reactors
- l. Design Evaluation
- 2.
01.W Reactor Transient Response Task Force 11.E.6 ln-5itu Testing of Valves
- l. Test Adequacy Study I*l.F Instrumentation and Controls
- l. Additional Accident Monitoring Instrumentation
- 2.
Identification of and necovery from Conditions Leading to Inadequate Core Cooling J.* Instrumentation for Monitoring Accident Conditions (Reg. Guide 1.97)
INITlf\\L USI SCllEENI NG 2
2 2
2 2
2 2
C/\\NO 10/\\TE Camlldate Candidate Candidate
~. S t~!_<!Y_,~x_J:g_~!!'!'l and ProtecUJ~_.f\\ct ion Des.iJJn Re_'lu ~e~~e~_ ts ---.--.------1-------____
1,_Cand i~da_t_e ____
~
- ,~lass1f1cal10n otfnstr11mcnl:at10n. Contro1 aliCJElectnct1IEq111pmenl Candidate
--rr:GK°"----.E..... I ectnca 1 1 Power
.c.___ ______
11------1---------
I l
- 11
- 1.
Power supplies for PORV. block valves. level indicators TMl-2 Cleanup and Examination
- 1.
Maintain Safely of TMI-2 and Minimize Environmental Impact
- 2.
Ohtain Technical Data on the Conditions Inside the TMI-2 Containment St rttture 2
3 3
I\\ r:T I Ot-1 f' l Ml IT £M INIT l/\\l USI SCHEEN rNG C/\\NOID/\\TE
--~-1-~~-------c
- 3.
Evaluate and feedback Information Obtained from TMI
- 4.
Determine Impact of TMI on Socioeconomic and Real Property Values 11.J General Implications of TMI for Oesiun and Construction Activities 11.J.l Vendor Inspection Program
- l. Establish a Priority System for Concluctin9 Vendor Inspections
- 2.
Modify Existing Vendor Inspection Program
~.
Increase Re9ulatory Control Over Present Non-licensees II.
J\\ssi911 Resident Inspectors to Reactor Vendors and /\\rchitect-Engineers 11.J.2 Construction Inspection Pro9ram 4
5 4
4
- 1.
Reorient Inspection Program More Toward Direct Observation, Proper Work 4
Pcrfonnance, and Verification of as-bu1lt Configurations Versus Desi9n
- 2.
Increase Emphasis on Independent Measurement in the Construction Inspection 4
Program
- 3.
~~si~n Resident Inspectors to all Construction Sites 4
11.J.3 Management for Design and Construction
- l.
- Organiz~tion and Staffing to Oversee Design and Construction
- 2.
Issue Regulatory Guide 11.* J.4 Revise Deficiency Reporting Requirements
- l.
Revise Deficiency Reporting Requirements 4
4
. 1 ****---------
ACT! Oii PLAN ITEM INITl/\\l USI CJ\\NOID/\\TE SCrtHNING I I. K Measures to Mitigate Small-Break LOC/\\s and loss of feedwater Accidents
- l. IE Oulletins 2
- 2.
Conuni ssion Orders on O&W Plants 2
- 3.
Final Reconuncndat1ons of O&O Task Force See Table (57 ft~ns total. see Table C.3 attached)
C-3 III. Emergency Preparations and Radiation Effect$
I 11./\\.l Improve Licensee Emergency Preparedness.. Short Tenn
- l. Upgrade Emergency Preparedness 2
2 *. Upgrade Licensee Emergency Support facil 1t1es 2
- 3. Maintain Sup~lies of Tl~rofd Olock1ng Agent (Potassium Iodide) 7 11./\\.2 Improving Licensee Emergency Preparedness - long-term
- l.
~nend 10 CfR 50 and 10 CFR 50, Appendix E 1.2
- 2.
Development of ~uidance and Criteria 2
11.J\\.3 Improving NRC Emergency Preparedness-
- l. NRC Role* in Respondi"ng to Nuclear Emergencies 2
- 2.
Improve Operations Center 7,2
- 3.
Conniun i cations 2
- 4.
Nuclear Data Link 2
- 5. Training, Orills, and Tests Candidate
- 6.
Interaction of NRC with Other J\\gencf es 4
- :~
"t
.. * -c
-. ------*-------~------ -----------------------*----------.-----~---------<
/\\CTI ON f'L/\\N I HM II I. 0 Emergency Preparedness of Sta le and Loca 1 Governments
- 1.
Transfer of Responsibilities to FEMJ\\
- 2.
Implementation of NRC's and FEMJ\\'s Respons.ibilities 111.C Public tnfonnation
- 1. llave Infonnation Available for the Hews Media and the Public
- 2.
The Office of Public Affairs will Develop Agency Policy and Provide Training for Interfacing with the Parties 111.D.l Radiation Source Control
- 1.
4 1
1 Included in 11.0.0 CJ\\NDIO/\\TE
- 2.
Radioactive Gas M_a_na~9.._e_m_e __ n_t _______________________
~_5 _____ ----------c J.
Ventilation System*and Radioiodine Adsorber Criteria Candidate
'1.
Radwaste System Design Features to /\\id in /\\ccident Recovery and Decontamination --
11.D.O 111.D.2 Public Radiation Protection Improvement
- 1.
Radiological Monitoring of Effluents 5
. 2.
Radioiodine, Carbon-1'1, and Tritium Pathway Dose Analysis 5
J.
Lifluid Pathway Radiological Control 5
- 4.
Offsite Dose Measurements 5
- 5.
Offsite Dose Calculation Manual 2
- 6.
Indeperrdent nadiological Measurements 2
111.D.J Worker Racli at ion Protection Improvements
.*. ' : *~
/\\CT ION 1'1.f\\N ITEM
- 1.
Radiation Protection Plans
- 2.
lleal th Physics Improvements
- 3.
lnplant Radiation Monitoring
- 4.
Control Room llabitab11 ity
- 5.
Radiation Worker Exposure Data Oase IV.
Practices and Procedures JV.A Strengthen Enforcement Process
- 1. Seek Legt sl att ve Authority
- 2.
Revise Enforcement Pol icy IV.O Issuance of Instruction and Information to Licensees JV.0.1 Revise Practices for Issuance of Instructions and Information to Licensees IV.C IV.C.1 IV.O IV.0.1 Extend Lessons Learned Extend Lessons Lea med NRC Staff Tra1ntnu NRC Staff Trainin9 to Licensed Activities Other Than Power Reactors from TMI to Other NRC Programs IV.E Safety Decision-Making
- 1.
Expand Research on Quantification of Safety Decision-Making
- 2.
Plan for Early Resolution of Safety Issues
- 3.
Plan for Resolving Issues at Construction Pennit Stage
- 4.
Resolve Generic Issues by Rulemaking
- 5.
Assess Currently Operating Reactors 4
4 4
4 4
5 4
4 7
7
- .'* ACTION PLAN ITEM IV.F Ffnancfal Disincentive to Safety
- 1.
Increased IE Scrutiny of Power Ascension Test Program
- 2.
Evaluate the Impact of Financial Disincentives to the Safety of Nuclear Power Plants IV.G Improve Safety Rulemaking Procedures
- 1. Develop a Public Agenda for Rulcmaking
- 2.
Periodic and Systematic Reevaluation of Existing Rules J.
Improve Ruelmaking Procedures
- 4.
Study ~lternative for Improved Rulemakfng Process IV.H
~RC Participation in the Radiation Policy Council
- v.
NRC Pol icy, *organization, and Management
- 1.
Oeve 1 op NRC Po 11cy Statement on Safety
- 2.
Study El fmination of Non.. safety Responsibil Hies Strengthen Role of ACRS
- 3.
- 4.
- 5.
- 6.
- 7.
- 0.
- 9.
- 10.
- 11.
- 12.
Study Need for Addi t ion a 1 Adv 1sory Con111ft tees Imr>rove Puhl k and Intervenor Participation in llearing Process Study Construction-Durin9.. J\\djud1cation Rules Study Heed for TMI-Related Legislation I
Study the Need to Establish an Independent Nuclear Safety Ooard Study the Reform of the Licensing Process Study NRC Top Manage1nent Structure and Process Reexamine Organizatio" and Functions of NRC Offices Revise Dcle9ation of J\\utl1ority to Staff INITI/\\L USI SCREENING 4
4
.,
- 7 14,7 4,7 4,7
~
~.7 l
~
7 7
7 7
7 4,7 4,7 4,7 7
13 *
- 14.
Clarify and Strengthen the Respective. Roles of Chair111<1n, Con11ifssion, and rno 1.7
..J\\uthority to Oclegale Emergency Response Functions to a Sin9le Conunissioner 7 l !i.
- /\\cllievr. SiwJle l.oc:iltion -
1011~1-l.r.nn 7
C/\\ND ID/\\ TE i--
i.-
~-
~ *..
J\\CTION PL/\\N IHM INITI/\\L USI Cl\\NDID/\\TE SCREENING
. *****----------------------------------------1-::...;:_-----t---------c
- 16.
Achieve Single Location - Interim
- 17.
Reexamine Co11111ission Role in Adjudication TABLE C.J - TASK Il.K.3 ACTION Pll\\N RECOMMENDATIONS
- 1.
Install automatic PORV isolation system and perfonn operational test
- 2.
r~eport on overall safety effect Qf PORV isolation system
- 3.
Report safety and relief valve failures prom ly and challenges annually
- 4.
Review and upgrade reliability and redundancy of non-safety equipment for sma 11 break LOCA mitigation 7
- 1, 7 2
2 2
2
- 5.
Continue to study need for Table C. 1.4.c and need for automatic trip of RCPs, 2
- 6.
- 7.
- 0.
then modify procedures or designs as appropriate Instrumentation to verify natural circulation Evaluation of PORV opening probability during overpressure transient Further staff consideration of need for diverse decay heat removal method independent of SG's
- 9.
Pressure Integral Derivative controller modification 2
11.E.3 2
- 10.
Anticipatory trip modification proposed by some 1 icensees to confine range 2
of use to high power levels
- 11.
Control use of PORV suppl fed by Control Components Inc. untfl further review 2
comp 1 etc *
- 12.
Confinl\\ existence of anticipatory trip upo'n turbi_.~e trip 2
- 13.
Separt1tion of llPCI and RCIC system initiation levels. l\\nalysis and implemen-2 tat ion
- 14.
Isolation of isolation condensers on high radiation 2
- 15.
Modify break detection logic to prevent spurious isolation of llPCI and RCIC 2
systems
- 16.
Reduction of challenges and failures of relief valves - feasibility study and 2
system modification
- 17.
llcport on outacJe of ECC systems - licens(!e report and proposed technical 2
spt~c i fka t ion <:ha11qe!;
r.:*;::.;... ***------**-------------------------------------------.------,---------:
/\\CT ION l'L/\\N ITEM
- 10.
Modification of l\\OS 1o9ic - feasibility study and mod1f1cation for 1ncreased diversity for some event sequences
- 19.
Interlock on recirculation pump loops
- 20.
Loss of service water for Oig Rock Point
- 21.
Restart of core spray and LPCJ systems on low level - design and modification
- 22.
l\\utomatic switchover of RCIC system suction
- verify procedures and modify design
- 23.
Central water level recording 2'1. Confinn adequacy of space cooling for llPCI and RCIC systems Effect of loss of /\\C power on pump seals Study effect on RllR rc11abil ity of its use for fuel pool cooling Provide common reference level for vessel level instrumentation Study and verify qualification of accumulators on ADS valves
- 25.
- 26.
- 27.
- 20.
- 29.
- 30.
Study to demonstrate perfonnance of isolation condensers with non... condcns1bles Revised small-break LOCI\\ methods to show compliance with 10 CFR 50, /\\ppendix K\\
- 31.
- 32.
Plant-specific calculations to show compliance with 10 CFR 50.46 Provide experimental verification of two-phase natural circulation models 3'1.
REL/\\P-'1 model development I
35.. Evaluation of effects of core flood tank injection on small-break LOCl\\s JG.
/\\dditional st;1ff auclit calculations of ll&W'sm.. 111-break LOCI\\ analyses
- 37.
/\\nalysis of 11/1.W plant response to isolated small-hre;ak LOCI\\
311.
/\\nalysis of plant response to a small-hreak LOCI\\ in the pressurizer spray
- 39.
rv,1luation of effects of Willer sluqs in pipin~ Ci111Scd hy llPI and CFT flows*
'10.
[Villuation or ncr seal d.Ullil!JC i111<.I lcak.:1ge tlurfnu.1 small-hrcak LOCI\\
'11.
Suhmit prcclictions for I.OFT Test U-6 with llCPs r11nnin!1 line INITJ/\\l USJ SCHEENING C/\\NO 10/\\ TE
--~-i*-------~:
2 2,3 2,3 2
2 2
2 2
11.E.2.1 2
2 2
l 2
2 6,
2 6,2 2,
2, 2,
2,
?,
11.E.2.:
I. C. 1 I. C. 1 I. C. 1 J.C. 1 I.C. 1 I. C. l Candidate
": *~
-H-
/\\CTION Pl/\\N ITEM INITIAL USI SCHEENING
'12.
Submit requestctl infonnation on the effects of non-condens1ble gcJses
'13.
Evaluation of mechanical effects of slug flow on steam generator tubes 1111.
Evaluation of anticipated transients with sinule failure to verify no signi-ficant fuel failure
'15.
Evaluate depressurization with other than full /\\OS
- 46.
Response to list of concerns from /\\CRS consultant
'17.
Test program for small-break LOCI\\ model verification pretest prediction.
test program and model verification
- 40.
/\\ssess change in safety reliability as result of implementing ll&OTF reconuncnda-t ions 119.
Review of procedures (NRC)
- 50. leview of procedures (NSSS vendors)
- 51.
S:Ymptom-based emergency procedures
- 52.
Operator awareness of revised emergency procedures
- 53.
Two operators in control room 5'1.
Simulator upgrade for small-break LOC/\\s
- 55.
Operator monitoring of control board
- 56.
Simulator training requirements
- 57.
Identify water sources prior to manual activation of /\\OS
- 2. I.C.l 2
2 2
2 I. c. 1 11.E.2.2
- 2. I.C.1 2, J.C. 9 2, l.C. 9 2, J.C. 9 2, I.ll.1 2
2, J./\\.'1.1 2
2 2
C/\\NOID/\\TE
INITIAL SCREENING RESULTS
/\\CRS. LETTrns /\\rlD REPOIHS
Oate 1 /9/79 2/15/79 2/1 G/79 3/12/79 3/13/7'J 3/11/79 3/111/79 3/21 /79 4/7 /79
'1/10/7'J
'1/1 B/79 i
USI INITIAL SCREENING
/\\CRS LETTERS - J/\\NU/\\RY 1979 TO 11/\\RCll 1900 (Note:
"Pass" under Initial Screenino
- Column means that the issue passed the initial screeninq criteria)
Subject Recommendation/Issue Concurrence in RG 1. l '11, Rev. 1 None Report on Salem Unit 2 Establish cri terfo for implementing RG 1. 97 Report on Fluor Power Services OOP None Std. SAR -
u~w 205 Std. NSSS Concurrence in Rf; l. 137, Rev. 1 and Mone l _ 14'.L llPV l
Transportation of Rad, t.1at'ls None
!leport on Zi11n11er Unit l None Combination of Dynamic loads Mone Status Rpt. No. 7 on Generic Items No new qeneric items Interim Report on Tftl -2 /\\cc i dent
- Further Analyses of transients anrf accidents in PHRs that involve initially or at sometime durinq their course, a sma 11 hreak in the orimary~stem Additional instrumentation for operator to fo 11 ow sma 11 breaks - provision for unambi CJUOU s l eve 1 in reactor vessel.
llioh-noint nrimarv svc;tr.m vr.ntc;.
- Concurrence in R('; 1.1'10, Rev. 1 None Transmits copy of 11/17 /79 l\\CRS oral Detailed analyses of M.it'l' Circulation recommendations on TMI to supported hl'. experimcn t.
Commissioners Placinq of Pressurizer heaters on emerqency power for pressure control clurin'l nat 'l r.irc.
IJSI Initial Screening
- No issue No (2)
No (J)
No issue No ( 1 )
Mo (3)
No (6)
No (6)
Pass (See Note above)
No(2)
No(?.)
No issue Pass tlo(2)
Tf-1 I Action pl 1.c.1-C II.K.3-c
(
n.11*e Suh i cc: t Rccon*nenda t ion/ Issue Suhcoolinci meter ilnd flow exit temp.
instruments ilS indication of nat'l circ.
Use of core thermocouples - other in-I s trumen ts for onera tor assistance
~/20/79 11emo to Gil fo s ~.Y on 1mp1Cii1en t 1 n~J Mone
'1/17 /79 ora 1 reco111nenda t ions 5/15/79 Review of Proposed Rules on Ship11Jent None of Spent Fuel 5/16/79 Report on Quantitative Safety Goa 1 s Establish quilntitative safety goals for overa 11 safety of reactors.
5/16/79 Interim Rpt. 03 on TMI Unit 2 Reactor Vessel 1 evel instrumentation (sanr. as /f 3).
Develop changes to operator qualifications, training and licensing.
Eval. of Lrns -
Review of o~erating ~rocedures Improver! reliability of offsite and ons ite /\\C and nc.
Station Olackout - Water llanl1'1er Licensce/NRC [merqcncy Planninq Oesiqn to facilitate the decontamination and recover~ of major reactor systems.
Staff Review Procedures
--'~-~J"_!ilh i 1 ity of NllC S t.1 ff
--~~~~l!'~_C:.Y._~__f lh_~~9Je fi!..!l~!!*e er i terion.
__ Sil_~~l:Y.. ~~!~~~'.1-!:~~!_ ___
)
rnn I ro I *1 I'd filll't'l'd I/ f' II t. *j I\\ '1 () f Ulll I.iii 111111*11 t*
US I Initial Screening No(2)
No(2)
No issue 1
Pass No(2)
No(2)
No ( '1)
No{2)
No(6)
Mo(6}
No('1,2)
Pass No('1}
No{ 4}
Pass
__ !!g_l,,J pi\\~~~)
THI l\\c t ion f4 JV. E.,
11 I.0.1-4 11.0.0.!!
11.C.2 IV.E.1
*--(
- 11. ILO
- ... -~
.., Oate Suhiect lleconvnencla ti on I Issue 5/16/79 Interim Report 112 Nil tura 1 Ci re. Procedures -
Pressurizer heaters from Emergency Power Sources S11hcoolin9 meter Use of core exit thermocou~les Instruments to follow the course of an accident.
Safety Research - Trclnsients//\\cciclents Status monitorinrt of availability of ESF 6/10/79 Power level increase at Millstone None Unit ?.
6/10/79 Comparison of Stainless steel and None Zircaloy Fuel Rod Cladding (to '1ilinsky) 6/19/79 Concurrence in RG 1.9, Rev. 2 None 6/19/79 Requests, /\\qrecments and Conmli tmcnts None 7 /16/79 Report on Bailly Pile Foundations None 7/lll/79 Relationship of NllC & 00[ llesearch None
!2!L!.runroyrcl llr.ador Safety 7 /l 9/79 llescarch Rcriuest Ueq11ire111ents None n11 J/79 Report on Short Term Lessons Learned Reexamination of llydrogen r.enera ti on a ncl (others already incl udcd a hove or in Control.
STLL)
B/1 ~/79 Near Term Operatin9 Licenses None USI Initial Screeninq No (2)
No (2)
No(?.)
No(2)
No(2)
No(5)
Pass No(J)
No issue No issue No issue No(J)
Uo issue No issue Pass No issue TMI fiction Pl (lr1 -
I. 0. 3 11.0.0
USI Initial TMr
~>~~e ____ ~~~Jr.c_t_* --**--------------------*--n
__ ~_c_on~nr.!Hla ti on/ Issue Screcninq
/\\ctlon r
*------<<----*-----(
ll/1'1/79 Sturlies to l111p1*ove llcaclor Safety 0/16/79 Pipe Cracking in LWRs Pass r r. E°. 2-4 Systematic Rer.valuation of Compressed rass 11.C.2
__ /\\ i r_~ys te!ns.
*-------- ------- ----c*
S turly of interrelationship between /\\HIS, MHJS, /\\fl an<I Control Systems.
Stucly of effects of shared systems.
Study of how operatinrr procedures should be written.
Reevaluation of basis for environmental qua 1 if i cation for equipment.
Oesign, Construction & Operation Review Reevaluation of f)llft desi9n basis.
Review of Reliability Improvements that could be achieved by employinq direct safety signals Systems Interactions lnvolvinq /\\ir. In-strument and llydraulic Lines Pass rass Pass No(2}
tlo(2},(3) rass Pass No(6)
/\\nalyses of transients that would lead Pass to riross overfi 11 i ng of secondary side of steam generator and equivalent in owns.
lJse of probabilistic techniques to examine Pass the reliability of safety systems.
Expan<1ed Scope of p1pe crack sturlies to rass address potential consequences, e.q.,
increased potential for co11111on mode failures. Develop optimum water specificatio 1s.
11.C.2 n-:-n~
11:0:0 l.C.9 IJ.E.3.3 I J.E.J.1 11.F.4; II. C.2 IV.E.l, II.C.2 +
11.C.'1
/\\lready eluded a an item "Other P tent i a 1
-~----~-~-~--i-~~-~~~~*~.:-Tssue 9/7'J neview of Lrns (1976-1970) NIJREG-0572 Separation of Control Roel from its <1rive Pass I
and mm lliqh Rod Horth Events (3.2.l.l, piHlf' 1-?.).
C.111rl i d.1 l
O.i l:e ----
9/79 (Continued)
Suhiect
/ neco111111endiltion/tss11e Unavailability of Vital Services (3.2. l.2, page 3-2)
Failures nue to Hater llanmcr and Fl01-1 Vi bra ti on (3.2.1.3, paqc 3-2}
Slstems Interaction (3.2.1.4! (!age 3-3}
Valve Failures ( 3. 2. 1. 5, pa ~e 3-3) leakage Between Interconnected Fluid
. Slstcms ( 3. 2. l. 6, pa9e J-'1)
Problems in Containment Isolation and Monitorinq (3.2.1.7, page 3-'1)
Failure of Containment Monitorinq Systems Oue to Environmental Condi lions (3.2.1.0,
_P-a~1e 3-'1) lnildequate Design Criteria (3.2.1.9, page 3-5)
En~ineered Safety Features Oeqraded by lhtman Errors (:l.2.1.10, page 3-5}
Loss of Iii qh-Prcssure Coolant Inject ion
( llPC I } and Heactor Core Isolation Coolinq (RCIC) Systems (3.2.1.11, [!il 11e 3-6)
Failures in Moni torinq Equipment for
/\\ir-Clcaninq and Vcntilatinq Systems (3.2.2.1, parre 3-f>)
Failure to neco11nize and Correct the Cause of an Event (3.2.2.2, pa oc 3-6)
Failures of Protective Devices for Es-sen ti al Eciu i pmcn t (J.2.?..J, paqc 3-7)
USI Initial Screeninl]
PilSS Hater llammer No(G) rlow Vibration Pass Mo{6}
Pass No(2)
No(2)
No(2)
No(4)
Pass Pass Pass Pass Pass Tt1I.
/\\ction Pli11 It. E. 3-'1 c
Candidate 11.C.2 c
11.1\\.1,J.C I.D.1-5 II. K.3-024 Candidate I.E.6 Candidate
.... j., Delle Suhject Reconanenctat ion/ J s sue 9/7'J Failures of Oiesel Generators (Continued)
(3.2.2.'I! ~age 3-7)
Set Point Drift in Instrumentation (3. 2. 3. 1 1_Qa ge 3-0)
End-of-life and Maintenance Criteria (3.2.3.2, page 3-n)
Inadvertent Actuation of Safety Injection in PWRs {3.2.3.3 1 ~age 3-0) 10/9/79 Xenon releases in accidents Stud.v to determine the appl icabil Hy and desireahility of available technolo9y to minimize release of rad. noble gases.
10/11/79
$[P None 10/12/79 Systems Interact ions Study for IP 03 Consideration for Phase II of 1\\-17 11 /l'l/79 l\\CRS Action on Proposed Rev's. to
( l )
Make R.~. 1.1'11, Rev. 1, imple-Reg Guides.
mentation consistent with Lessons Learned recommendations.
(2)
Concur in R.G. 1.97, Rev. 2 being issued for public comment.
ll/1c1/79 NUREG-0600, Investigation of TMl-2 Prepration and Issuance of a Sunmary Report Accident by l&E of the findings of the various NRC Task Forces on HU
- 12/10/79 RecoMnendations of President's Concur in Reco111nenda tion by President's Commission re~1a rd in!) l\\CRS Activities Co111nission to strenqthen J\\CRS role.
12/11/79 Adequacy of Procedures for Transmitting
/\\CHS It cconanenda t Ions may rcqu I re Comm Is s I on Arns co111nen ts to the NRC staff attention to indicate priori t.v anrl author-izatiun of resources.
i 12/11 /79 Interim low Power Operation of Plans by staff and TV/\\ to monitor per SP.flllOYclh N11clr.ar Power Pl.ant Unit 1 fonmmr.e of ice condenser containments shou 1 rl he i11111 l cmen led.
USI Initial Screeninq No (2,6)
Pass Pass Pass Pass tlo ( '1)
No(6)
No(2)
No issue No(4)
No(4)
No( '1)
No(J)
TMI*
Action 1 r er Candidate Candidate Candidate 111.D.1-2
~ *,
- -. 0 t i1.c S Ii t
11 1. ec *
- r.
I ti /I
.ccommenr a on ssue 12/11 /79 Co11111ents on the Pause in licensing (1) Perform 1 owpower test i nri of systems or the entire plant.
(2)
For new plants that meet tlUREri-0571l, allow startup and testinq at 50-75X level; then shutdown, available for call in the event of a national need, until sufficient resolution of TMl related requirements to allow nonnal operation.
(1) Expeditiously act on Recornnendations from TMI Task Forces, the President's Connlission, and /\\CRS, but assign a timetable for imrlementation (i.e.,
shortterm items, lonq term).
( 4)
Provide guidance to early CP plants or those not yet in construction to general guidance and design changes that will result from TMI.
12/13/79 Identification of NRC Regulatory A procedure to deffne /\\CRS participation Requirements which need changing in rulemakinq would be useful.
12/13/79 Report on Tm-2 Lessons learned (1) Personnel Oualifications Task Force Initial Report
. Better definition of qua11ff-cation reciu1renmnt~ for Shift Supervisor.
. Adequacy ~f staffing of perator Licensing Branch
. Criteria on adequate degree of in-house technical capability for each 1 i censee;.
. Availability of NSSS technical support for design changes or accident conditions.
(2)
Emerqency Procedures
. tJHC C]l VI'? Jlriori t_y to P.mP.rC]Cncy nrnrPd11rnc: ;if*
nnnr;i~inri 11l.1nf"<;
s USJ Initial crcenrng Pass tlo ( 4)
No(7)
No(4)
No( 4)
No(2)
No(4)
Pass No(2), (3)
No(~)
- A t.
P1 c 1on a* -
I.G.
1.0.1. 1
- -~.,
---* S11hJ£_c_t _______________ _
12/13/79 (Continued)
-n-USI TM!
lnitiill Pc~o11111cn<la ti on./_J_s_s __ 11_e ___________
~..,...S_cr_c_c_n_i_n..,..'J ___
1 ~c t1 on ~l.
L icr.nsees use intcrdi sci pl inary qroups to cleve 1 op these procedures *
(3) Miln/Machine Interfilce -
licensee evaluation should include data recording and recall require-ments of parameters critical to safety.
(4) Reliability Assessment - Licensees perfonn Reliability assessments in addition to IREP.
Pass Pass Pass (5) Studies by licensees of hydro9en con-Pass trol and filtered venting systems.
{ 6)
I ntcrim mea surcs for ice-condenser containments.
(7) Periodic review of NRC rules, philo-sophy, etc.
(fl). netter definition of NRC role in cmerqency response.
(9)
Need to develop methods to uncover significant design errors, equipment degradation, and to test systems under conditions simulating transient and accident situations.
Pass No('1)
No(4)
No(2)
(10) Effect of lar']e raclfoactive release on Pass abi 1 ity to sa fc ly shutdown another unit on the Silme site.
(11) Studies to ascertain contingency de-siqn measures, beyond LL Task Force reco11n11enda ti ons, that may i prove ahil Hy to cope with accidents heyond the design has*is.
Pass 1.c.. 9 1.0. 2 11.C.4 11.D.O 11.D.O 11.D.8 11.D.O
- ..9_
Oil te USJ Initial TMI Action Phr<
Subject Reconnnencla t ion/Issue Screen mg
~~~~~-1-~~~~~~~~~~~~~~~~~11--~~~~~~~~~~~~~-'--~~~--.~~~~~~~-..~~~~-1c 12/13/79 (Con ti nuerl) 12/17 /79 Review of NRC Regulatory Processes an'I Fune ti on s b.v the ACRS, dated December, 1979 (12) Seismic capability of auxiliary Pass feedwater supplies; failure of non-seismic Class I equipment, earthquake emerqency procedure.
(13) Provision of dedicated shutdown heilt removal systems, possibly able to function at normal system pressure.
(1) Regulatory function requires strong leadership (2) Oversi ~ht conm1ittee is not re-quired.
(3)
(.,)
Future orCJanizational arrangements at NHC ~houl cl recognize need to investigate beyond the "design basis".*
"eed to strenqthen certain NRC staff functions.
Pass No ( 4)
No ( 4)
No ( 4)
No(4)
(~) Role of ACRS should be strenqthened.
No(4)
(6) Nuclear industr.v must strengthen Pass its technical and manaqerial capability to handle s~fety matters.
(7) Knowledge gained during plant design Ho(2) ~ (3) and construction should be trans-ferred more effectively to personnel responsible for plant operation.
(8) Consider accidents beyond design basis on future reactors; should furtlier reduce prol>abil i ty of
. serious accidents ;ind provide mea-sures to mitiqate their consequences.
(9) Silfet.v ~o.ils should he as qoo<I as rr..isonahl.Y achievilhle Pass Pass
( nepr..i t from
!i/H1/l'l lPLl.Pr) 11.C.3 I I.E.3.3+4 1.0.1-1 I 1.0.8 IV.E.l.
-=
"'.j.
'.*I Oate Suh.feet Recon111enda ti on/Ts sue 12/17 /79 (10) l.lse quanlitative approach to (Continued) set safety criteria; assess en-hancement *of safety; and to provide comparative risk asses.sment to other technological aspects of society.
( 11) Mod1fy single-failure-criterion for more consideration of pro-gressive, conunon cause, and multiple failures arisinq from a sinqle initiating event; Establish reliability re<Juirements for components. eqt.* and systems based on safety importance.
(12) Use separate and dedicated ~afety systems, consider safety influence of nor.
( 13) Give increased attention to systems interact ion.
( l '1) Give more attention to man/machine interactions. *---
(15) Seek improvements beyond those re-quired by regulations; investigate filtered vented containment, dc*licatcd shutdown systems and de-sf ~1n changes to reduce the pro bah n ity of successful sahotaqe, ancl. imple-ment those found appropriate.
( 16) l\\pply techniques of probabilistic analysis to iflen t if.Y areas for reducinu risk.
- ,_\\
USJ TMI Initial Screeninq l\\ction Pi.
---c Pass IV.Ll,.
(Repeat from II.C:.2, 0/14/79 letter) 11.C.4 Pass (Helated to I I. C. 2 item in 5/l 6/79 letter); also part of Systems Interaction-1\\-17 Pass (Related to IV.E.l, 9 and 10 above)
II *E:~
c Pass 11.E.J.3 11.E.3.4 11.C.2 No(6)
Pass J\\l so 1>art -
1\\-17. II.Cl 11.C.'1 Pass 11.0.8, 11.E.J.4 Pass (Repeat of 11.c *. 2 item in R/1 '1/79 11.C.'1 letter).
- **-.*---'. *:*'<"'* *'M'*
y.. USI lni l.il TMl P!!JQ _____ Su~jec: t ______ --*-------------*-----' _Pcc~~~'.!_e!)cl*"!_~f!"l!!LJ2~!~~-**--------- Screen ~!~~L _____
l\\_c_t_1o~!_1_,~
12/17 /79
{Conti nuecl) l/15/00 neconu11cndations of President's Co11111ission on /\\ens Role
{ 17) Peri odi ca 11 y rccx.imi ne operating plants in lif)ht of current criteria iln<I s tan<lards.
No ( '1)
(10) Reorient NllC safety research to assisting resolution of identified safety concerns ilnd hy exploring for issues or problems of potential s i qnif i ca nee.
No ( '1 )
(19) JlefJuires NSSSvendors, 1\\-E'g.=rnd No(2) licensee's to report safety concerns raised in their orqanizations (I/\\W f>ilrt 21).
(20) Maintilin hi9h level of competence in Pass NSSS organizations or develop equivil-lc11t source. of knowledge to support licensees.
(21) Chanqe approach hy /\\-E and plant owners to emphasize safety.
(22) Design check and audit Of nor (23) Continue to develop Un evaluation program.
No(4) IV.F of TMI /\\P Pass (Rel atcd to item 12 above)
No ( 4)
(2'1) Encourage stand.ird LWR plant clesirins.
No(7)
/\\CllS Conuncnts No ( '1) 1.0.1.l Candida tel
*-~-~~**'-*.&* ______ _...._ ___.... -..-.. -~-**-* **-***&..*---.. ----**--***-*--**---.-...-------- --------t------
l /15/00 Draft NUnEG-0660 (TMI /\\ction Plan) ncvelop priorities and identify items of No(4)
i------------------------------------
primary_2~p~rtance.~-----------------i----~----~~t--~~~~
Jncl udcd -
2/11 /30 Low Pressure Turbine Oise Cracking neevaluate problems associated with tur-l>ine missiles, safety consequences.
Pass "Other* Po:
tent1a 1 I
~------- -----------------*-----.. -*-------* ***-*-***..... -*--**-*--*------*******-**-**-----*----*------- ------*---- -----c
?./11/llO NlmEr.-OGGO. llr,1 ft 2 (TMI /\\ct ion l'lan)
Sil me as a hove. but rcq11cs ts opportuni ty f o
- No ( ~ )
/\\(((S co111111r.nt h(~forc implementation.
- j !late Subject fleconmcnclation/lssue 2/1 J/no 2/13/00 NRC Acceptance Criteria for Mark I tong Term Program SRV rlischarqe p1prn<<J failure; verifi-cation of Mark I containmenl motlifi-cations.
-~-----~---*-------~----* ****5*t:;c-;i~-t*ii-en--;,;q;*ii.r'Cii1eiiE-*r-o-t=*-rc*r-s-o.iiieT-a-ii<r-Qua1 ifkations of Radioactive Waste mana~iement attention and direction for rad*
-A------a--Sys tern Op~!-~2..~:r~~~~:~-*--*--*-*--*-* -~~~-~~-_:>1.~!~'-~.: ______ _ ---***---*-----
2/14/00 NUREG-0625, "Report of lhe Si tin9 Policy Task Force Criteria on siting should he strenqthencd US I
/
TMI Initial Screeninq
- Action r No(G)
Pass 1./\\.2.2 (in ref<
Pass I I.I\\. 1
!~------**---------------*--------------------- -~--*----.. ----
3/ 11 /00 3/11/00 tlTOL Pl ant Test Pro<<Jrams V~rification testing of decay heat re-moval under upset concli lions; (for low power testin<<J on NTOL's).
No(2),(3)
NTOL I terns From Ora ft 3 of NURE<i-0660, TMI Action Plan (1)
Orrianization anrl Manaqement Ci1p-abilitics at NTOL Applicants.
No ( 2)
~*---~ ------ ----*----*-----c (2)
AE review of Emerqency Procedures.
(3)
Non-scherlulecl random checkinq of operatinq personnel w.r.t. i1bility to cope with accidents.
No{2)
No(2)
~-------------- ---------1-----c.
(4)
Reevalui1te criteria in IE bulletins related to RCS Pump trip; llPS I ter-mination; automatic PORV blocking~
( 5)
Improve control room ha bi ta bi 1 i ty rcriuirements.
(6)
Consider hydrogen accumulation at hi<Jh points in containment in 1 oca ti on of re combiner penetrations.
(7) Utility should have lead in handlinq an emergency.
(n)
Systems for measurino concentrations of contilminants in containment and in effluents should he clcsiqned so t.ha t sa111p 1 es a re reprr.sr.n tat i vr., and should receive attention to iJ'.;Sure
,, clr~q11.1 cy cl IHI rr. l iii I> i 1 i ty.
No(2)- llPI & ronvIJ.K.J -
Pass - rtCP Trip ll.K.3 -
Pass 111.D.3-c
~
Pass I I. E.'1 (rtef.)
No(7)
No(2)
(see TMI J\\ction Plan 11.0.3)
I
-T3-US I Ini Li,11 D~te _____
S_u_h'"'ie
___ r:_t_* -------------------t-Rr.co~.!1_nendi1 ti°'~~'-' s_s_11_e ____. ___ ------* Screeninq
~
3/l l /IJO
(<:ontinued) 3/11 /00 3/12/no Recommenc1iltions of the NRC Task Force on 11ulletins and Orders
/\\CllS Connnen ts on Recommendations of ~me Spec i ill Inquiry Group Hcr1ard in~J /\\CltS /\\cl i vitles (9)
Re<itiire NTOL /\\ppl icants ilnd licensees to develop reliilhility assessments of their plilnts (10) llyclroqen control and filtered vented containment, ancl interim measures for ice-condenser con-lil ir1mcn ls.
(1)
Review criteria for llPI termination ilnd RCP trip.
PilSS (Hell1tecl to r.ilrl ier one)
Mo(5)
PilSS (Relilted to I/II in letter of 3/11 /80 on NTOL Items)
(2)
Availability of diverse heat re-Pilss mova l pil th, sue h as feed "nd bl eecl, is desirable.
(3)
Reevaluate lowerinq of hicih-pressure Pass reactor trip setpoint in B&H plants
( 4)
Auto initiation of aux. f eeclwater Mo(2) in the event of a milin steam line hreilk in containment.
(5)
Schedule for NRC position related No ( '1) to smilll hrcak LOCI\\ analysis should be compatible with schedu 1 e req11 i reel for NSSS vendors to revise their n1ocle ls.
(6)
Extend schedule for review ilncl 1mplc-No('1) 111cn talion.
Use of /\\CllS as al ternalive to Nuclear Safely Ooard; selective /\\CRS role; ildvisory i11Hl not fonrli.ll participant role In l icensinq.
No(4)
/\\ct ion re I I.C.2 RCP:
canclida llPI:
NC 11.E.3.:{
11.E.3.(
I I.E. 5-;f
INITIAL SCREENING RESULTS OTllER POTENTIAL ISSUES
POTENTIAL ISSUE USI INITIAL SCREENING OTHER POTENTIAL ISSUES Degraded Engineered Safety Features - Defic1cncies in electrical distribution system operation and design at ANO Units l and 2 Defi ci enc1 es in Piping Design - Errors in seismic desiqn Three Mile Island Indication of Low Water level at Oyster Creek Damage to New Fuel Assemblies at Surry Deficient Procedures at ANO Unit 1 - Emergency Feedwater System controls positioned such that automatic actuation would not occur Major degradation of primary containment boundary - valve misalignment at Pillisades
- See key on page preceding tables.
- Source Abnormal Occurrence 79-1 Abnormal Occurrence 79-2 Abnormal Occurrence
- 79-3 Abnormal Occurrence 79-5 Abnormal Occurrence 79-6 Abnormal Occurrence
- 79-7 Ahnorma 1 Occurrence 79-0 Initiill USI Screening
- 2 2
2,3 2,3 2,3 2,3 TMI J\\c t f on Plan Section All Related to 11.E.'1.J
~ --* ---*-*-* ----*- -**------* *-*--*---**--* - ---- --
T.
Source Initial TMI POTHIT J /\\I. ISSUE USl
/\\ct ion Screen 1 ng Pltu1 Section Turhine Oisk Cracking Misc.
Candidate r~m Pi r>C? Cracks.
Misc.
Candidate The Need for addf.tional pipe? crack studies should be detennined and if so (5,6) generic tasks developed. mms (Sec Memo Denton to Carbon, 10/211/79).
Misc.
No.
(
/\\n analysis of the prohahil ity and consequences of a loss-of-offsite power Misc.
Candidate suhscquent to a LOCI\\ should be undertaken to determine if i:lctions arc necessary on plants (old l\\WS concern not considered as a USI last year -
sec 3/3/70 memo Boyd to Case).
/\\11 investigation of the effects of control system failures and design in-USI 1\\-17; 1 I.
Misc.
Candidate l I.C.2; 11.r ailequacies is needed (see Gossick to l\\hcarnc mco, 10/22/79).
I. F, l Analyses of the effects on plant safety of (1) failures of non-seismically Misc.
11.C.3 riua 1 i f i cd components 011 the operation of seismically qual Hied components and ( 2 ) mu ll i r> l e failures of non seismically qualified components.
(See letter Fleischaker to Denton, 10/19/79).
Emer!Jency procedures should prepare opera tors for anoma 1 ies in system Misc.
behavior likely to occur during an earthquake (Fleischaker letter).
l.C lluman factors considerations in control room designs should be analyzed Misc.
to optimize operator response during and following a severe earthquake (rlcischaker letter).
J.D Testing rcqui rcmcn ts for valves isolatin!J low pressure systems from the Misc.
No(2) reactor coolant system should be optimized to further reduce the risk associated with an inter system LOCI\\.
(To 1 d Conmi ss i oh we vmul d recqns i der as a US I - see excerpt from SECY-Glfl).
POT ENTl/\\L ISSUE Source Initial TMI USI
/\\ct 1 on Screening Plan Section
~
l\\n ass essment of the potential for and consequences of a sma 11 LOCI\\ from burned Mi SC.
Candidate out pr essurizer heaters should be undertaken.
The issue is one of assessing the need to 1) require filtration [llEPJ\\
and charco.11 (or equivalent)] of PHH auxiliary building and certain other gaseous release pathways; and 2) upgrade requirement~ for char-coal adsorbers to provide greater assurance that radioiodine 1s ef-fectively removed from gaseous effluents.
Presently. release pathways need to be filtered only if it is necessilry to meet /\\ppendix I require-ments* or if filtration is necessary to reduce Part 100 accident conse-quences.
TMl-2 has sh9wn that public tolerance to even small accidental releases is very low.
A systemiltic evaluation to look. at l) accident scenarios; 2) other ir:iprovernents that are being implemented as a result of TMJ-2. such as improved containment isolation and system integrity;.
J) placing all or most systems which contain primJry coolant inside containment; and '1) benefits and cost of requiring filtration which satisfies cxistin9 guidance and benefits and costs of requirinu improved filtration is needed.
(DSE)
OSE Proposal 11.D.O
~-----*--------*--------------------- -----1------~------
Defin il ion of exclusion area boundary over large bodies of water.
(DSE)
DSE No
('1 7)
~P.rqnosal
/\\utomatcd remote real-time acquisition of onsite meteorological data by NRC and persons responsible for making decisions durin9 a nuclear power plant incident.
(OSE)
DSE Proposal 111.A.3.3
*---- ------1-----'---
rtcquircmcnts for upurading of onsite meteorological programs at operating nuclear power plants where serious deficiences exist. (DSE)
OSE Proposal II I.A. 1 POTEIHI/\\L ISSUE
--* -*--""'\\-*--..,.-_ ~~*..--....-,~
.. -,--.--~-,*----~--~,~---
n Main ypc of There have been a number of LERs relating to excessive leakage i Steam Isolation Valves (MSJVs) in OWR facilities.
It was this t experience that led to the development of Regulatory Guide 1.96.
of Main Steam Isolation Valve Leakage Control Systems for Ooilin Reactor fluclear Power Plants" and the Unresolved Safety Issue C-Steam Une Leakage Control Systems."
llowever, not only have the relating to excessive MSJV leakage continued, but recent LERs ha identified failures in the MSIV leakage control systems (MSIV-LC
Design g Water 8 "Main LE~s ve Ss).
He believe that C-0 should be expanded to include:
- 1.
Review MSIV and* MSIV-LCS operating experience in considerati on for a potential revision to RG 1.96.
- 2.
Consider an MSIV rel;ability improvement program. similar to the program currentl~* underway for Target Rock safety/relief va 1 ves ( 0- 55).
We are ~oncerned that consequential control system failure due t the hostile environment (humidity. pressure, temperature, jct impin(_)cmcnt) fol lowin9 a high energy pipe break may occur.
Such failures might compound anticipated accident scenarios and hampe Lite operators ability to successfully cope with a potential acci 0
He have informed all operating plant licensees of our concerns a received and screened Lheir responses.
We have found no specifi hh:ntifiable Sclfcty problem to elate.
We ar:e.rn.ncer11ed about tt1e
.i<lequacy of the hren<llh aml depth of these 1n1 tlnl l \\ccnsee rcv1c
.ind will be proposin9 a 10119 term systematic and methodical plan l>y pl ant review.
This issue may be incorporated in NRC efforts to resolve reconvnc 9, TMl-2 Lessons Learned Task force Final Report, NUREG-0505, Co Systems Interactions.
r dent.
nd c
ws l"
ndation n trol Source Initial USI Screening DOR Candidate Proposal DO fl Candidate Proposal nu l\\c ti on Plan Section
. (11. c. 1 ' I I.C.
11.C.3, 11.F.
11.F.~
2, 4,
-5..
POHNll /\\L ISSUE The Transverse lncore Probe (TIJ') monitoring system in a BHR penetrate$
the primary containment.
In order to provide effective isolation capability when the TIP probes are in the reactor core, each line is equipped with an explosive (squil>) holation valve which will sever and!.al the line
\\-1hen manually actuated.
These valves would be used when containment isolation is needed and the TIP probe(s) could not be withdrawn from the core.
At present, there are no surveillance requirements for these valves, since they Ciln only be destructively tested.
Mevertheless, there is a finite "shelf-life" associated with the squib charges.
Periodic replacement of the squib chargers and circuit checks should be included in the Technical Specifications of each BWR plant, similar to the requirements for the explosive valves in the Standby liquid Control System.
Source DOR Proposal Initial USI ScreeninCJ No. (2)
/\\ct ion Plan Section
*---------------*-----------------1------1------+-----
The l\\rchitect/Engineers I structure and practices should be revie\\*1ed for all operating plants.
This review shoulil include the study of:
- 1) lhe management structure incorporated for the desil.)n and con-struction of the plant to insure thclt they were adequate to enr:ourilgc and al lm-1 for required interfaces between various groups.
- 2)
The required procedures for inter.face between lhe various groups.
- 3)
The mechanisms for insuring that re<Juired procedures were implemented.
The results of this study should he evaluated and the impact on overall plant safety assessed.
DOR Proposal No specific Issue -
Should be 1n eluded under Sys tema tic Review of Operating Plants
~***~
-Ci-POTENT I /\\L ISSUE The rlcsign/analysis/construction practices and procedures ~~Pd by architect/
engineers for all operating reactors should he reviewed to delennine thcit the desi9n and analysis of structures, systems, equipment and components reflected *their as-built configurations and that modifications to these items, required clS a result of design ilnd anillysis, were pro-perly incorporilted.
This review should include iln assessment of I.he impact of the practices and procedures on overilll pl.111t safety.
Source DOR Proposal lnf tial USI
. Screening No Speci fie Issue.. Sys, Pro!Jrarn for nev1ew of Opcril ting Plants TMI
/\\c ti on Plan Section -
-----1-------1------
The methods and criteria (loads, load combinations and acceptance criteria) used for the design and analy~is of structures, systems equipment and components should he re-reviewed for all operilting reactors.
This should not be an effort which is as in-depth as that being conducted under the SEP, but should be detailed enough to identify and cursorily evaluate the methods and criteria actually used in the design and analysis of the plant.
This
\\*1ould supplement and update the FSl\\R's which are either incom-plete, out of date, or do not represent the criteria actually used to <lesign the fe:cility.
DOR Proposal No Spec1fic Issue ~ Should 13e Included Under Sys tema..
tic Review of Operating P I an ts & f Sf\\R Update Rule
............. -********* -*-*-- -*- *-"*....... --** *-*-**-***------*-------*-*--*--*---**-***-**----------.. ----* ------1--------1------
Oesig11, fahrication, ancl operation of the secondary systera of PUils should he revie\\*1ecl and evaluated as they rcl,1te to lon*, te.-m mechanical t\\tHI materials perrormance of the primary systcrr.* pri:11ilry and secondary systems inter.1ction 1 and primary and secondary systems responsr. anti inte']rity during normal and postulated accidents.
Specil'ic items include component desi9n; materials, and integrity~*
si91iificant secon<lilry system transients which may affect primary system inte9rity; and the 9eneral safety aspects of the balance or plant Oil the primary system.
DOR Proposal 11.C.l, 11.C.2 11.C.J, 11.f.5,.
T/\\P-/\\3,M,/\\5 Materials Jmpac on S/G Tube In tc9r~ ty T/\\P-/\\ 17 f Sys ten n tcrac t; l'OTENTI/\\L ISSUE Operability and reliability of safety related p:.JJil::>S and valves* re-quired for safe shutdm-m of power plants in the event of an acci-dent is a requirement of a plant license.
Recent reviews of specific systems and components indicate that these vital components may be designed inproperly. misapplied or that simply manufacturers ClJcll ity assurance is inadequate.
Further, the requirements* for demonstration of operability are either nonexistent or are left to Lile discretion of enclors or 'licensees.
Potential failures of this equipment could affect the operation of required safety systems dur-ing and after a postulated accident.
Safety-related mechanical and electrical equipment are required by the staff to ~e seismically qu_ali_fied.
Rece!lt. reviews !Jf the.s~ism)c qualifications on SEP plants have rnd1cated potent1al genenc dcf1c1enc1es.
These deficiencies arc both in the equipment installation practices and in the seismic qualification procedures and records.
Potential failures of this equipment could affect the operatian of required safely systems during a seismic event.
In a post accident situation, as at TMl-2, gases that evolve from reactor coolant in the makeup and purification system are collected in the waste gas system.
Small leaks in these systems can cause very hiqh concentrations of airborne activity i11 the auxiliary huilclinu and fuel hanclling building and can result in much higher than normal environmental releases via ventila-tion exhausts from these huildinus.
llecausc of the Vilrlcly of problems 11ssociilted with these hi!'}h radiation levels a review of the need for le11k l'iuhtness in these systems or con-li1ir1111e11t withi11 scaled areas should be considered.
Source DOR Proposal DOR Proposal non Proposal Jni tial TMI USI
/\\ct ion Screeninq Plan Section 11.C.l 11.C.2 11.C.4 and Seismic Qualif i-ca t1 on Issue inuncdiately b el4 Candidate 11.11.0
~
~
~*....
~0-PO ITNT I /\\L ISSUE llurln!J a recent foedwater tt*ansicnt at Oconee Unit J all electrical power was temporarily lost to the inte9rated control system.
The power loss was clue to the failure of a non-safety grade inverter power supply to trans-fer automatically from the O.C. power source to the regulated /\\C power source.
I\\ 11 RCS indicators and recorders (except one wi derange RCS pressure recorder) were also lost.
The ability of the operator to be aware of and to adequately control temperature and pressure arc of concern.
Jn this case, all parameters were powered from a single non-safety grade bus.
The licensee is installing a redundant transfer switch, but this does not resolve the pro-blem of all parameters being powered from a single bus nor the question of whether these parameters should in fact be on a safety grade bus.
In November 1970, we issued a 9eneric let'ter on containment purging during normal plant operations (copy attached).
Two earlier events, subsequently classified os abnormal occurrences, involving the manual override of isola-tion signals required to automatically close containment purge valves, also resulted in the safety signals for containment isolation in the event of an accident to be defeated.
The staff's position is detailed on page 5 of the letter.
Responses from licensees indicate that this issue will not be easily or quickly resolved.
Therefore, in October 1979 1 we transmitted a second letter (copy attached) that requested that licensees cor11nit to an interim position as dcscrihccl in the enclosure to th.it letter.
Oefore the lon!J tenn resolution to this issue can be finnlizecl.,several ques-tions such as the advisabi l tty of a complete restriction on containment pttr!Jinl) durin!J plant opcr,1t ion, need to L>e answered.
Source OOrt Proposal DOR Proposal Initial TMI USI
/\\ct ion Sc re cn.:....i'-11~9 __ 1--'-P-'l a 11 Sc c ti on 11.C.l 11.C.2 11.C.3 11.F.tl 11.F.5 11.E.4
~-------~----------*-----------------*!----- -----~*~-----
mm,Jct Pump Integrity (Mcmo-t1ichi1elson to Denton, 5/23/00)
O/\\EOP Candidate ttcaclor Vessel Supports Fracture Tou9hncss DC Power System Rel iahil ity OOR Pro1>osa 1 Last year Candidate Candida tc May be incl udcd t 11.C.l, 11.C.2
~~ 1LU~J_!.,_~ *-
0/\\[0IJ f.i11Hlidi\\te f.0111hined \\'lflh r.elc
/\\l111r1*;pl11!ric ll11111p Valvr~ Opcllill!J on Los~; of' l'owr!r to Jf.S (M1~1110-Mich.iclson to i*.r*"" "",111
.ii<
- '~ *.
y..
r.*.
- r.
yt-.. :***
t ENCLOSURE 4 SU~ARY DISCUSSION OF CANDIDATE ISSUES NOT RECOM.V.ENDED AS UNRESOL VEO SAFETY ISSUES Of the issues that passed the initial screening criteria, most were not.
recorrmended for designation as Unresolved Safety Issues. Certain of these issues were identified as requiring further study to better define the safety concern or to assess their safety significance in order to detennine whether they should be designated as Unresolved Safety Issues. The issues reco~nded for further study are listed below.
Furt~er Study
- 1.
Reliance on ECCS (II.E.2.1)
- 2. *In-Situ Testing of Valves (II.E.6.1)
- 3.
Protective Dev1ce Reliability
- 4.
D.C. Power System Re.liability
- 5.
?WR Pipe Cracks
- 6.
Sl~R Main Steam Isolation Valve Leakage Control Systems
- 7.
Radiation Effects on Reactor Vessel Supports The following provides a sumr~ry discussion of the candidate issues not recoi:'iiienced for report1ng as Unresolved Safety Issues, including those requiring further study, and the bases for these conclusions.
The numbers in parenthesis following titles of certain of the issues refer to the corresponding action item in the iMI Action ?lan (NUREG 0660).
The issues are listed in the order that they ap;::>ear in the candidate 1ssue list of Enclosure 3.
r *::."
f'
- .* ~*. '* Training and Qualifications of Operations Personnel (Item I.A.2.2)
This issue involves a short term potential improvement in the training and qualifications of operations personnel, other than plant operators, and in-cludes maintenance and technical personnel.
Human error in the performance of plant operations can dominate the unavailability of plant equipment.
Such errors, however, appear to be more a result of poor procedures, administra-tive controls and communications than a significant deficiency in the training and qualification of operations personnel. In any case this involves a short tenn improvement by licensees with limited guidance from NRC.
The development of firm guidelines in this area is included in an issue that is recommended for designation as an Unresolved Safety Issue "Long Tenn Upgrading of Training and Qualifications of Operating Personnel" {Item I.A.2.6).
Because a large reduction is not likely to result from these short term improve~ents alone, this issue has not been recommended for designation as an Unresolved Safety Issue.
. long Term Training Simulator Upgrade (Item I.A.4.2)
This issue involves a potential improvement in operator performance by up-grading the capabilities ~f training simulators to include programing of WASH 1400 accident sequences and adding capability to test operator diagnos-tic capability.
The issue is not recommended for designation as an Unresolved Safety Issue because significant improvements in operator performance will be obtained by resolution of the proposed Unresolved Safety Issues on'long Term Upgrading of Training and Qualificati6~s of Operating Personnel"and on 110perating Procedures"(see Enclosure 1) and by improvements in areas where requirements have already been established and implementation is underway or planned, such as the short tenn simulator improvement (Item I.A.4.1), use of a shift technical advisor (Item I.A.l.l),
improverrents in shift manning (Item I.A.1.3), and improvements in
.. operating procedures (Item I.C). The short term simulator upgrade includes: establishing and sustaining a higher degree of realism in training using simulators including dealing with transients; modeling saturation conditions; progra!l":i'!i ng rnu1tipl e failure accident sequences' incorrect instrUiilent responses, and active and passive failures; and including training on natural circulation operation under solid water conditions. These actions already taken or unden-tay will provide the significant improverrcnt in operator performance; the long term simulator upgrade is expected to provide refinements on the actions previously -
taken and, although these will provide some improvement in operator per-formance, and will be implemented as part of the NRC Action Plan (NUR~G 0660), it is not expected to be large enough to designate this issue as an Unresolved Safety Issue.
Requirements for Ooerator Fitness (Item I.A.3.3)
This issue involves a potential deficiency with respect to lack of qualification criteria to screen out individuals with a poor ability to perfoliil under stress or that have dependencies on alcohol or drugs.
The issue is not recOl'l'r.ended for designation as an Unresolved Safety Issue because:
(1} there is no evidence of any significant problem in this area; (2} additional operators and use of a shift technical advisor will reduce*the likelihood of operator err~rs due to this deficiency; (3) operator errors would*generally on1y result in interrup-tion of some equipment but not necessarily a failure of the equipment; (4} alarr.s and indicators would warn the other operators of conditions, resulting from the error, that require corrections; and (5) corrective
').
action may be taken by the operator or other operators to restore the safety function. Resolution of this issue w111 provide some safety improvement and will be imple~ented as part of the NRC Action Plan (NUREG 0660), but it does ~ot warrant designation as an Unresolved Safety Issue.
Oroanization and Managenent - Long Term Improvements (Item I.B.1.1)
This issue involves a potential improvement in plant organizations and
.rnanagenent regarding their capability to assure safe plant operation and to respond to e:nergencies. This issue has not been recOlilTlended for designation as an Unresolved Safety Issue because it is judged that addi-tional upgrading in this area beyond that already implenented is not likely to result in a large reduction in risk.
The first line of defense in an accident situation are the operators who are directly supported by a Shift Technical Advisor. Upgrading in operator training, qualifications, procedures, etc. have already been made and more are planned.
In addition, interim upgrading of plant emergency organizations and mana~ement have already been implemented. Resolution of this issue will provide some improvement in licensee response to emergencies, but does not warrant designation as an Unresolved Safety Issue.
Exoand QA List (Item ?.F.1)
This issue involves a potential i~prove~ent related to application of 10 CFR 50 Appendix a QA criteria to systems and corr.ponents that in the past have not been considered safety related. Such equipment would 1nclude balance of plant equipment that could perform a safety function or whose failure could place demands on safety related equipment.
The
. :..*. ~
F~. ::*:f :*~.. :*
jJ issue is not recommended for designation as an Unresolved Safety Issue because:
{1} Many.of the criteria for the electrical equipr.:ent are being established under Action Plan item II.F.5, and II.F.5 and was found to not warrant designation as an Unresolved Safety Issue; (2} application of QA criteria to this balance-of-plant equipment will not provide a large improvement in reliability of the equipment; and (3) license~s already place importance on reliability of such equip~ent because of ecor.c~ic considerations.
Note that implementation of this issue will rely en the results of other ongoing studies such as IREP and Systems Interaction.
Develoo More Detailed QA Criteria (Item I.F.2)
This issue involves a potential deficiency related to insufficient detail bei~g specified in certain QA criteria. Such lack of detail could lead to incorrect interpretation of how to satisfy the intent of the QA criteria.
The issue is not recorrrnended for designation as an Unresolved Safety Iss~e because:
(1) experience has shown that in general licensees and applicants have been making a conscientious effort to satisfy the intent of the Q.'\\ criteria; (2) even if the intent is not satisfied in a particular area, ct~er QA measures such as technical specification surveillance will provide oiear.s to prevent and detect faults in equipment; and (3} the potential deficiency involves developing further detail in only a limited nurrber of areas.
Although resolution of this issue will provide some im~roveT-ent in application of QA criteria, it does not warrant designation as an Unresolved Safety Issue.
Preoperational and Low Power Testina-Tra ining Reouire:.:ents ( Ite::i I. S. 1)
This issue involves a potential improvement in operator training by re~uiring "hands on" training during low power test programs.
The issue has not ~een reccrrrnended for designation as an Unresolved Safety Issue because it is not likely that such training can significantly improve the operator's ability to respond to potentially serious accidents.
Classroom and simulator training are better able to provide such training. Although resolution of this issue may provide some improvement in safety, it does not warrant designation as an Unresolved Safety Issue.
Reliance on ECCS (Item II.E.2.1}
This issue involves a potential deficiency in the reliability of Emergency Core Cooling Systems.
The concern results from a higher than anticipated frequency of ECCS challenges in operating reactors, due in part to reliance on ECCS for other than loss-of-coolant accidents.
The reliability of ECCS is believed to be high, but it is not clear that it is sufficiently high !o accomplish its safety function with high assurance considering the increase in expected challenges.
Further study is recommended to determine if this issue should be reported as an Unresolved Safety Issue.
The further study would be in the form of scoping calculations related to ECCS challenges and system reliability.
Erneroenc Core Coolino S stem - Uncertainties in Perfor~ance Predictions Item II.E.2.3 ihis issue involves potential uncertainties in small break ECCS performance evaluations as a result of uncertainties due to modeling assumptions or inaccuracies. The issue has not been recorrmended as an Unresolved Safety Issue because small break analyses are believed to be conservative. Resolu-tion of this issue is needed to confir.n the adequacy of the existing analyses.
~. -* ":"..
- . Containri:ent Design - Integrity Check (Item II.E.4.3)
This issue involves a potential improvement related to developing a method to verify gross integrity of the containment structure. Containment inte-grity is presently verified by monitoring the integrity of components (valves, penetrations, etc.) and by administrative controls on valve posi-tions and seal integrity. The issue has not been recommended for designa-tion as an Unresolved Safety Issue because monitoring or periodically veri-fying gross integrity is expected to only provide marginal improvement over current practice. Study of the feasibility, need, and possible methods for such testing will be carried out as part of the NRC's Action Plan (NUREG 0660); however, it does not warrant designation as an *unresolved Safety Issue.
Design Sensitivity of B&W Reactors (Item II.E.5.1)
This issue involves a potential improvement that might be achieved by modi-f ications in systems or procedures to reduce B&W reactor sensitivity to transients. Under Item II.E.5.2 of the NRC's Action Plan recol'TlTilendations were made, based on a short ter:n study, on improvements that should be made to reduce B&W reactor sensitivity to transients. These recommendations are contained in NUREG 0667, "Transient Response of Babcock and Wilcox Designed Reactors," May 1980.
Because iWREG-0667 has been issued, item II.E.5.2 was screened out from being a candidate issue as noted in Enclosure 3.
Item II.E.5.1 of the Action Plan involves a longer ter:n evaluation of B&W reactor sensitivity and identification of any further reco~Jnended improvements to those identified in NUREG 0667.
The staff does not believe that this longer-tenn study will result in significant improvements beyond
r *.*. those resulting from NUREG 0667 (Item II.E.5.2). The results of the long-term evaluation are expected to confirm the adequacy of the changes resulting from NUREG 0667. Accordingly, this issue is not recor.:7'ended for designation as an Unresolved Safety Issue.
In-Situ Testing of Valves (Item II.E.6.1)
This issue involves a potential improvement that might be achieved by demon-strating the functional perfonnance of valves in Engineered Safety Feature Systems.
In-servi~e testing and technical specification surveillance provides* some measure of the operability of valves.
However, these tests are not performed under the same loadings and conditions that the valve may experience in an accident or emergency situation. A valve reliability study, based on test and operational data, indicates valve reliability is about the same as was estimated in WASH-1400.
However, a current study of valve test frequency and furt~er consideration of testing valves under severe condi-tions may indicate a potential for risk reduction greater than currently anticipated. Accordingly, further study is recorrnnended to estimate if the test adequacy study to be performed under Item II.E.6.1 is likely to result in:
(1) a significant irr.provement in valve reliability by changing the test frequency, and (2) a significant reduction in risk if methods were developed for testing of valves closer to the design conditions.
- '-:.. **. Study of Control and Protective Action Design Reouire:nents (Ite~ II.F.4).
This issue involves a potential deficiency related to:
(1) basing protec-tive actions on derived variables rather than direct reading of process variables;. (2) protective actions relying on coincidence of independent process variables rather than relying on either variable; and (3) lack of testing of control circuit components at expected degraded power supply conditions.
It is believed that existing requirements alrF.~~Y preclude these deficiencies. The issue is not recommended for designation as an Unresolved Safety Issue because the recorr.mended action involves add~n9 further clarification to existing requirements in the Standard Review Plan and only minor improvement in protection is expected to resuit.
C1assification of Instrumentation, Control and Electrical Equi~ent (It~~ II.F.S)
This issue involves a potential improvement by developing a standard for establishing design criteria and performance requirements for instrumentation, control and electrical equipment in accordance with the equiµnent's safety importance. ihis would likely result in upgraded requirenents for sc~e equip-ment.
The current classification scheme (Class IE) is judged to provide reasonably good criteria for many systems and components important to safety.
Additionally, Revision 2 to Regulatory Guide l.97, "Instrumentation to Follow The Course Of An Accident" has been developed and issued for public corrrnent.
Although development of an improved classification scheme could improve the reliabiiity and performance of some equipment, and will be dcne in accordance with the NRC's Action Plan, the issue is not recorr.":lencied for reporting as an Unresolved Safety Issue because the reduction in risk is not expected to be large from this. improvement.
N.':\\C E;..:rcenc Preoarediiess - Trainin, Drills, and Tests Item 111.A.3.5 This issue involves a potential i~prove~ent in NRC e~ergency preparedness through NRC observation and evaluation of joint exercises between the licensees, State and local age~cies, and Federal response organizations (including FHt:..).
The issue is not recor..1iended for designation as an Unresolved Safety Issue because:
(l) the significant reduction in risk resulting fro~
ir.;prove:-::ents in the area of emergency preparedness at each NRC licensed nuclear facility will evolve from an intensive NRC program to upgrade e~~rgency pre-paredness and issuance of upgraded emergency preparedness re~ulations, the criteria and requirements have already been established for these rules; and (2) the NRC has develo~ed a program for the improve~ent of NRC emergency pre-paredness that enco~passes far more than just observation of joint exercises.
The importance of observation of joint exercises should not be down-played, but they will provide only a small impact on the improvement of N~C emergency pre;ia redness.
Radiation Source Contro1-Ventilation System and Radioiodine Adsorber Criteria (Item III.0.1.3}.
This issue is a potential deficiency in means to control and process air;,orne radioactivity in the auxiliary and radwaste buildings and in ffiaintaining filter r.dia efficiency. Operating experience and research have identified certain areas where charcoal filter efficiency and use may be improved; further researc1 is planned. The issue is not recor.rr.ended for designation as an Unresolved
. Safety !ssue because;
{1) existing criteria on charcoal filters are believe::l to be generally quite good, with perhaps only minor changes require~
based en operating experience and research; and (2) the improve~er.ts that will be made are not expected to result in a significant reduction in risk.
It should be noted that the rule change under item II.B.8 of the T!~I Action Plan r.dy also require further changes to ventilation syste~ filtration capability; the i~~=ct of these further changes is not considered under this issue.
Inplant Radiation Monitoring (Item III.0.3.3)
This issue involves a potential improvement that might be achieved by in-creased in-plant radiation monitoring capability, including installation of radiation monitors with reo.ote readout, high dose rate readout instru-ments, and additional portable radiation monitoring equipment.
This issue is not recor.u~ended for designation as an Unresolved Safety Issue because the increase in radiation monitoring equip~ent will provide only an incremental improvement in reducing dose to plant personnel for postulated accidents bey~nd the protection provided by present monitoring capability.
Evaluate Elimination of pcqv Function (Item II.K.3(33))
This issue involves a potential improve~ent that might be achieved by ei:her reducing demands on the PORV (revising set-points) or providing an improved means to cope with a stuck open PORV (automatic operation of the PORV block valve).
The issue is not recor..mended for designation as an Unresoived Safety Issue because:
(1) a study by the Probabilistic Analysis Staff has indicated that th~se improvements would not significantly reduce the potential for core dar;iage; (2) licensees have already been required to provide improved methods of indication to the operator of a stuck-open PORV; and (3) changes to operator training and eme~gency procedures are being made so that the operator is better able to cope with a stuck open PORV.
Reliability of Ventilation Monitoring Ecuipment This issue involves a potential deficiency related to low reliability of air flow monitoring equipment.
The issue was identified in the ACRS report on Licensee Event Reports, NUREG-C5i2.
The issue is not recc;;-;;:e:iced for designation as an Unresolved Safety Issue because loss of tr.e air ~cnit:r~
ing equip~ent, in itself, will not cause loss of a safety function.
Loss of a safety function would also require the follc~ing to occur:
a failure in the ventilation equip~ent, a failure in the plant cperatcr shift tours, a fa i 1 ure in the redundant syste:.i to that aff:cted by the ventilation failure, and failures in any non-safety syste~s tr.at could perfoliil the safety function.
Protective Device Reliability This issue involves a potential deficiency in that high failure rates of protective devices (fuses, circuit breakers), could result in a lo~er reliability of safety equip~ent. The issue was identified in t~e ACRS report on Licensee Event Reports (LER's) NUR~G-0572. This report noted an apparent large nur.;bc:r of LER's re1ated to failures of protecth*e devices.
Such failures will result in unavailability of the related safety equi~ent. However, it was not known whether the failure rate of safety equi~~ent mi5ht be greater than previously assumed (in ~ASH-14CO or other reiiability reports) as a result of protective device faiiur:s, o:
if the reliability of safety equipment could be significar.tly ir.:proved ~Y increasing protective device reliability. Accordingly, further st~dy has been recol7Iiiiended to estimate protective device failure rate ar.d to dete;.;iine (1) if this failure rate is excessive.and leads to a lowering of failur~
~
.... 1 (2).f.&.'
- 'l ra~e es~1~a~es.or essen~1a equipment, or i
~ne ra1 ure ra~e es~1;.:a~es for essential equipr::ent could be significantly rect.:ced by i;.:provin; pro:ec-tive device reliability. Such a study should be conducted before a ce:isfon is made on whether this iss~e should be designated as an Unres~lvec Se~e~y Issue.
I;
~-~ *.-
t*., Instr~7.:e~t2tion Set-Point Drift This issue involves a potential deficiency related to an excessive drift in instrumentation set points beyond Technical Specification limits.
The issue was identified in the ACRS report on Licensee Event Reports,
- uRE:G-0572.
The issue is not recor.;;:;enced for desisnation* as an l!nresolve~
Safety Issue because:
(1) given the set point drift, the affected c~ar.~el in most cases would be only slightly out of tolerance and therefore would tr~p at close to the desired setting; (2) other channels would senera11y be av2ilable; an_d (3) in most cases operators may take lii2.nua1 actions to acco~plish the safety function.
End-of-Life and ~aintenance Criteria This issue involves a potential deficiency related to lack.of adeq:.ia:e criteria for establishing maintenance periods and end-of-life ex~e:tan:y fer m2terials that may degrade significantly with tir.e or use.
The issue was identified in the ACRS report on Licensee Event Reports, NUREG 0572.
The issue is not recomrr.ended for designation as an Unresolved Safety Issue because: it is not likely that failures due to this deficiency will occur simultaneously in redundant systems; periodic testing of equip~ent and
~n-service inspection will detect such degradaticn; and requirements are being established for certain identified issues related to material de~racaticn and these have been previously designated as Unresolved Safety Iss~es,
(.!.-3, A-4, A-5) Stea:n-Generator lube Integrity and (A-11) Reactor Vessel Materials Toughness.
Although establishment of such criteria ~ay provi~e some irr.prove~ent in safety. it does not warrant designation as an Un~esolved Safety Issue.
- \\
-,~ -
Desion Check and Audit of Balance-of-Plant Eaui2~ent This issue involves a potential improve::ient that might be achieved by requirements for verification that the balance-of-plant "as built" con-figuration satisfies the design intent. Such action could improve the reliability of balance-cf-plant equipment and reduce d;;:-:ands on safety equipment.
The issue was identified in the ACRS report on Licensee Event Reports, NUREG 0572.
The issue is not r~cornmended for designation as an Unresolved Safety Issue because transients or safety systems challenges result more frequently from opera tor errors and "random" cc:::ponent fa i 1 ures rather than from deviations from the intended plant design. Additionally other ongoing studies (!REP and Systems Interactions) wil 1 identify poten-tial adverse impacts from balance-of-plant equipment.
BWR Control Red Worth This issue involves a potential deficiency in the manner of account~ng for xencn following a reactor trip in assessing its effect on control rod worth.
The issue was identified in the ACRS report on Licensee Event Repor~s, NUREG 0572.
The issue is not recommended for designa!ion as an Unresolved Safety Issue because an initial staff review has found that the rod worths are relatively insensitive to xenon distribution.
The results of the final staff review wi11 be documented when the review is co~pleted.
Flow Induced Vibration The issue involves the potential for single and multiple failures of piping, valves, snubbers, and nearby electrical and mechanical cc;;:~onents as a result of flow-induced vibrations.
The issue was identified in the ACRS report on Licensee Event Reports, NUREG-0572.
T~e issue is not reco~~ended for designation as an Unresolved Safety Issue because:
(1) failures in electrical equip~ent as a result of flow-induced vibrations would likely occur in only one safety division because of the physical separaticn of _redundant co~ponents to satisfy other regulatory requirements s~ch as flooding and fire protection; (2) failures in mechanical components are likely to be in piping restraints or supports rather than in.the piping, and as such would not likely result in loss of a safety function; and (3) many sources of flow induced vibration failures noted in LER's have been identified and corrected, such as p~"i':p high cycle fatigue and reactor in-ternal vibration.
Inadvertent Actuation of Safety Injection This issue involves a potential tendency for operators to ter~inate safety injection when actually required, because their judgement has been influenced by the large number of inadvertent safety injections that have occ~rred in the past.
The issue was identified in the ACRS report on Licensee ~vent Reports, NUREG-0572.
The issue is not recommended for designation as an Unresolved Safety Issue because improvements in training and procedures, subsequent to TMI, related to safety injection operation have stressed the need for operators to obtain multiple indications to determine if actuation was inadvertent.
Further evaluations of ECCS challenge frequency will be perfor:ned under Action Plan Item II.E.2.1, currently designated for further study.
(
Reevaluation of Reactor Coolant Pump Trip Criteria (Item II.K.3(5))
The issue involves a potential improvement that might be achieved by establish-ing better criteria on when to allow reactor coolant pump operation and when to trip the pumps.
Better criteria might allow use of reactor coolant pumps to aid in recovering from certain transients, while still assuring that these pu~ps are tripped for a small break LOCA.
The issue was iden-tified by the ACRS in its letter of March 11, 1980 concerning recommenda-tions of the NRC Task Force on Bulletins and Orders.
A study is currently ongoing to resolve this issue and is expected to be completed in early 1981.
The issue is not recommended for designation as an Unresolved Safety Issue because, although such use of reactor coolant pumps would provide a small improvement in safety, the safety evaluations of transients used for.licensing acceptability do not assume the availability of the reactor coo 1 ant pumps.
Turbine Disk Crackina This issue involves a potential deficiency related to turbine disk inte-grity that could increase the likelihood of a severe accident from turbine missiles. This issue has recently been raised because of the discovery of stress corrosion cracking in low pressure turbine disks.
The issue has not been recoliTTlended for designati.on as an Unresolved Safety Issue because the probability of turbine missile generation assumed by the staff in acci-dent calculations has been unaffected ~Y the discovery of these cracks.
Requirements for periodic inspection of the disks may actually decrease this probability.
In addition, a serious accident is not considered likely even if the affected disks fail because to date the cracking has only been cs-served in the smaller, lower energy disks.
Missiles generated from these disks would not likely escape the turbine casing or penetrate other struc-tural barriers.
DC Power System Reliability This issue involves a potential improvement related to DC power system reliability. The issue was originally raised by an ACRS consultant.
The NRC staff has funded a contractor study of the probability of core damage following certain transients as a result of loss of shutdown cooling from DC power system failures.
The DC power system analyzed in the study was a DC system meeting the staff's current minimum require-ments.
The preliminary results indicate that relatively significant reductions in the probability of core dar.Gge from event sequences in-volving DC power system failures could be realized by making certain improve-ments in the DC power system design.
However, such improvements would have little effect on the overall core damage probability unless other improve-ments in the shutdo'l-m cooling system design are made.
Further it is r.ot clear how typical the DC power system analyzed is compared to those found at most plants. It is thought likely that most actual installations exceed the staff's minimum requirements.
Accordingly, further study *is recol'17i1ended to detennine if this issue should be designated as an Unresolved Safety Issue.
BWR Jet Pump Integrity This issue involves the potential for degraded core cooling as a result of jet pump failure that occurs because of a large LOCA and degraded structural jet pump members.
Failure could potentially result if jet pump structural members were cracked during normal service by water hai'i:iier events prior to the LOCA, or result from flow induced vibration caused by ECCS flow follow-ing a LOCA.
This issue was the subject of a mer:lorandum from C. Michelson
. -~..
- Radiation Effects on Reactor Vessel Supoorts This issue involves a potential deficiency in reactor vessel supports related to a reduced fracture resistance as a result of irradiat1on damage from low energy neutrons. Although the consequences of reactor vessel support failure under large loads such as LOCA or earthquake loads could be severe, there are a number of uncertainties regarding the likeli-hood for low support fracture resistance. Further study to better charac-terize the support materials, the neutron spectra, the potential radiation damage and the structural loading of supports is recor.7ilended before making
/....
a judsment regarding whether this issue should be designated as an Unresolved Safety Issue.
Loss of Offsite Power Subsequent to a LOCA This involves a potential improve~ent that might be achieved if it were required to consider loss of offsite power subsequent to a LOCA in the plant design. This issue has not been recorrmended for designation as an Unresolved Safety Issue because the probability of the combined event is judged to be very 1 ow (on the order of 1 o*6 /RY) and the consequences would likely be insignificant, because adequate core coolirig would be provided by vessel inventory during the time required (less than 1 minute ) for diesels to start and assume load.
OELD FILE COPY.
July 14, 1980 CORRECTION NOTICE To All Copyholders of SECY-80-325 DELD FILE COP'D Attached are two pages, pps. 17 and 18, which were omitted from of the subject paper.
Please attach them to your copy.
Attachment:
pages 17 and 18 The Secretariat
to H. Denton dated May 23, 1980, and work has been unden*1ay in NRR on this subject since the first reports of degraded structural me~bers. The issue is not recommended for designation as an Unresolved Safety Issue because the occurrence of degraded core cooling would require the combination of:
(1) a large LOCA; (2) a degr~ded jet pu~p; and (3) a*jet pump failure that results in inadequate core cooling.
The likelihood of this combina-tion of events is judged to be low.
Small Break LOCA from Extended Overheating of Pressurizer Heaters This issue involves the potential for failure 6f the pressurizer pressure boundary in the event of extended overheating of the pressurizer heaters.
The issue has not oeen recommended for designation as an Unresolved Safety Issue because the possible scenarios involved multiple equip~ent failures and operator inaction for relatively long time periods.
Such scenarios are judged to be of low likelihood.
PWR Pipe Cracks This issue involves a potential deficiency in plant equipment related to cracking in various PWR piping systems.
The prin:ipal causes of cracking have been thennal fatigue, vibration induced fatigue and intergranu1ar stress corrosion cracking.
This issue has been the subject of a recent investigation by the Pipe Crack Study Group. Although thennal fatigue cracking has been observed in a number of feedwater lines, analyses indi-cate th3t such cracking is not likely to result in complete severence of the line even when severely loaded.
Complete severence of small ("-3/4 11
)
vent or drai~ lines in certain locations in ~mergency core cooling systems
as a result of vibration induced fatigue could potentially result in degraded core cooling.
However it is not evident that the particular scenarios envisioned are likely enough to involve a significant contribu-
. tion to risk. Accordingly, further study has been recom:nended to determine if this issue should be designated as an Unresolved Safety Issue.
BWR Main Steam Isolation Valve Leakage Control Systems This issue involves a potential deficiency in the ability to control leakage through the main steam isolation valves (MSIVs) in BWR plants.
As a result of excessive leakage experience for the MSIV~ in operating plants, the staff developed requirements for HSIV leakage control systems, as des-cri~ed in Regulatory Guide 1.96. However, the initial operating experience with the leakage control systems suggests that they are also prone to failures.
In addition, it appears from recent leakag2 test results that there are improved maintenance procedures that may significantly reduce esxcessive leakage from the MSIVS.
Accordingly, further study hc.s been reco!T."7lended to estir.~te the MSIV and leakage control system failure rates and to determine if the leakage control system failure rate is exces-si~e in order to determine whether the issue should be designated as an Unresolved Safety Issue.