ML19261B387
| ML19261B387 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0510, NUREG-510, NUDOCS 7902210066 | |
| Download: ML19261B387 (95) | |
Text
NUREG4510 IDENTIFICATION OF UNRESOLVED SAFETY ISSUES RELATING TO NUCLEAR POWER PLANTS I
Report to Congress y,
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Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 790221006 G I
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National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $6.00 ; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Infonnation Service.
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NUREG-0510 IDENTIFICATION OF UNRESOLVED SAFETY ISSUES RELATING TO NUCLEAR POWER PLANTS Report to Congress Manuscript Completed: January 1979 Date Published: January 1979 Program Support Branch Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555
TABLE OF CONTENTS PAGE INTRODUCTION AND
SUMMARY
1 BACKGROUND...
IDENTIFICATION OF " UNRESOLVED SAFETY ISSUES" t
APPENDICES APPENDIX A -
1978 Annual Report Text on " Unresolved Safety Issues" APPENDIX B -
Generic Issue Grouping by Type of Activity APPENDIX C -
Risk-Based Categorization of Generic Tasks APPENDIX D -
Events Reported to Congress as Abnormal Occurrences
INTRODUCTION AND
SUMMARY
In December of 1977, the Energy Reorganization Act of 1974 was amended to include a new Section 210 which required that the Nuclear Regulatory Commission develop and submit to the Congress a plan for the specification and analysis of " Unresolved Safety Issues" relating to nuclear reactors.
The plan was to be submitted on or before January 1, 1978 and progress reports are to be included in the Annual Report of the Commission thereafter.
In accordance with the new Section 210, a report, NUREG-0410, entitled "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," was submitted to the Congress on December 30, 1977 identifying 133 generic tasks and describing the NRC program that was then already in place.
The report, however, pointed out that the NRC program was consider-ably broader than the " Unresolved Safety Issues" plan required by Section 210.
The NRC program, for instance, included plans for the resolution of generic environmental issues as well as plans for the resolution of issues related to nuclear power plant safety.
In the letter transmitting NUREG-0410 to the Congress on December 30, 1977, the Commission indicated that "the progress reports, which are required by Section 210 to be included in future NRC annual reports, may be more useful to Congress if they focus on the specific Section 210 safety items."
It is the NRC's view that the intent of Section 210 is to assure that plans are developed and implementeu on issues with potentially significant public safety implications.
Over the past year, the NRC has undertaken a review of the generic issues addressed in the NRC program described last year to determine which issues fit this description and qu:lify as
" Unresolved Safety Issues" for reporting to the Congress in the NRC Annual Report.
The NRC review included the development of proposals by the NRC staff and review and final approval by the NRC Commissioners.
As a result of this review, 17 " Unresolved Safety Issues" addressed by 22 tasks in the NRC program were identified.
The issues are listed below and the progress on these issues is discussed in the 1978 NRC Annual Report.l The number (s) of the generic tasks (e.g., A-1) in the NRC program addressing each issue is indicated in parentheses following the title.
UNRESOLVED SAFETY ISSUES 1.
Water Hammer (A-1) 2.
Asymmetric Blowdown Loads on the Reactor Coolant System (A-2) 3.
Pressurized Water Reactor Steam Generator Tube Integrity (A-3, A-4, A-5) 4.
BWR Mark I and Mark II Pressure Suppression Containments (A-6, A-7, A-8, A-39) 5.
Anticipated Transients Without Scram (A-9) 6.
BWR Nozzle Cracking (A-10) 7.
Reactor Vessel Materials Toughness (A-11) 1P e text of the 1978 NRC Annual Report section addressing " Unresolved Safety Issues" is included in this report as Appendix A.
_ 8.
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (A-12) 9.
System Interactions in Nuclear Power Plants (A-17) 10.
Environmental Qualification of Safety-Related Electrical Equipment (A-24) 11.
Reactor Vessel Pressure Transient Protection (A-26) 12.
Residual Heat Removal Requirements (A-31) 13.
Control of Heavy Loads Near Spent Fuel (A-36) 14.
Seismic Design Criteria (A-40) 15.
Pipe Cracks in Boiling Water Reactors (A-42) 16.
Containment Emergency Sump Reliability (A-43) 17.
Station Blackout (A-44)
This NUREG report describes the review undertaken over the last year that resulted in identifying the 17 issues listed above as " Unresolved Safety Issues."
In addition, the report provides specific discussions of why certain issues were not included.
The report also provides a brief back-ground discussion describing Section 210 of the Energy Reorganization Act and the NRC program for the resolution of generic issues described in NUREG-0410 last year.
BACKGROUND The NRC staff continuously evaluates the safety requirements used in its reviews against new information as it becomes available.
Information related to the safety of nuclear power plants comes from a variety of sources, including experience from operating reactors, research results, NRC staff and Advisory Committee on Reactor Safeguards safety reviews, and vendor, architect / engineer and utility design reviews.
As new concerns or safety issues are identified from one or more of these sources, the need w
for immediate action to assure safe operation is assessed.
This assess-ment includes consideration of the generic implications of the issue.
In some cases, immediate action is taken to assure safety, e.g., the derating of boiling water reactors as a result of the channel box wear problems in 1975.
In other cases, interim measures, such as modifications to operating procedures, may be sufficient to allow further study of the issue prior to making licensing decisions.
In most cases, however, the initial assessment indicates that immediate licensing actions or changes in licensing criteria are not necessary.
In any event, further study may be deemed appropriate to make judgments as to whether existing NRC staff requirements should be modified to address the issue for new plants or if backfitting is appropriate for the long-term operation of plants already under construction or in operation.
These issues are sometimes called " generic safety issues" because they are related to a particular class or type of nuclear facility rather than a specific plant.
These issues have also been referred to as "unrec ilved safety issues." However, as discussed above, such issues are considered on a generic basis only after the staff has made an initial assessment for individual plants and has made a determination that the safety signifi-cance of the issue does not prohibit continued operation or require licensing actions while the longer term generic review is underway.
- As a result of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-20S' on December 13, 1977 to include, among other things, a new Section 210 as follows:
" UNRESOLVED SAFETY ISSUES PLAN" "SEC. 210.
The Commission shall develop a plan providing for specification and analysis of unresolved safety issues relating to nuclear reactors and shall take such actions as may be necessary to implement corrective measures with respect to such issues.
Such plan shall be submitted to the Congress on or before January 1, 1978 and progress reports shall be included in the annual report of the Commission thereaf ter."
The Joint Explanatory Statement of the House-Senate Conference Committee for the FY 1978 Appropriations Bill (Bill S.ll31) provided the following additional information regarding the Committee's deliberations on this portion of the bill:
"SECTION 3 - UNRE5OLVED SAFETY ISSUES" "The House amendment required development of a plan to resolve generic safety issues.
The conferees agreed to a requirement that a plan be submitted to the Congress on or before January 1, 1978.
The conferees also expressed the intent that this plan should identify and describe those safety issues, relating to nuclear power reactors, which are unre-solved on the date of enactment.
It should set forth:
(1) Commission actions taken directly or indirectly to develop and implement corrective measures; (2) further actions planned concerning such measures; and (3) timetables and cost estimates of such actions.
The Commission should indicate the priority it has assigned to each issue, and the basis on which priorities have been assigned."
In response to the reporting requirements of the new Section 210, the NRC staff submitted to Congress on January 1, 1978, r report on the "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," (NUREG-0410).
The NRC program described in NUREG-0410 was already in place when PL 95-209 was enacted and is of considerably broader scope than the " Unresolved Safety Issues Plan" required by Section 210.
Although the NRC program does include plans for the resolution of generic technical issues of varying degrees of safety significance, it also includes generic tasks for the resolution of environmental issues, for the development of improvements in the reactor licensing process, and for consideration of less conservative design criteria or operating limitations in areas where over-ccnservatism may be unnecessarily restrictive.
The major elements of the NRC program are described in NUREG-0410 and are summarized below.
A Technical Activities Steering Committee was established to increase high level management involvement in and to improve management oversight of generic technical activities.
The Steering Committee is chaired by the Deputy Director, Office of Nuclear Reactor Regulation (NRR) and includes, as members, the four Division Directors in NRR.
The Committee's functions include assigning proposed generic tasks to priority categories, assigning lead responsibility to an NRR division for defining and executing each generic task, approving Task Action Plans, and regularly reviewing the progress of ongoing tasks.
w The Steering Committee's judgmental decisions regarding priorities and other mattcrs, such as the assignment of an NRR division with lead responsi-bility and approval of the Task Action Plan for each task, are based upon recommendations resulting from an extensive internal review process.
This process begins in the NRR line organizations through staff development, review, comment and concurrence on proposals for high priority tasks and Task Action Plans.
In addition, specific recommendation regarding these proposals are provided by the Steering Committee's Advisory Group following its detailed review.
The Advisory Group is made up of five senior technical staff representing each of the NRR divisions and the Director, NRR.
Implementation of the program began in 1977 by the Technical Activities Steering Committee's determination of the relative priority of a large number of ongoing, planned or suggested generic efforts.
The generic issues that were considered included those from the Advisory Committee on Reactor Safeguard's listing, those listed in NRR's former Technical Safety Activities Report, the 27 issues discussed in NUREG-0138 and NUREG-0153,2 and a number of other generic issues that were identified from a variety of sources as described above.
The Steering Committee adopted four priority category definitions as descriptive of the various generic technical issues.
These definitions are presented in Table 1.
As indicated by these definitions, issuas were assigned to the various priority categories 2
NUREG-0138 and NUREG-0153 published in November and December 1976, respec-tively, provided the staff's discussion of 27 technical issues identified by one or more members of the NRR staff as problems whose priority, progress or resolution was, in their opinion, unsatisfactory.
based on their judged safety, environmental or safeguards importance, or their potential for improving the efficiency or effectiveness of the licensing process.
Initially, each of the NRR divisions described and proposed to the Technical Activities Steering Committee those generic issues it considered to warrant high priority effort (Category A and Category B tasks).
Proposals were received for over 130 Category A tasks and over 225 Category B tasks in April and May 1977, respectively.
Many of these proposals were duplicates or could be readily combined as a part of another proposed generic issue.
The Steering Committee reviewed the division proposals and the recommenda-tions of its Advisory Group, assigned each task to a priority category and designated an NRR division with lead responsibility (Lead Division) for each task.
Table 2 provides a listing of the issues assigned to Priority Categories A, B, C and D by the Steering Committee from among those issues originally proposed as Category A and B tasks.
A Task Action Plan has been approved by the Technical Activities Steering Committee for each of the Category A tasks listed in Table 2.
Copies of Task Action Plans for the Category A tasks listed in Table 2 are contained in NUREG-0371, " Task Action Plans for Generic Activities, Category A," published in November 1978 and trans-mitted to the appropriate Congressional oversight committees on November 22, 1978.
. Each Task Action Plan in NUREG-0371 provides a description of the problem and the staff's approach to resolution; a general discussion of the bases upon which continued plant licensing or operation can proceed pending completion of the task; technical organizations involved in the review and estimates of the manpower required; a description of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards and outside organizations; estimates of any funding required for contractor supplied technical assistance; prospective dates for completing the task; and a description of any potential problems that could impact the plans.
With regard to Category B, C and D tasks, the NRC staff has compiled a brief description of the Category B, C and D tasks listed in Table 2 in NUREG-0471, " Generic Task Problem Descriptions, Category B, C and D Tasks,"
published in June 1978.
NUREG-0471 was transmitted to the appropriate Congressional oversight committees at the same time as NUREG-0371 on November 22, 1978.
Task Action Plans have not as yet been approved for these lower priority tasks.
IDENTIFICATION OF " UNRESOLVED SAFETY ISSUES" As indicated above, the NRC program for resolution of generic issues is considerably broader than the " Unresolved Safety Issues Plan" required by Sction 210.
In the letter transmitting NUREG-0410 to the Congress on December 30, 1977, the Commission indicated that "the progress reports,
...........---._- - - which are required by Section 210 to be included in future NRC annual reports, may be more useful to Congress if they focus on the specific Section 210 safety items."
It is the NRC's view that the intent of Section 210 is to assure that plans are developed and implemented on issues with potentially significant public safety implications.
Over the past year the NRC has undertaken a review of the generic issues addressed in the NRC program to determine which issues fit this description and qualify as " Unresolved Safety Issues" for reporting to Congress.
Consistent with this view, the following definition of an " Unresolved Safety Issue" was developed:
"An 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 conditions not likely to be acceptable over the lifetime of the plants affected."
In applying this definition, matters that pose "important questions con-cerning the adequacy of existing safety requirements" were judged to be those for which resolution is necessary to (1) compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety.
Quite simply, an " Unresolved Safety Issue" is potentially significant from a public safety standpoint and its resolu-tion is likely to result in NRC action on the affected plants.
... - The issues addressed in the NRC program were evaluated against this defi-nition.
A number of issues received limited consideration as part of a general screening of all of the issues.
The remaining issues were evaluated individually in greater detail.
Several different groupings or categorizations of the issues were utilized in screening the issues.
One set of candidates for " Unresolved Safety Issues" were those issues that had been assigned to Priority Category A.
As indicated by the priority category definitions in Table 1, those generic tasks assigned to Category A are those that had been judged to be the most important, including those judged to be most important in terms of safety significance.
Those issues assigned to Category B, although also important, were originally judged to be of " lesser" safety significance and therefore, Category B issues were not likely to qualify as " Unresolved Safety Issues,"
Similarly, Category C and Category D issues were not likely to qualify.
One additional consideration was, as indicated earlier, that many of the issues in the NRC program are not directly related to safety, e.g.,
some are environmental issues.
Accordingly, the NRC staff grouped the generic tasks in the NRR program in eight groups by activity type.
The definitions of the eight groups are as follows:
Group 1 - Analyzing generic technical problems related to plant safety that have arisen from new information such as experience from operating reactors, identified design or construction deficiencies, or research and test results, for the purpose of (1) determining whether existing safety requirements or review procedures require upgrading and developing new requirements or procedures if needed, and/or (2) developing technical positions regarding the acceptability of long-term solutions for operating plants.
Group 2 - Evaluating existing safety requirements and review procedures in areas related to plant safety that could require upgrading.
Group 3 - Performing studies to confirm the adequacy of current staff safety requirements.
Group 4 - Performing studies for the purpose of quantifying safety margins provided by current requirements or determining whether or not current safety requirements can be relaxed.
Group 5 - Developing, maintaining or improving staff capabilities to perform independent calculations and audits for safety and environmental reviews.
Also performing generic audit reviews and calculations.
Group 6 - Improving guidance to applicants, licensees and/or staff reviewers regarding staff safety and environmental requirements or devel-oping documentation describing the basis for staff safety and environmental requirements.
Group 7 - Performing studies related to staff environmental reviews as necessary to (1) address new information, (2) determine whether or not current staff requirements can be relaxed, and (3) con-firm the adequacy of current staff requirements.
Group 8 - Other tasks that do not fit the definitions of Groups 1 through 7.
The distribution of the Category A, B, C and D generic tasks in these groupings is provided in Appendix B.
As can be seen from the group defi-nitions above, those issues in Groups 1, 2 and 3 should encompass those that could potentially qualify as " Unresolved Safety Issues."
-- An additional source of infoncation was also used in screening the issues.
A preliminary risk-based evaluation of the generic issues in the NRC program had been performed at the request of the Director of the Office of Nuclear Reactor Regulation (NRR) for the purpcse of reevaluating the ongoing and planned efforts in NRR's generic issuts program.
The objec-tive of this reevaluation was to develop a basis for refining priorities for action on the various tasks.
A principal element of this study was the examination of the significance of each planned generic task, based on risk-related criteria.
Drawing on the insights gained from the Reactor Safety Study (WASH-1400) and related follow on efforts, an attempt was made to identify which tasks had, potentially, the greatest relative worth from a risk standpoint.
This generally involved use of tounding (conserva-tive) fault-tree / event-tree evaluations as a scieening device to group the tasks into various categories of potential " payoff." The draft report describing these evaluations indicated that each generic risk was assigned to one of four categories:
Category I:
Potential High Risk ltems; Category II:
Potential Low Risk Items; Category III:
Items With Negligible Risk Potential and Category IV:
Items Not Directly Relevant to Risk.
The distribution of the Category A, B, C and D generic tasks in these four risk-based categories is provided in Appendix C.
From these three somewhat different groupings or categc rizations of the generic tasks in the NRC program, the most likely candidates for "Unre-solved Safety Issues" would be those issues that appeared in all three of the most important groupings from a safety standpoint, i.e., Categor A,
Groups 1, 2 or 3 and the risk-based Categories I or II (15 issues in all).
However, to assure that all issues that could possibly qualify were considered, all issues that were included in either Category A, Groups 1, 2 or 3, or in the risk-based Categories I or II received further, more 3
detailed consideration.
This resulted in the individual review of 86 of the 133 generic tasks in the NRC program.
The 86 tasks considered included all 40 Category A tasks, 33 Category B tasks, 11 Category C tasks, and 2 Category D Tasks.
In addition to considering these 86 generic tasks, one additional source of information was reviewed.
In accordance with Section 208 of the Energy Reorganization Act of 1974, the NRC is required to ".
submit to the Congress each quarter a report listing for that period any Abnormal Occurrences at or associated with any facility which is licensed or other-wise regulated pursuant to the Atomic Energy Act of 1954, as amended, or 3Information considered in the review of each of the 86 tasks included (1) Task Action Plans for Category A tasks (NUREG-0371), (2) Problem descriptions for Category B, C and D tasks (NUREG-0471), (3) the draft risk-based evaluation referred to on page 13 above, and (4) NRR staf f comments on the draf t risk-based evaluation.
pursuant to this Act.
For the purpose of this Section, an Abnormal Occur-rence is an unscheduled incident or event which the Commission determines is significant from the standpoint of public health and safety To make the requisite determination, the NRC applies the criterion promul-gated in an NRC policy statement.
This statement defines an Abnormal Occurrence as an unscheduled incident or event which involves "a major reduction in the degree of protection of the public health or safety.
Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:
"(1) Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission"
"(2) Major degradation of essential safety-related equipment; or
"(3) Major deficiencies in design, construction, use of, or manage-ment controls for licensed facilities or material."
As noted above on page 10, this " major reduction in the degree of protec-tion of the public health and safety" criterion was also utilized in determining which issues qualify as " Unresolved Safety Issues." Because the same criterion is used, a generic issue requiring long-term study that is identified as a result of an incident reported as an Abnormal Occur-rence, would be expected to qualify as " Unresolved Safety Issue." For this reason, the Abnormal Occurrences reported to Congress to-date were reviewed to determine if any issues qualifying as " Unresolved Safety Issues" could be identified.
The list of Abnormal Occurrences considered is included as Appendix D.
The result of the review of the 86 generic tasks referred to above and of the Abnormal Occurrences listed in Appendix D was the identification of 17
" Unresolved Safety Issues" addressed by 22 generic tasks in the NRC program.
The text of the 1978 Annual Report section addressing these issues is provided in Appendix A.
These issues are listed below:
UNRESOLVED SAFETY ISSUES 1.
Water Hammer (A-1) 2.
Asymmetric Blowdown Loads on the Reactor Coolant System (A-2) 3.
Pressurized Water Reactor Steam Generator Tube Integrity (A-3, A-4,A-5) 4.
BWR Mar: I and Mark II Pressure Suppression Containments (A-6, A-7, A-8, A-39) 5.
Anticipated Transients Without Scram (A-9) 6.
BWR Nozzle Cracking (A-10) 7.
Reactor Vessel Materials Toughness (A-11) 8.
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (A-12) 9.
System Interactions in Nuclear Power Plants (A-17) 10.
Environmental Qualification of Safety-Related Electrical Equipment (A-24) 11.
Reactor Vessel Pressure Transient Protection (A-26) 12.
Residual Heal Removal Requirements (A-31) 13.
Control of Heavy Loads Near Spent Fuel (A-36) 14.
Seismic Design Criteria (A-40) 15.
Pipe Cracks in Boiling Water Reactors (A-42) 16.
Containment Emergency Sump Reliability (A-43) 17.
Station Blackout (A-44)
As can be seen from this listing, three new Category A generic tasks (A-42, A-43 and A-44) have been identified.
These tasks were identified as a direct result of this review to identify " Unresolved Safety Issues."
Task A-42, Pipe Cracks in Boiling Water Reactors, was identified from the list of Abnormal Occurrences.
This Abnormal Occurrence was an incident at the Duane Arnold nuclear facility regarding cracking in primary system piping.
This recent incident and the fact that pipe cracking has recently been reported in large diameter piping in a boiling water reactor in Germany, resulted in identifying this issue as an " Unresolved Safety Issue" and designating it as a Category A task in the NRC program.
Task A-43, Containment Emergency Sump Reliability, and Task A-44, Station Blackout, address issues that were previously assigned to Categories B and C.
The issues addressed by Task A-43 were originally to be addressed in Tasks B-18 and C-3.
Task A-44 was originally to be addressed by Task B-57.
However, the review of the relevant information on these issues indicated that they were of potentially greater safety significance than judged at the time of their original priority category assignment.
These two issues were elevated to Category A and were identified as " Unresolved Safety Issues."
As an aid in understanding why particular issues in the NRC program were not included as " Unresolved Safety Issues," discussions of a number of the issues that were considered individually are provided below.
The issues selected for discussion were Category A issues that were judged not to qualify and several lower priority category issues that the preliminary risk-based evaluation (Appendix C) identified as potentially risk significant.
CATEGORY A ISSUES THAT WERE NOT REPORTED AS " UNRESOLVED SAFETY ISSUES" The assignment of an issue to Category A does not necessarily mean that the issue is safety significant, since, for example, some Category A issues are generic environmental issues.
Accordingly, all Category A generic tasks would not be expected to involve " Unresolved Safety Issues."
Nonetheless, because Category A tasks were repo.'ted to Congress last year in NUREG-0410 as the highest priority tasks in the NRC program, a discus-sion is provided below for each Category A issue that did not qualify as an " Unresolved Safety Issue." The discussion describes the basis for not including the issue as an " Unresolved Safety Issue." A more complete description of each Category A task is provided in NUREG-0371.
Generic Task A-13 Snubber Operability Assurance Shock and vibration arrestors (snubbers) are used in nuclear power plants as pipe whip and seismic restraints.
Operating experier.ue reports show that a substantial number of snubbers have leaked hydraulic fluid and the rejection rate from functional testing and inspection has been high.
The
..... - - - purpose of this task is to evaluate current practice and develop comprehen-sive procedures to better assure the operability of snubbers.
The types of snubber problems that have been experienced do not represent a " major reduction in the degree of protection of the public health and safety" because the faults experienced only represent degraded conditions rather than conditions that prevented operation of the affected snubbers.
In addition, as a result of the faulty snubber experience, augmented inservice surveillance and operability tests were required at operating facilities.
These current requirements provide assurance that faulty snubbers will be detected should they occur and corrective actions (i.e.,
repair or replacement) will be implemented.
Implementation of these requirements has markedly increased the availability of snubbers.
Based on the above considerations, this tasks does not involve an " Unresolved Safety Issue."
Generic Task A-14 Flaw Detection Assurance of the integrity of the pressurized systems in a nuclear power plant, including the reactor pressure vessel, the primary system piping and other safety systems, is based, in part, on caref'P and repeated inspections for possible flaws in the metals that form the pressure retaining boundaries of the system components.
The purpose of this task is to assess the capability of current and new advances in flaw detection
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. methods and recommended improvements in equipment, methods, and require-ments for inclusion in industry and regulatory standards, codes and guides.
Although improvements may be possible, current requirements are considered to be adequate from a safety standpoint and this task does not involve a
" major reduction in the degree of protection of the health and safety of the public." Areas where a more immediate need,or improvement in flaw detection techniques may be required are included in Tasks A-10, "BWR Nozzle Cracking," and A-42, " Pipe Cracks in Boiling Water Reactors," which are identified as " Unresolved Safety Issues." Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Task A-15 Primary Coolant System Decontamination and Steam Generator F
Chemical Cleaning The presence of a layer of highly radioactive corrosion products adhering to the interior surfaces of the primary coolant system has, in one case, prevented licensees from carrying out some of the less important inservice inspections required by their technical specifications.
Because of the safety significance of the system and components being inspected, an approach should be developed to permit these inspections while at the same time minimizing personnel radiation exposures.
Several methods of decontami-nation to reduce radioactivity levels in the primary system are under review by the nuclear industry for application in operating reactors.
.... - _ _ _ _ _. These include chemical decontamination, electropolishing, mechanical and hydraulic decontamination.
This generic task involves reviewing the existing and ongoing decontami-nation technology and providing guidance to the staf f and industry relating to acceptable methods of decontamination of reactor primary coolant systems.
This task will also address the chemical cleaning of steam generato a
.Jr the purpose of reducing the rate of steam generator tube degradation.
Task A-15 does not involve an " Unresolved Safety Issue" because the radia-tion levels in operating reactors do not presently prevent licensees from carrying out the important safety-related inspections, modifications and repairs.
One plant, Dresden 1, which is not shut down for Commission ordered modifications, will undergo a primary system decontamination during 1979 to reduce occupational exposures during modifications and primary coolant system insp(ctions.
On other plants, remote methods of inspection may be utilized or the radiation levels are currently low enough to permit inspections to be carried out.
With regard to the chemical cleaning of steam generators, steam generator tube integrity is assured by leak monitoring and by periodic inspections, the extent and frequency of which are varied according to the condition of the steam generator.
Because chemical cleaning is a technique to extend the service life of the generator, review of this procedure does not involve an " Unresolved Safety Issue."
Generic Task A-16 Steam Effects Oa BWR Core Spray Distribution The core spray systems are one component of the emergency core cooling system for all boiling water reactors.
Core spray systems have a nozzle or a set of nozzles arranged to distribute water over the top of the core following a postulated loss-of-co:
<,t accident.
Each fuel bundle must receive a specified minimum flow from se core spray system in order to provide the post-loss-of-coolant accident spray cooling assumed in the analyses performed 'or the purpose of determining the acceptability of emergency core cooling system pefarL:nce.
Acceptability of the spray flow assumed to be available to each fuel bundle following a postulated loss-of-coolant accident has traditionally been justified by utilizing full scale upper plenum mockup tests (in air),
with complete core spray spargers and full scale nozzles.
However, the steam environment that would be present following a loss-of-coolant accident can, in c6.rtain cases, significantly alter the spray distribution.
The problem is to quantitatively assess the degree of conservatism present in t!.e assumed spray flow (and the dependent spray cooling heat transfer coefficients) including the steam effects.
The subject is not considered to be an " Unresolved Safety Issue" because (1) results from preliminary bounding tests i.dicate the presence of adequate margin in the quantity of spray flow provided by presently installed core spray systems, (2) results of an ongoing GE-EPRI-NRC program are expected to confirm the preliminary test results, and (3) for the majority of boiling water reactors (jet pump BWRs) spray flow degradation would not cause major reduction in the degree of protection provided, because the separate core flooding systems would not be affected and would provide a diverse means of cooling.
Generic Task A-18 Pipe Rupture Design Criter.a Current criteria for postulating pipe breaks and specifying the protection therefrom have been developed over a long period of time and accordingly, lack consistency when applied inside and outside of the containment and are subject to misinterpretation in certain areas.
In addition, the effect on normal operation of piping design requirements for postulated accidents needs to be further considered.
The purpose of this tasks is to review these criteria and recommend changes in these areas.
Although these criteria may be improved, the possible deficiencies do not represent a major reduction in the degree of protection of the health and safety of the public.
Adequate protection from pipe breaks is now provided and this task does not involve an " Unresolved Safety Issue."
Generic Task A-19 Digital Computer Protection Systems Some reactor protection systems which initiate control rod insertion now incorporate digital computers.
The staff has reviewed one such system and others are under review.
The purpose of this task is to standardize and document the acceptance criteria and methodology for the safety review of digital computer protection systems, and thereby improve the guidance available to NRC staff reviewers.
Digital systems when proposed by applicants are currently being reviewed on a case-by-case basis, which is adequate.
This task is not directed toward affecting the level of safety, but toward improving the efficiency of licensing reviews.
Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Task A-20 Impacts of the Coal Fuel Cycle Compliance with the National Environmental Policy Act (NEPA) requires that alternatives to a proposed Federal action be considered and that required alternatives be balanced against the base case in terms of associated environmental impacts.
A coal fired plant is currently the only realistic alternative to a nuclear power plant.
Present treatment of the coal alternative is aimed essentially
- at economics and public health impacts.
It is relatively incomplete in other areas of impact.
This task will provide a comprehensive summary which evaluates the environmental effects of the coal fuel cycle in a form directly comparable to that for the uranium fuel cycle.
In the absence of such a generic treatment of the effects of using coal for generating electric +ser, it is necessary for the NRC staff to develop an analysis de novo for each licensing action, to present this individual analysis in detail in the staff's Environmental Impact Statement, and to defend it throughout the hearing process.
This repetitive staff effort will be avoided by preparing a generic statement suitable to support rulemaking proceedings.
After the rulemaking procedure, such a statement would avoid repetitive staff effort in individual cases.
Since this task is associated with an environmental issue that does not affect plant safety, it does not involve an " Unresolved Safety Issue."
Generic Task A-21 Main Steam Line Break Inside Containment Safety-related equipment inside of the containment of a nuclear power plant is qualified for the most severe accident conditions under which it is expected to function.
In a pressurized water reactor, this has, in the past, been assumed to be the pressure and temperature that would accompany a loss of coolant accident resulting from the failure of the largest pipe in the reactor primary system.
However, preliminary calculations indicated
.....__.__ - that the failure of a main steam line inside of the containment may result in a temperature that is higher than the temperature calculated for a loss-of-coolant accident and, therefore, possibly higher than the tempera-ture for which the safety-related equipment is qualified.
The purpose of this task is to recommend acceptable methods of calculating environmental conditions that would result from a steam line failure within the contain-ment for the purpo;e of qualifying safety-related equipment.
Although preliminary calculations indicated that the temperature within the containment following a steam line break could be significantly higher than following a loss of-coolant accident, the duration of the high tempera-ture was calculated to be short.
Because of the relatively low heat transfer rate in superheated steam and the heat capacity of the affected safety-related equipment, the equipment itself would not be expected to exceed the temperature for which it was qualified as a result of this short duration peak in the temperature of the containment atmosphere.
Therefore, although this task may result in an improved basis for deter-mining the environmental conditions for equipment qualification, it does not involve a major reduction in the degree of protection to the health and safety of the public, and therefore, does not involve an " Unresolved Safety Issue."
Generic Task A-22 PWR Main Steam Line Break, Core, Reactor Vessel and Containment Building Response Several aspects of the main steam line break analyses for pressurized water reactors as currently provided by license applicants have been questioned.
This task involves evaluating these questions or concerns to confirm or modify the present NRC staff position on these analyses.
The first concern involves the current reliance oi the operation of non-safety grade equipment as a backup for assumed single active failure in safety grade equipment following a main steam line break.
This task will evaluate plant response t-the operation or nonoperability of various nonsafety grade systems and components, and will develop a reliability assessment of such equipment.
The majority of the components in the secondary system are essential to plant operation or availability, and are in a state of continuous or frequent operation.
The considerable experience gained from both fossil and nuclear plant operation has demonstrated the high reliability of such components.
Awareness of this reliability level led to the current staff position cf permitting credit in accident analyses for selected nonsafety grade equip-mer,t as backup to safety grade equipment, even though documented evidence of the reliability levels for such systems and components was not available.
This task effort is likely to confirm the reliability of this equipment and thus support the present staff position.
An additional concern involves the mechanicci response of the pressure T.s task will consider safety vessel following a main steam line break.
systems and operator actions required to (a) maintain acceptable pressure vessel stress levels, and (b) achieve long-term cooling.
This potential safety problem related to reactor vessel integrity does not become important until the vessel has been subjected to extended neutron irradiation during plant operation.
The irradiation effect is to reduce the allowable stress at reduced temperatures late in the life of the vessel.
There must also be a potential for this concern to be significant.
When considering the sequence of conditions following a main steam line break, the primary system is first depressurized by overcooling through the secondary system.
The reduction in primary system pressure causes a reactor trip and actuation of the emergency core cooling system (ECCS).
Pressure reductions in the primary system are accompanied by temperature decrease with shrinkage of the liquid volume.
Actuation of the ECCS replenishes the volume of liquid.
Unless terminated or controlled by the operator, the ECCS could eventually refuel and repressurize the primary system to the safety valve set point.
This task involves evaluating the timing requirements for operator actions, the nature of the actions, and
- the likelihood of accomplishment and thus confirm that the operator actions necessary to maintain pressure vessel integrity can be reliably accomplished.
The ncnsafety grade equipment and operator actions currently relied on to mitigate the consequences of a main steam line break accident provide substantial protection of the public health and safety.
This task is expected to confirm that such reliance is appropriate.
Accordingly, this task does not involve an " Unresolved Safety Issue."
Generic Task A-23 Containment Leak Testing One of the requirements of all operating licenses for water-cooled power reactors is that the primary reactor containment meet the leakage test requirements of Appendix J to 10 CFR Part 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." The NRC staff and reactor licenses have experienced some difficulties in implementing Appendix J since its inception.
The purpose of this task is to revise this appendix so as to clarify existing requirements, and resolve conflic-ting and impractical requirements.
Such clarification is being done now on a case-by-case bcsis as part of the NRC staff review process.
This task does not involve any change in the degree of protection of the public health and safety, but will improve the efficiency of licensing reviews.
Therefore, this task does not involve an " Unresolved Safety Issue."
G_eneric Task A-25 Nonsafety Loads On Class lE Power Sources Present regulatory practice permits nonsafety loads to be connected to the emergency onsite power system (Class 1E power sources) that provides the electrical power to safety systems.
The purpose of this task is to determine whether or not the re'i-hility of the Class 1E power sources is significantly affected by the sharing of safety and nonsafety loads.
This task is expected to confirm that current practice is acceptable and may even result in the relaxation of some requirements.
Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Tas< A-27 Reload Applications The purpose of Task A-27 is to provide updated and formalized review procedures for licensee core reload submittals.
Completion of this task will provide for more efficient NRC staff review of core reload submittals.
Existing licensing acceptance criteria utilized in the NRC staff's review of reload submittals provide assurance that plants will be operated safely.
Since this task involves improving licensing efficiency and does not affect plant safety, it does not involve an " Unresolved Safety Issue."
Generic Task A-28 Increase in Spent Fuel Storage Capacity This task involves the development of consistent and formalized acceptance criteria regarding the use of high density storage racks in existing spent fuel storage pools.
Revisions of current guidelines are being developed that incorporate insights gained in the case-by-case reviews of applications for increased spent fuel storage pool capacity.
This task involves documenting and formalizing the acceptance criteria currently being used for the review of applications for increased spent fuel storage capacity at nuclear power plants.
Therefore, it does not involve an " Unresolved Safety Issue."
Generic Task A-29 Design Features to Control Sabotage The objective of this task is to identify and evaluate possible plant design variations which could improve the inherent sabotage resistance of nuclear power plants.
Should this program identify promising design alternatives, appropriate changes in the NRC's regulations will be developed for future plants.
For current plants high assurance of protection against industrial sabotage is achieved by the physical security measures required by 10 CFR 73.55.
Although Task A-29 may identify design concepts that could provide alter-nate or more effective means of protection against sabotage, the implemen-tation of such design improvements is not necessary to provide a high level of protection of nuclear power plants.
This task, therefore, does not involve an " Unresolved Safety Issue."
Generic Task A-30 Adequacy of Safety-Related DC Power Supplies This generic task originated from a letter to the NRC's Advisory Committee on Reactor Safeguards from one of its consultants that questioned the reliability of DC power supplies at nuclear power stations.
The specific concern expressed was as follows:
Wnile a nuclear power plant is operating, one of two redundant DC power supply systems fails causing a reactor scram and subsequently causing loss of all offsite power.
At this point, safe shutdown of the plant requires that the residual heat from the decay of radioac-tivity be removed from the reactor.
Control of valve position and pumps needed to remove residual heat after plant shutdcwn depends on availability of the DC power supply.
If all remaining sources of DC power were lost, continued cooling of the reactor core cannot be assured.
_ _ _ _ _ _ _ The NRC staff's view is that the simultaneous and independent failure of redundant DC power supplies is so unlikely as to be incredible and that their failure from a common event is judged to be low enough in likelihood that adequate protection of the public health pre.ently exists, but that additional technical studies to be provided as part of this task should and will be performed to add confidence to this judgment.
This view stems from the following:
(1) the postulated scenario is highly unlikely; (2) the period of vulnerability to the above cited single failure of the redundant DC power supply is limited, i.e., both the DC power supply failure initiating the scenario, and the second failure of the remaining source of DC power must occur within 30 seconds to defeat starting of the redun-dant diesel and acceptance of critical loads; and (3) the degree of vulnerability is mitigated substantially by the availability of alternative measures for restoration of power or for removal of decay heat and of sufficient time (at least an hour) for operator implementa-tion of these alternative measures.
A more detailed discussion of the design of DC power supply systems and of the NRC staff's view on the postulated accident scenario described above is provided in NUREG-035, " Technical Report on DC Power Suppliers in Nuclear Power Plants."
w_ Accordingly, while this issue is important and warrants the quantita-tive assessment of reliabilities of DC power supplies that will be provided by Tass A-30, particularly with respect to common mode failures, it does not involve an " Unresolved Safety Issue."
Generic Task A-32 Missile Effects Current regulatory requirements include protection of safety-related structures, systems and equipment from missiles that might result from a tornado, failure of the main turbine or failure of equipment within the plant during an accident.
Design criteria for the barriers used as protec-tion from missiles are based on limited available experimental data and analyses.
The purpose of this task is to reevaluate these criteria and quantify the safety margin in the design of missile barriers.
The current criteria were conservatively established, recognizing the limitations of the available data and analyses, and provide substantial safety margins.
This task is expected to confirm that the criteria are conservative and may even result in some relaxation of requirements.
Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Task A-33 NEPA Review of Accident Risks In 1971, the AEC determined that, consistent with NEPA, the environmental assessments of requests for construction permits and operating licenses
_ should include consideration of the possible impacts from accidents.
An Annex to 10 CFR 50 Appendix 0 was proposed which provided guidance to applicants in this regard.
The purpose of this Task is to conduct limited additional analyses and prepare a summary survey document which could be used as a standard reference regarding accident risks in the context of the staff's environmental reviews.
This same docum'snt would serve as the principal basis for a decision regarding finali',ag the proposed Annex to 10 CFR 50 Appendix 0.
This task is associated with evaluating accidents in the context of environ-mental reviews of nuclear power plants and accordingly, is not directly relevant to plant safety.
Therefore, it does not involve an " Unresolved Safety Issue."
Generic Task A-34 Instrumentation for Monitoring Radiation and Process Variables During Accidents The purpose of this task is to develop criteria and guidelines to be used by applicants, licensees and staff reviewers to support implamentation of Regulatory Guide 1.97, Revision 1, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident."
Such criteria and guidelines will provide specific guidance on functional and operational capabilities required of the various classes of instruments.
Where such guidance can not be provided, the rationale to be used to derive requirements for specific situations will be provided.
This task is intended to increase the efficiency and consistency of the licensing process by developing implementation and review criteria for implementing the regulatory positions of a Regulatory Guide that is already in effect.
Although there are some philosophical and technical issues that must be resolved concerning implementation of the Regulatory Guide, the safety issue involved (i.e., instrumentation to follow the course of an accident) has been resolved by issuance of the Regulatory Guide.
Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Task A-35 Adequacy of Offsite Power The NRC requires that electric power for safety systems be comprised of two redundant and independent divisions, each capable of providing the necessary plant protection functions during all normal operating conditions and following various design basis accidents.
Each division includes an offsite AC power connection (the preferred power source), a standby emergency diesel generator AC power s, ply (capable of powering essential safety systems should the offs' ource be lost), and DC power sources.
Events at several plants involving the loss or degradation of the offsite power system or involving its connection to the emergency onsite power system have indicated that a reassessment of current staff requirements was appropriate.
This task was undertaken to perform such an assessment and to determine the need, if any, for upgrading the offsite power sources and/or their interfaces with the onsite power system at nuclear power stations.
Although this task may identify areas where current criteria should be modified to increase safety margins beyond those currently provided, extensive modifications are not anticipated.
Where such deficiencies have been identified on specific plants, corrective actions are now underway.
This issue does not involve a " major reduction in the degree of protection of the public health and safety" and, therefore, does not involve an
" Unresolved Safety Issue."
Generic Task A-37 Turbine Missiles Protection of essential systems from turbine missiles is required by the NRC staff unless the combined generation, strike and damage probability is very small.
For most new plants, adequate protection against turbine missiles is provided by favorable turbine placement and orientation and adherence to the guidelines cf Regulatory Guide 1.115, Rev. 1, " Protection Against Low-Trajectory Turbine Missiles."
For plants that have safety-related structures, systems and components that are potentially susceptible to turbine missile strikes because of unfavorable turbine placement for example, a more detailed evaluation of turbine missile protection is required.
Currently, each such plant is reviewed on a case-by-case basis to assure that probability of unacceptable damage is acceptable or, if not, that appropriate measures are taken to reduce this probability.
The purpose of Task A-37 is to assess the methods currently used to estimate the probability of damage to essential systems used in these case-by-case reviews, to quantify the effect of steps that can be taken by applicants to reduce the damage probability, and to recommend means of assuring that the probability of unacceptable damage is sufficiently small.
Alt'ough this task will provide a more uniform review by providing better guidance to reviewers and applicants, the currentiy used case-by-case methods are sufficiently conservative to assure adequate protection of the public health and safety.
Therefore, this task does not involve an " Unresolved Safety Issue."
Generic Task A-38 Tornado Missiles The requirement of General Design Criterion 2, that structures, systems, and components of nuclear plants important to safety be designed to withstand the effects of tornadoes without loss of capability to perform their safety functions, has imposed new demands on the practice of structural engineering.
For other than nuclear safety-related structures, tornadoes have always been considered too rare to be included in the design basis.
Consequently, no body of practice existed and design criteria for tornado resistance have had to be developed.
The NRC staff requirements regarding tornado missile protection that have evolved result in conservative design practice.
Many believe that the requirements are excessively conservative and result in unnecessary costs to plant applicants.
However, the NRC staff has lacked an adequate technical basis for further reducing the conservatism that currently exists.
The purpose of this task is to define a set of design basis tornado missiles (whose effects envelope those expected at plant sites) that do not impose unnecessary design requirements on plant construction and for which a sound technical basis exists.
Since this task involves consideration of the relaxation of current requirements, it does not involve an " Unresolved Safety Issue."
SELECTED CATEGORY B ISSUES THAT WERE NOT REPORTED AS " UNRESOLVED SAFETY ISSUES" Several Category B issues and one Category C issue were identified in the preliminary risk-based evaluation discussed on page 13 as potentially risk significant (i.e., assigned to Category I or II in Appendix C).
As noted
... - - - _ on pages 16 and 17 above, because of this and other considerations, two of these Category B and the one Category C task were elevated to Category A and included as " Unresolved Safety Issues" in the 1978 NRC Annual Report.
The remaining Category B issues were considered in detail for reporting as
" Unresolved Safety Issues," but were judged not to qualify.
These Category B issues are discussed below.
Each discussion includes a description of why the issue was considered to be potentially risk signifi-cant in the preliminary risk-based evaluation and why, nonetheless, the issue was judged not to qualify as an " Unresolved Safety Issue."
Generic Task B-30 Design Basis Floods and Probability The purpose of this task was to prepare a paper for presentation to the Advisory Committee on Reactor Safeguards (ACRS) detailing the bases for design basis flood events used by the NRC licensing staff in case reviews.
Additionally, the task was to address the possible use of probability estimates for the principal flood producing events.
This task has been completed and a report to the ACRS was issued in Ju;y 1977.
The report presents discussion and definitions of flood events which may be used as Design Basis Floods for review of nuclear power plants.
It supports continued use by the staff of a deterministic approach for identifying the Design Basis Flood events in preference to possible use of a probabilistic approach.
The deterministic approach identifies the upper limit of flood
.. _., potential physically possible.
As indicated in the report, the NRC licensing staff does not feel that a probabilistic approach is appropriate for use in licensing reviews at the present time because of the lack of confidence one has in estimates of extreme flooding events using current techniques.
The preliminary results of the risk-based evaluation discussed on page 13 indicate that the probability of a flood-induced core meltdown accident at most sites is very low.
However, the study notes that detailed probabilis-tic estimatas have not been performed for specific sites and that prelimi-nary indications from flood-data analyses performed by the NRC's Office of Nuclear Regulatory Research using probabilistic techniques indicate that flooding events could potentially be risk significant for some sites.
On this basis, this issue, i.e., the use of deterministic versus probabilistic methodology for determining design basis floods, was categorized as poten-tially risk significant in the preliminary risk-based evaluation.
As indicated above the licensing staff's conclusion is that the determin-istic approach is acceptably conservative and, at present, is the preferred approach for use in licensing reviews because of the large uncertainties associated with probabilistic estimates of very infrequent events.
Nonethe-less, ongoing research efforts aimed at developing improved methodological techniques for the probabilistic analysis of flooding are being undertaken by the NRC's Office of Nuclear Regu'atory Research.
In any event, the NRC
s
. staff's view is that this generic task does not involve an " Unresolved Safety Issue."
Generic Task B-34 Occupational Radiation Exposure Reduction This task involves the development of oJJitional criteria and guidelines to provide aa improved basis for the staff to review reactor plant designs and operations to support full implementation of the NRC's regulatory requirement that radiation exposures should be maintained as low as is reasonably achievable.
The preliminary risk-based evaluation discussed on page 13 points out that occupational radiation exposures at operating nuclear facilities are averaging roughly 400 man-rem per reactor year and have generally been increasing with time.
Further, the expected value for the annual accident exposure associated with plants analyzed in the Reactor Safety Study (WASH-1400) is predicted to be approximately 250 man rem per reactor year.
Although it was recognized in the study that a meaningful comparison of the occupational exposure risks with those associated with accidents is difficult, the study concluded that reduction of occupational exposures can be very important to reducing the overall radiologically-associated risks associated with the nuclear reactor industry.
On this basis, this task was categorized as potentially risk significant in the preliminary evaluation.
The preliminary evaluation did not indicate what, if any, J
__ reduction in occupational exposures could be achieved by completing this particular generic task.
This assessment of the significance of occupational exposures in the preliminary risk-based evaluation is consistent with the NRC staff's view of the importance of occupational radiation exposure reduction, as evidenced by the reouirement to maintain such exposures as low as is reasonably achievable.
In this regard, general guidance is now available to the industry in Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Station Will Be As Low As Is Reasonably Achievable." This guidance has been utilized by the staff in performing licensing reviews for a number of years.
Task B-34 will draw from that experience and, with the aid of supplementary studies, will develop additional criteria regarding techniques and methods to maintain occupational radiation exposure as low as is reasonably achievable.
Although the preliminary risk-based evaluation was correct in that occupa-tional radiation exposures are important, current NRC requirements and staff review procedures assure that they will be maintained as low as reasonably achievable.
This task may provide some improved guidance to designers and operators.
However, substantial improvements over current guidance are not anticipated.
On this basis, this task does not involve an " Unresolved Safety Issue."
Another aspect of Task B-34 is an attempt to develop criteria for making decisions regarding trade-offs between decreased occupational exposures and increased potential exposure to the public resulting from accidents.
Such trade-offs are considered in determining whether or not to waive inspec-tion requirements for safety-related components in plant locations where significant radiation exposures could occur.
The accomplishment of this task may result in improved decision making criteria for evaluating these trade-offs, however, this aspect of the task was judged not to involve an
" Unresolved Safety Issue."
Generic Task B-55 Improved Reliability of Target-Rock Safety-Relief Valves The purpose of this task is to monitor current General Electric Company and Target-Rock (valve manufacturer) valve programs related to improving the reliability of Target-Rock safety-relief valves and to develop generic technical positions for use in the review of individual plants.
The preliminary risk-based evaluation indicated that this task was combined with a number of others that relate to potentially damaging loads in boiling water reactor pressure suppression containments (A-6, A-7, A-8 and A-39) and, from the risk-based perspective, all were categorized as risk significant.
Tasks'A-6, A-7, A-8 and A-39 all have been included as tasks that address " Unresolved Safety Issues" (see Appendix A).
The effects of all modes of inadvertent safety-relief valve operation are being consiriered by the NRC staff as part of these tasks.
With regard to safety-relief valve malfunctions such as spurious openings and " failure-to-close" af ter openings that have occurred, licensees, valve manufacturers and General Electric have been jointly working to improve valve reliability.
This effort has resulted in corrective programs that are expected to improve the reliability of safety-relief valves.
In addition, an NRC staf f analysis presented in NUREG-0462, entitled "Teuinical Report on Operating Experience with BWR Pressure Relief Systems," concludes that inadvertent operation of safety-relief valves, even at higher frequencies than actually experienced, would not have significant load effects or result in significant offsite radiological consequences.
On this basis, this task does not involve an " Unresolved Safety Issue."
Generic Task B-63 Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary There are several systems connected to the reactor coolant pressure boundary which have design pressures considerably below the reactor coolant system operating pressure.
The Reactor Safety Study (RSS) identified a potential inter sysi 'm loss-of-coolant accident (LOCA) in a pressurized water reactor as a significant contributor to the risk of core melt.
The inter-system LOCA identified in the RSS was the failure of two check valves in the injection lines of the residual heat removal system (or low presssure injection system), that would allow the high pressure reactot coolant to communicate with the low pressure piping outside of containment.
Rupture of the low pressure piping could result in loss of reactor coolant outside of containment and subsequent core meltdown.
On this basis, the risk-based evaluation concluded that the possible improvement in procedures for examining interfacing system isolation devices, that might result from thir task, was potentially risk significant.
Improved procedures for reducing the likelihood of such accidents were developed by the NRC staff and implemented for CP and OL reviews after the results of the draft RSS became available.
Task B-63 was originally undertaken to review representative operating plants to assess their isolation capabilities with regard to the reactor coolant system-low pressure system interface.
This review was completed with a conclusion that adequate high pressure-low pressure isolation protection axisted in operating reactors.
Further optimization of requirements for various valve configurations to provide additional protection, particularly during valve testing, are still being considered by the NRC staff.
Current configurations and review procedures provide substantial protection and pressure isolation during valve testing is addressed in plant specific staff reviews of Inservice Testing (IST) Programs.
Therefore, this task was judged not to involve an " Unresolved Safety Issue."
M-Generic Task B-64 Decommissioning of Reactors The Code of Federal Regulations 10 CFR 50.82 provides criteria by which licensees may terminate their licenses.
Under this regulation, the Commis-sion may require information from the licensee to demonstrate that the methods and procedures to be used for decontamination and for disposal of radioactive materials provide reasonable assurance that the dismantling and disposal will not be inimical to the common defense and security ar to the health and safety of the public.
10 CFR 50.33(f) includes the require-ment that operating license applicants show that they possess or have reasonable assurance af obtaining funds necessary to covar the " estimated costs of permanently shutting the facility down and maintaining it in a safe condition."
Since 1960, about 50 research-type reactor facilities and 15 small power and test reactors have been decommissioned in accordance with the above regulations.
In addition, the NRC reviews the general plans for decommis-sioning and financial arrangements for decommissioning as a part of its review of operating license applications.
Based on acceptable findings, including this area, the NRC has issued operating licenses.
As a result of the need for increased guidance to the industry in this area, the staff published in June 1974 a Regulatory Guide (1.86) on the " Termination of Operating Licenses for Nuclear Reactors." This guide includes methods and procedures considered acceptable by the staff for the termination of licenses for operating reactors.
However, because of the increasing interest in decommissioning, additional guidance is needed on this topic.
The preliminary risk-based evaluation indicates that the possibility of significant exposures to a large number of plant personnel exists if decommissioning activities are not carried out properly.
On this basis, this issue was categorized in the evaluation as risk significant.
Such exposures are an important consideration of any decommissioning program.
Accordingly, the studies and resultant safety acceptance criteria and guidelines for decommissioning operations developed under this task currently include consideration of occupational radiation safety.
In addition, current requirements to keep occupational exposures as low as reasonably achievable (ALARA) require that decommissioning plans proposed by licensees are reviewed within the context of ALARA regulations.
While it is anticipated that improved guidance will be forthcoming as a result of this task, its completion is not expected to significantly reduce occupational exposures during decommissioning operations.
Accordingly, this task does not involve an " Unresolved Safety Issue."
TABLE 1 PRIORITY CATEGORY DEFINITIONS Category A:
Those generic technical activities judged by the staff to warrant priority attention in terms of manpower and/or funds to attain early resolution.
These matters include those the resolution of which could (1) provide a significant increase in assurance of the health and safety of the public, or (2) have a significant impact upon the reactor licensing process.
Category B:
Those generic technical activities judged by the staff to be important in assuring the continued health and safety of the public but for which early resolution is not required or for which the staff perceives a lesser safety, safeguards or environmental significance than Category A matters.
Category C:
Those generic technical activities judged by the staff to have little direct or immediate safety, safeguards or environmental significance, but which could lead to improved staff understanding of particular technical issues or refinements in the licensing process.
Category D:
Those proposed generic technical activities judged by the staff not to warrant the expenditure of manpower or funds because little or no importance to the safety, environmental or safeguards aspects of nuclear reactors or to improving the licensing process can be attributed to.the activity.
November 1978 Table 2 LIST OF GENERIC TASKS IN THE NRC PROGRAM Category A Tasks fask No.
Tit;e A-1 Water Har r
A-2 Asymmetric Blowdown Loads on PWR Primary Coolant Systems A-3 Westinghouse Steam Generator Tube Integrity A-4 Combustion Engineering Steam Generator Tube Integrity A-5 Babcock & Wilcox Steam Generator Tube Integrity A-6 Mark I Short Term Program A-7 Mark I Long Term Program A-8 Mark II Containment Pool Dynamic Loads A-9 ATVS A-10 BWR Nozzle Cracking A-ll Reactor Vessel Materials Toughness A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports A-13 Snubber Operability Assurance A-14 Flaw Detection A-15 Primary Coolant System Decontamination and Steam Generator Chemical Cleaning A-16 Steam Effects on BWR Core Spray Distribution A-17 Systems Interaction in Nuclear Power Plants A-18 Pipe Rupture Design Criteria A-19 Digital Computer Protection Systems A-20 Impacts of the Coal Fuel Cycle A-21 Main Steam Line Break Inside Containment - Evaluation of Environmental Conditions for Equipment Qualification A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Response A-23 Containment Leak Testing A-24 Qualification of Class IE Safety-Related Equipment A-25 Nonsafety Loads on Class IE Power Sources A-26 Reactor Vessel Pressure Transient Protection (Overpressure Protection)
A-27 Reload Applications A-28 Increase in Spent Fuel Pool Storage Capacity A-29 Nuclear Power Plant Design for the Reduction of Vulner-ability to Industrial Sabotage A-30 Adequacy of Safety-Related DC Power Supplies A-31 RHR Shutdown Requirements A-32 Missile Effects A-33 NEPA Reviews of Accident Risks A-34 Instruments for Monitoring Radiation and Process Variables During Accidents
Table 2 (cont'd)
Task No.
Title A-35 Adequacy of Offsite Power Systems A-36 Control of Heavy Loads Near Spent Fuel A-37 Turbir.a Missiles A-38 Torrade Missiles A-39 Deterr ination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containments A-40 Seismic Design Criteria - Short Term Program Category B Tasks B-1 Environmental Technical Specifications B-2 Forecasting Electricity Demana By State in the United States on an Annual Basis B-3 Event Categorization B-4 ECCS Reliability B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containment B-6 Loads, Load Combinations, Stress Limits B-7 Secondary Accident Consequence Modeling B-8 Locking Out of F.CCS' Power Operated Valves B-9 Electrical Cable Penetrations of Containment B-10 Behavior of BWR Mark III Containment B-ll Subcompartment Standard Problems B-12 Containment Cooling Requirements (Non-LOCA)
B-13 Marviken Test Data Evaluations B-14 Study of Hydrogen Mixing Capability in Containment Post-LOCA B-15 CONTEMPT Computer Code Maintenance B-16 Protection Against Postulated Piping Failures in Fluid Systems Outside Containment B-17 Criteria for Safety-Related Operator Actions B-18 Vortex Suppression Requirements for Containment Sumps B-19 Thermal-Hydraulic Stability B-20 Standard Problem Analysis B-21 Core Physics B-22 LWR Fuel B-23 LMFBR Fuel B-24 Seismic Qualification of Electrical and Mechanical Components B-25 Piping Benchmark Problems B-26 Structural Integrity of Containment Penetrations B-27 Implementation and Use of Subsection NF B-28 Radionuclide/ Sediment Transport Program B-29 Effectiveness of Ultimate Haat Sinks
Table 2 (cont'd)
Task No.
Title B-30 Design Basis Floods and Probability B-31 Dam Failure Model B-32 Ice Effects on Safety-Related Water Supplies B-33 Dose Assessment Methodology B-34 Occupational Radiation Exposure Reduction B-?5 Confirmation of Appendix I Models for " Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Light-Water-Cooled Power Reactors" B-36 Develop Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems B-37 Chemical Discharges to Receiving Waters B-38 Reconnaissance Level Investigations B-39 Transmission Lines B-40 Effects of Power Plant Entrainment on Plankton B-41 Impacts on Fisheries B-42 Socioeconomic Environmental Impacts B-43 Value of Aerial Photographs for Site Evaluation B-44 Forecasts of Generating Costs of Coal and Nuclear Plants B-45 Need for Power - Energy Conservation B-46 Costs of Alternatives in Environmental Design B-47 Inservice Inspection Criteria for Supports and Bolting of Class 1, 2, 3 and MC Components B-48 BWR CRD Mechanical Failure (Collet Housing)
B-49 Inservice Inspection Criteria for Containment B-50 Requirements for Post-0BE Inspection B-51 Assessment of Inelastic Analysis Techniques B-52 Fuel Assembly Seismic and LOCA Responses B-53 Load Break raitch B-54 Ice Condenscr Containments B-55 Improved Reliability of Target-Rock Safety-Relief Valves B-56 Diesel Reliability B-57 Station Blackout B-58 Passive Mechanical Failures B-59 Review of (N-1) Loop Operation in BWRs and PWRs B-60 Loose Parts Monitoring Systems B-61 Allowable ECCS Equipment Outage Periods B-62 Reexamination of Technical Bases for Establishing SLs, LSSSs, etc B-63 Isolation of Low Pressure Systems Connected to RCPB B-64 Decommissioning of Reactors l:
L
Table 2 (cont'd)
Task No.
Title B-65 Iodine Spiking B-66 Control Room Infiltration Measurements B-67 Effluent and Process Monitoring Instrumentation B-68 Pump Overspeed During a LOCA B-69 ECCS Leakage Ex-containment B-70 Power Grid Frequency Degradation and Effect on Primary Coolant Pumps B-71 Incident Response B-72 Development of Models for Assessing Risk of Health Effects and Life Shortening from Uranium and Coal Fuel Cycles B-73 Monitoring for Excessive Vibration Inside the Reactor Pressure Vessel Category C Tasks C-1 Assurance of Continuous Long-Term Integrity of Seals on Instrumentation and Electrical Equipment C-2 Study of Containment Depressurization by Inadvertent Spray Operation to Determine Adequacy of Containment External Design Pressure C-3 Insulation Usage Within Containment C-4 Statistical Methods for ECCS Analysis C-5 Decay Heat Update C-6 LOCA Heat Sources C-7 PWR System Piping C-8 Main Steam Line Leakage Control System C-9 RHR Heat Exchanger Tube Failures C-10 Effective Operation of Containment Sprays in a LOCA C-11 Assessment of Failure and Reliability of Pumps and Valves C-12 Primary System Vibration Assessment C-13 Non-Random Failures C-14 Storm Surge Model for Coastal Sites C-15 NUREG Report for Liquid Tank Failure Analysis C-16 Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Category D Tasks 0-1 Advisability of a Seismic Scram D-2 Emergency Core Cooling System Capability for Future Plants D-3 Control Rod Drop Accident (BWRs)
APPENDIX A - 1978 ANNUAL REPORT TEXT ON
" UNRESOLVED SAFETY ISSUES" Unresolved Safety Issues Plan in 1977, the NRC's Office of Nuclear Reactor ment oversight of task progress, and public Regulation (NRR) instituted a program to dissemination of information related to the tasks define, categorize and manage generic technical as they progress. The NRR program is, however, activities on a systematic basis. The initial effort of considerably broader scope than the "Unre-under this program resulted in the identification solved Safety issues Plan" required by Section of 133 generic tasks. These tasks cover a variety 210. As noted above, the program also includes of topics. Some are related to safety, some to other generic tasks of importance to the NRC's environmental matters, and some to improving mission, such as those for the resolution of en-the regulatory process.
vironmental issues; for the development of im-Subsequent to the inception of the NRR pro-provements in the reactor licensing process; for gram, the Congress acted, in late 1977, to consideration of less conservative design criteria amend the Energy Reorganization Act of 1974 or operating limitations, in areas where overly to include, among other things, a new Section conservative requirements may be unnecessarily 210, as follows:
restrictive or costly; for the maintenance and
- "*E"
" UNRESOLVED SAFETY ISSUES PLAN" perform mdependent audit calculations; and for "Section 210. The Commission shall develop a the actual performance of independent audit plan providing for specification and analysis of calculations.
unresolved safety issues relating to nuclear reac.
This Annual Report section is limited to tors and shall take such action as may be describing the progress on that portion of the necessary to implement corrective measures with NRR program required to be reported to the respect to such issues. Such plan shall be submit-Congress by Section 210.
ted to the Congress on or before January 1, The following definition of an " Unresolved 1978 and progress reports shall be included in Safety Issue" was developed for use in identify-the annual report to the Commission ing the generic issues in the broader NRR staff thereafter."
program that should be reported to Congress, In response to this reporting requirement, the pursuant to Secdon 2R "An Unresolved Safety NRC provided a report to the Congress issue is a matter affecting a number of nuclear (NUREG 0410) in January 1978 describing the P wer plants that poses important questions generic issues program of the Office of Nuclear C ncerning the adequacy of existing safety re-Reactor Regulation that had been implemented quirements for which a final resolution has not earlier in 1977. The NRR program described in yet been developed and that involves conditions NUREG-0410 provides for the identification of n t likely to be acceptable over the lifetime of generic issues, the assignment of priorities, the the plants affected."
development of detailed Task Action Plans to resolve the issues, projections of dollar and manpower costs, continuing high level manage-
.__.______._m___ _ _
All of the generic issues reported to the Con.
and/or the consequences of an accident gress last year in NUREG4410, as well as any scenario, for which the issue under study is an other issues identified since that time, were con.
important consideration, is smaII.
sidered as candidates for " Unresolved Safety The NRC staff's conclusions in this regard are issues." A systematic review of these issues was subjected to the scrutiny of the licensing process undertaken by the NRC staff. As an aid to this in individual cases. Specifically, the NRC staff's review, an evaluation was made of the subject conclusions on individual applications are areas involved according to their relative impor.
reviewed by the Advisory Committee on Reactor tance from the standpoint of public risk. This Safeguards and are specifically addressed in the risk-based characterization was used together public hearing process (see previous section in with a substantial body of additional informa.
this chapter describing the licensing process).
tion (e.g., heavy weight was given to issues aris.
The seventeen generic issues listed in Table 2 ing from events reported to the Congress as were determined to be " Unresolved Safety
" Abnormal Occurrences") to determine which Issues." These issues are addressed by twenty-issues met the definition of an " Unresolved two generic tasks in the NRR Program for the Safety Issue." The review resulted in the iden.
Resolution of Generic Issues. The task numbers tification of seventeen " Unresolved Safety of the applicable generic tasks are provided in Issues." The Subcommittee on Generic Items of parentheses following the title of each issue in the Commission's Advisory Committee on Reac.
Table 2. Three of the twenty-two generic tasks tor Safeguards (ACRS) has been briefed on the addressing these seventeen issues have been com-identified issues. The NRC staff will continue to pleted. Generic Task A-6 was completed and coordinate with the ACRS on these issues and documented in a report NUREG4408, " Mark I future issues considered for reporting as Containment Short Term Program Safety
" Unresolved Safety Issues." (The selection pro.
Evaluation Report," in December 1977; Generic cess and the rationale for decisions regarding Task A-26 was completed and documented in particular issues are described in a separate NUREG-0224, " Reactor Vessel Pressure Tran-report, NUREG-0510 " Identification of sient Protection for Pressurized Water
' Unresolved Safety issues' Relating to Nuclear Reactors," in September 1978; and Generic Task Power Plants-A Report to Congress.")
A-31 was completed and documented in
- E" *
" e 1.139, " Guidance for Residual Although the term " Unresolved Safety issue" Heat Removal,,, in May 1978.
has been in use for some time, and the Congress A discussion of each of the " Unresolved Safe-used the term to identify those issues about ty Issues" follows~
which it wished to be kept informed, it has been frequently misunderstood. An immediate ques-tion is: if a generic safety issue (i.e., a safety issue relating to more than one plant)is
" unresolved," then how can NRC grant a license to operate a specific nuclear power plant for which that issue is relevant? The answer is that before the license is granted the NRC staff must determine that licensing and operation of the specific plant can continue pending a generic resolution of the issue. The bases for such a determination include one or more of the following: (1) the issue does not apply to or has been resolved for the plant under consideration; (2) interim measures assuring adequate safety of operation are being required at affected plants pending final resolution of the issue; (3) resolu-tion of the issue can reasonably be expected before the plant under consideration begins operation; or (4) the likelihood of occurrence A-2
I Table 2: Unresolved Safety Issues and Related Task Numbers 1.
Water Hammer -(A-1) 2.
Asymmetric Blowdown Leads on the Reactor Coolant System -(A-2) 3.
Pressurized Water Reactor Steam Generator Tube Integrity - (A-3, A-4, A-5) 4.
BWR Mark I and Mark 11 Pressure Suppression Containments -(A-6, A-7, A-8, A-39) 5.
Anticipated Transients Without Scram -(A-9) 6.
BWR Nozzle Cracking -(A-10) 7.
Reactor Vessel Materials Toughness -(A II) 8.
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports -(A-12) 9.
System Interactions in Nuclear Power Plants -(A-17)
- 10. Emironmental Qualification of Safety-Related Electrical Equipment - (A-24)
- 11. Reactor Vessel Pressure Transient Protection -(A-26)
- 12. Residual Heat Removal Requirements
'A-31)
- 13. Control of Heavy Loads Near Spent Fuel-(A-36)
- 14. Seismic Design Criteria -(A-40)
- 15. Pipe Cracks in Boiling Water Reactors - (A-42)
- 16. Containment Emergency Sump Reliability - (A-43)
- 17. Station Blackout -(A-44)
Water Hammer Water hammer events are intense pressure partially empty lines, and rapid valve motions.
pulses in fluid systems, such as commonly ex.
Most of the damage has been relatively minor, perienced when rapidly closing a water faucet, though there have been several cases of failure and they often occur in nuclear power plant or partial failure of system piping.
fluid systems. Since 1971, about 100 incidents While no water hammer incident has resulted involving water hammer in nuclear power reac.
in the release of radioactivity outside of a plant, tors have been reported. These incidents have in.
the concern is that water hammer could result in volved many types of fluid systems, including the failure of a pipe in the reactor coolant steam generator feed-rings, feedwater and steam system or disable a system required to cool the supply piping, residual heat removal systems, plant after a reactor shutdown.
emergency core cooling systems, containment The means to prevent one particular type of spray systems, and service water systems. Water water hammer caused by the rapid condensation hammer has been attributed to various causes, of steam in the steam generator feed-rings of such as the rapid condensation of steam pockets, some pressurized water reactors are being in-steam-driven slugs of water, pump start-up with stituted. In addition, applicants with new samm A-3
I generator designs are being required to demon-components, and (4) causing other ruptures in strate through test or analysis that water ham-the initially unbroken reactor coolant system mer will not occur in these designs. Plants with piping loops and attached systems.
steam generators-of the top feeding type that The NRC staff's review of this safety issue are subject to water hammer-are being required has been incorporated in the NRC Program for to modify the feed-rings and/or test the systems Resolution of Generic Issues as Generic Tark to assure water hammer will not occur. And A-2. The issue was originally identified in May other actions to correct the specific causes of 1975 by the Virginia Electric and Power Com-water hammer identified to-date are being re-pany in relation to its North Anna Units 1 and 2 quired.
nuclear power piants. A survey cf all operating The NRC staff's review of this safety issue pressurized water reactors (PWRs) was con-has been incorporated in the NRC Program for ducted in October 1975 which showed that Resolution of Genecic Issues as Generic Task asymmetric blowdown loads had not been con-A-1. The potentia' for water hammer in various sidered in the design of the reactor vessel sup-systems is being rvaluated and appropriate re-ports for any operating PWR facility. In June quisements and systematic review procedures are 1976, the NRC staff requested all operating being developed to ensure that water hammer is PWR licensees to assess the adequacy of the given appropriate consideration in all areas of reactor vessel supports at their facilities with licensing reviews. A technical report providing respect to these newly identified loads.
the results of a staff review of water hammer Most licensees with plants using Westinghouse events in nuclear power plants is scheduled for nuclear steam supply systems initially proposed publication m February 1979. Issuanse of this an augmented in-service inspection program report completes a major subtusk of Genene (ISI) of the reactor vessel safe-end-to-end pipe Task A-1. The remainmg subtasks are expected welds in lieu of providing the detailed analysis to be completed m 1980.
requested by the NRC staff. Licensees with Combustion Engineering nuclear steam supply systems submitted a probability study in support Asymmetn.c Blowdown Loads of a conclusion that the probability of a break On the Reactor Coolant System at the location in the piping necessary to pro-duce the postulated load was so low that no fur-In the very unlikely event of a rupture of the ther analysis was necessary. Licensees with Bab-primary coolant piping in light water reactors, cock and Wilcox nuclear steam supply systems large non-uniformly distributed loads would be took an approach similar to Combustion imposed upon the reactor vessel, reactor vessel Engineer ng licensees.
internals, and other components in the reactor The NRC staff's review of these proposed coolant system. The potential for such asym-alternatives to detailed plant-specific analyses metric loads, which result from the rapid has been completed with the conclusion that depressurization of the reactor coolant system, proposed alternatives to the requested analysis was only recently identified and was not con-should not be accepted. Accordingly, the NRC sidered in the original design of some facilities.
staff sent letters on January 25,1978, to all The forces associated with a postulated break in PWR licensees and applicants stating that an the reactor coolant piping near the reactor analysis must be undertaken to assess the design vessel, for example, could affect the integrity of adequacy of the reactor sessel supports and the reactor vessel supports and reactor pressure other structures to withstand the loads when vessel internals. A significant failure of the reac-asymmetric loss-of-coolant accident forces are for vessel support system, besides impacting the taken into account. As part of Task A-2, the reactor internals, has a potential for (1) damag.
NRC staff will review and approve analytical ing systems designed to cool the core following models and computer codes deveiuped by reac-the postulated piping break, (2) affecting the tor vendors to calculate asymmetric blowdown capability of the control rods to function prop.
loadings, prior to their use by licensees and ap-erly, (3) damaging other reactor coolant system plicants in plant-specific analyses. In addition, A-4
the staff will develop explicit guidelines and ac.
workers.
ceptance criteria for the asymmetric load anal-A detailed discussion of the specific problems yses and will conduct a pipe break probability associated with steam generator tube integrity study.
that were occurring at operating reactors was Plant modifications to assure that the provided in the 1977 NRC Annual Report, page postulated loads are accommodated have been
- 95. The information below is provided to sup-implemented late in the construction stage of plement and update that information.
several plants and have been proposed and are Corrosion resulting in steam generator tube under staff review for some operating plants.
wall thinning has been observed in several For plants still under operating license review, Westinghouse and Combustion Engineering (CE) the NRC staff requires that plant-specific plants for a number of years. Major changes in analyses be completed and any necessary plant their secondary water treatment process essen-modifications completed prior to issuance of an tially eliminated this form of degradation.
operating license. The generic efforts for Another major corrosion-related phenomenon pressurized water reactors under Task A-2 are has also been observed in a number of plants in currently scheduled for completion in early 1979.
recent years, resulting from a build-up of sup-The NRC staff has been investigating this port plate corrosion products in the annulus be-phenomenon as it applies to boiling water reac.
tween the tubes and the support plates. This tors and has determined that asymmetric loads build-up eventually causes a diametral reduction are also significant and therefore need to be of tubes, called " denting," and deformation of evaluated for these lower pressure systems. The the tube support plates. This phenomenon has staff is currently developing plans for expanding led to other problems, including stress corrosion Task A-2 to resolve this issue for boiling water cracking, leaks at the tube / support plate in-reactors.
tersections, and U-bend section cracking of tubes which were highly stressed because of sup-port plate deformation.
PWR Steam Generator Tube Integrity The significant developments in Westinghouse and Combustion Engineering steam generators, The heat produced in the reactor at a nuclear since June 1977, were the following:
power plant is used to convert water into steam which will drive the turbine-generators. In plants Continued tube denting at Indian Point employing pressurized water reactors, the Unit 2, San Onofre Unit 1, Surry Units I primary coolant water which extracts heat by and 2, Turkey Point Units 3 and 4, and circulating through the reactor core is kept lesser amounts of denting at a number of under pressure sufficient to prevent boiling. This ther Westinghouse designed reactors.
high-pressure water passes through tubes around Steam generator replacement is planned which a secondary coolant (also water) is cir-f r early 1979 r 1980 at Surry Units I culating, under somewhat lower pressure. The and 2. Replacement or retubing is also water in the secondary system is allowed to boil being considered for Turkey Point Units 3 and 4. In the interim, the units are and produce steam to drive the turbine-generators. The assembly in which the transfer perating under restrictions imposed by the NRC.
takes place is the steam generator. The tubes Discovery of support plate cracking within it are an integral part of the primary coolant boundary, keeping the radioactive (related to denting) at Indian Point Unit 2 primary coolant in a closed system and isolated and San Onofre Unit 1.
from the environment. The primary concern is Removal of several tubes and a section of the capability of steam generator tubes to main-support plate at Indian Point Unit 2 to tain their integrity during normal operation and investigate the potential for steam postulated accident conditions. In addition, the generator cleaning revealed continued ac-requirements for increased steam generator tube tive corrosion of the support plate.
inspections and repairs have resulted in signifi-Continuation of tube denting at Mi!! stone cant increases in occupational exposures t Unit 2 and Maine Yankee and discovery A-5
of denting in St. Lucie 1. Millstone Unit steam generator tube problems, it has under-2, Maine Yankee, and Arkansas Nuclear taken a number of generic reviews and studies as One Unit 2 have removed lugs and por.
part of three generic : asks in the NRC Program tions of the solid rim in the uppermost for the Resolution of Generic Issues; specifical-support plates to reduce the susceptibility ly, Generic Tasks A-3, A-4, and A-5 each of the plates to denting-related cracks directed at the particular problems of (CE designs).
Westinghouse, Combustion Engineering, and Babcock and Wilcox plants, respe:tively.
Palisades Nuclear Power Station is sleev-Under these tasks generic studies will be con-ing degraded tubes instead of plugging ducted to (1) evaluate inservice inspection results them. This process restores the structural ir m cperating reactors, (2) evaluate the conse-integrity of the tubes while keeping them quences of tube failures under postulated acci-in service (CE designb dent conditions, (3) evaluate tube structural in-Another form of steam generator tube tegrity, (4) establish tube plugging criteria based degradation m Babcock and Wilcox (B&W) on new information, (5) define the requirements steam generators was found in the Oconee for monitoring secondary coolant chemistry, (6)
Nuclear Plant where the first tube leak occurred evalute inservice inspection methods, and (7) m July 1976. To-date,14 tube leaks, all at the review design improvements proposed for new Oconee um,ts, ha've occurred in B&W steam plants. These studies will be used to revise cur-generators. The majority of these leaking tubes rent NRC staff requiiements and guidance re-were located adj,acent to the open inspection garding these subjects. In addition, under Task lane. Laboratory examination of removed defec-A-3, the NRC staff will review and evaluate the tive tubes mdicated that the tube failure were first proposed steam generator replacement caused by the propagation of circumferential operation to establish acceptance criteria and fatigue cracks by flow-induced vibration.
guidance on a generic basis for use in the review The sigmficant developments in B&W steam of subsequent replacement operations. These generators, since May 1977, were the following:
generic tasks are currently scheduled to be com-Continued tube leaks at the Oconee units.
pleted in early 1980.
Initiation of a demonstration tube sleev-ing program by Duke Power Company at BWR Mark I and Mark II the Oconee units. The tube sleeves will Pressure Suppression Conta,intnents not serve as part of the primary coolant boundary but will be installed to change the vibrational characteristics of the tubes In the course of performing large scale testing and decrease the dynamic stresses and the of an advanced design pressure-suppression con-susceptibility of the tubes to fatigue tainment (Mark Ill), and during in-plant testing cracking.
of Mark I containments, new suppression pool Following inspections by licensees of their hydrodynamic loads were identified which had steam generators and the completion of any not explicitly been included in the original Mark necessary repair programs, the NRC approves or I or Mark 11 containment design basis. These concurs in the restart of each of the severely af-additionalloads result from dynamic effects of fected facilities.* To-date, the units severely af-drywell air and steam being rapidly forced into fected by the tube denting have completed in-the suppression pool (torus) during a postulated spection and repair programs and received NRC LOCA and from suppression pool response to approval for operation for limited time periods.
various modes of safety relief valve (SRV) lafe op,.uw i: nsured by the imposition of operation generally associated with plant tran-strict conditions on licensed operation, requiring sient operating conditions. Since these new the plugging of affected tubes and restricting hydrodynamic loads had not been explicitly con-allowable leak rates during opcration.
sidered in the original design of the Mark I and As the NRC staff continues to closely Mark II containments, the NRC staff deter-monitor, evaluate, and appiove the acceptability mined that a detailed reevaluation of these con-of continued operation of plants experiencing tainment system designs was required.
A-6
t -
As a result of the need for this reevaluation the objectives of the Short-Term Program had the affected utilities formed ad hoc Mark I and been satisfied and documented the basis for this Mark II Owners' Groups and each has engaged conclusion in the " Mark I Containment Short-the General Electric Company as its program Term Program Safety Evaluation Report,"
manager. Both Owners' Groups developed two-NUREG4408, dated k~: ember 1977. (Thus phase programs consisting of a short-term pro-Task A-6 was completed in December 1977.)
gram and a long-term program for resolution of The objectivo of the Mark I Long-Term Pro-the pool dynamic concerns for their respective gram are: (1) to establish design basis loads that containment designs. The Owners' Groups' pro-are appropriat : for the anticipated life of each grams include a number of comprehensive ex-Mark I BWR.acility, and (2) to restore the perimental and analytical programs to establish original inten: ed design safety margins for each generic pool dynamic loads, load combinations Mark I containment system. The Mark I Long-and design criteria.
Term Program consists of a series of major The NRC staff has identified and initiated a tasks and subtasks which are designed to provide number of generic tasks to review and evaluate a detailed basis for hydrodynamic load oefini-the results of the Mark I and Mark II Owner's tion and the methodology and acceptance Group short-term and long-term programs to criteria for the structural assessments. The develop technical positions for use in licensing generic aspects of the Mark I Long-Term Pro-actions on individual plants utilizing the Mark I gram will be described in a Plant Unique and Mark Il containment designs. These generic Analysis Applications Guide, scheduled to be tasks are included in the NRC Program for completed in February 1979, and in the Load Resolution of Generic Issues (described in Definition Report, a portion of which was com-NUREG4410 as noted above). Specifically, they pleted in December of 1978. The remainder of are Task A-6, Mark I Short-Term Program; the Load Definition Report is scheduled to be Task A-7, Mark I Long-Term Program; Task completed in March 1979. Subsequently, each i
A-8, Mark 11 Containment Program; Task A-39, utility with a Mark I plant will perform a plant-Determination of Safety Relief Valve (SRV) unique analysis using approved load definition Pool Dynamic Loads and Temperature Limits and structural analysis techniques to for BWR Containments.
demonstrate conformance with the Mark I The objectives of the Mark I Short-Term Pro-Long-Term Program structural acceptance gram were: (1) to examine the containment criteria. These analyses are currently scheduled system of each BWR facility with a Mark I con.
for completion in October 1979.
tainment design to verify that it would maintain The scheduled completion date for the Mark I its integrity and functional capability when sub-Long-Term Program (Task A-7), including the jected to the most probable hydrodynamic loads issuance of license amandments and the im-induced by a postulated design basis loss-of-piementation of any plant modifications coolant accident; and (2) to verify that licensed necessary to satisfy the Mark I Long-Term Pro-Mark I BWR facilities may continue to operate gram structural acceptance criteria, is December safely, without undue risk to the health and 1980. In recognition of th;s schedule, a number safety of the public, while a methodical, com-of facilities are adopting their own schedules to prehensise Long-Term Program is conducted.
implement anticipated plant modifications and The NRC determined that, for the Short-Term minimize the potential for extended plant Program, " maintenance of containment integrity utages or unscheduled outages, and function" would be adequately assured if a The objective of the NRC staff's efforts under safety factor to failure of at least two were Generic Task A-8 related to the Mark II Short-demonstrated to exist for the weakest structural Term Program (STP) was to review and evalute or mechanical component in the Mark I contain_
the pool dynamic loads associated with a ment system (i.e., if the calculated stresses in all p stulated large loss-of-coolant accident pro-components of the affected containment struc-p sed by the Mark 11 Owner's Group to deter-ture were shown to be less than one-half the mine their acceptability for use in plant unique stress which would cause the component to lose an lyses. The Mark II Short-Term Program was its structural integrity). The NRC concluded that A-7
completed in October 1978 and documented in very low probability events from the design NUREG-0487, " Mark 11 Containment Lead basis. At issue in the ATWS discussions is Plant Program Load Evaluation and Acceptance whether or not the probability of an AT"/S Criteria." With regard to the hiark 11 Long.
event is sufficiently low to warrant the con-Term Program (LTP), the NRC staff will tinuance of the current staff practice with regard evaluate the results of the Mark 11 confirmatory to ATWS,i.e., continued exclusion from the experimental and analytical programs to assess design basis for nuclear power plants because of the margin for selected loads. The hiark 11 its low probability.
Long-Term Program is currently scheduled for Because of the perceived potential for serious completion in October 1980.
consequences resulting from ATWS events, a Under Generic Task A-39, the NRC staff will number of studies have been undertaken to resiew and evalute the results of the Mark I and assess the probabilities and consequences of such Mark 11 Owners' Group's experimental and events. These studies have been performed by analytical programs to establish and justify the vendors, utility groups, and by the AEC and safety relief valve-related pool dynamic loads for NRC regulatory staff. The ATWS issue was in BWR Mark I and Mark Il containment designs.
corporated in the NRC Program for Resolution The results of Generic Task A-39 will be an in-of Generic issues (described in NUREGot10, as tegral part of the f' mal acceptability of the Mark noted above) as Generic Task A-9.
I and Mark 11 pressure suppression containment In September 1973, the then-AEC staff designs. This generic task is currently scheduled published WASH-1270, " Technical Report on for completion in December 1979. An interim Anticipated Transients Without Scram for Water assessment of multiple-consecutive SRV Cooled Power Reactors," which set forth staff discharges was performed for the operating
" acceptance criteria" to protect against ATWS Mark I facilities to support deferral of the events. During the two-year period following resolution of this issue until the completion of publication of the staff report, each of the four the Mark I Long-Term Program. This review reactor manufacturers submitted analyses and was completed in December 1978 and deferral supporting information on ATWS which was was found to be acceptable. A safety evaluation reviewed by the NRC staff and addressed in describing the NRC staff's interim :.sessment four status reports published in December 1975.
will be issued in early 1979.
The staff reports evaluated the information for conformance to the WASH-1270 criteria and noted where design changes and additional Anticipated Transients Without Scram analyses were required.
The vendors and owners have questioned whether the NRC staff's requirements are Nuclear plants have safety and control systems necessary and justified. The industry contends to limit the consequences of temporary abnor.
that the probability of an ATWS event is mal operating conditions or " anticipated tran.
sients." Some deviations from normal operating significantly less than estimated by the NRC staff and so low as to make ATWS events minor conditions may be minor; others, occurring less safety concerns in light water reactor operations.
frequently, may impose significant demands on Because of the continuing controversy over plant equipment. In some anticipated transients, the NRC staff position since its publicatior. in rapidly shutting down the nuclear reaction (ini-WASH-1270, a staff review and evaluation of a, tiating a " scram"), and thus rapidly reducing the information available on the subject of the generation of heat in the reactor core, is an ATWS, and in particular, the material develope, important safety measure. If there were a poten.
subsequent to the publication of the staff status tially severe " anticipated transient" and the reports referred to above, was undertaken in the reactor shutdown system did not " scram" as latter part of 1977 and early 1978. A report, desired, then an " anticipated-transient-without.
NUREGot60, was published in April 1978 pro-scram," or ATWS, would have occurred.
viding the results of this review and evaluation.
This issue has been discussed throughout the It was concluded in NUREGot60 that con-nuclear industry for a number of years.
sidering the expected frequency of transients, the Historically, the regulatory staff has excluded A8
reliability of current reactor scram systems Excessive crack growth could lead to im-necessary to meet the safety objectives has not pairment of pressure vessel safety margins been demonstrated and may well have not been requiring more complicated repair work attamed. NUREG-0460 recommended that than simple grinding.
means of mitigatmg the consequences of ATWS events be provided in plant designs.
The design safety margin could be re-The recommendations presented in NUREG.
duced by excessive removal of base metal.
The exposure to radiation of the person-0460 have been criticized by industry and some members of the NRC staff as unnecessarily con-nel performing inspection and repair tasks servative and therefore too costly. The staff is can be considerable.
now evaluating alternative means of reducing the The repair of these kinds of cracks can probability or consequences of ATWS events, result in considerable shutdown time at other than that recommended in NUREG-0460.
the plant affected.
The effectiveness, cost and other factors, such as The reactor vendor (the General Electric Coni-the effect on the licensing process of these alter-pany) and the NRC have concluded from their natives, is being evaluated. Based on this evalua-ive studies that the cracking is caused by tion, the staff will re:ommend to the Commis-fluctuations or " cycling" of the temperature on sion the alternatives which provide the best the inside surface of the nozzles; that the balance between safety and cost for new designs, stainless steel cladding exhibited less resistance plants under construction and operating plants.
to crack initiation than the underlying low-alloy The staff expects to provide its recommenda-steel; and that, after initiation in the stainless tions to the Commission m early 1979.
steel cladding, cracks can be propagated by operational startup and shutdown cycles or other per ti n Ily-induced transients. The vendor has BWR Nozzle Cracking performed extensive analysis and testing to con-firm the suspected cause of the cracking and to Over the last several years, inspections at 21 uncover possible long-term solutions - a newly of the 23 boiling water reactor (BWR) plants designed sleeve, removal of the stainless steel licensed for operation in the U.S. have disclosed cladding, reduction of the temperature differen-some degree of cracking in the feedwater nozzles tial at the nozzle, or some combination of these, of the reactor vessel at all but three facilities.
The licensees involved have mereased the Two facilities have not yet accumulated signiii-
- r. umber and extent of inspections of feedwater cant operating time and have not yet been in-nozzles, with careful repair and reinspection spected, although all BWR plants will eventually where cracks were found. The vendor advised be inspected for this problem.
these licensees to closely monitor startup and The feedwater nozzles, part of the " pressure shutdown procedures in an effort to substantial-vessel," are an integral part of the primary ly reduce the time during which cold feedwater pressure boundary of the reactor coolant system s being injected into the hot pressure vessel.
and the second barrier (after the fuel cladding)
In a closely related area, the NRC was in-to the release of radioactive fission products. All formed in March 1977 by the General Electric of the repaired BWR feedwater nozzles met the Company that a crack had been found in the ASME pressure vessel code limits, however, and noz2le of the " control rod drive (CRD) return no immediate action was necessary. Because on-line" in a reactor vessel in a foreign country.
ly relatively small amounts of base metal have The CRD return line nozzles are the openings in been removed by repair operations, there has BWR pressure vessels through which the high been no significant reduction in safety margins.
pressure water in excess of that needed to Several plants have removed the stainless steel operate and cool the CRDs is returned to the nozzic cladding as a means of eliminating crack pressure vessel. Later in March, the Philadelphia initiation, since the clad thickness was not Electric Company reported that similar cracking necessary to meet code reinforcement re-had been found in the CRD return line nozzle at quirements. Nevertheless, the cracking is poten-its Peach Bottom Atomic Power Station, Unit 3.
tiary serious because:
The cracks resembled those found in the feed-A-9
water nonles and seemed to be the result of the same kind of cyclic thermal stresses that were Reactor Vessel Materials Toughness causing feedwater nonle cracks. Both the foreign reactor and the Peach Bottom Unit 3 reactor are representative of a small number of Resistance to brittle fracture, a rapidly prop-BWRs which do not have a thermal sleeve in the agating catastrophic failure mode for a compo-CRD return line noule.
nent containing flaws, is described quantitatively The licensee removed the cracks in the Peach by a material property generally denoted as Bottom CRD nonle by grinding out the cracked
" fracture toughness." Fracture toughness has area, the maximum crack depth being 7/8-inch, different values and characteristics depending and returned the unit to operation with the CRD upon the material being considered. For steels return line " valved out" and with the flow and used in nuclear reactor pressure vessel, three pressure in the CRD hydraulic system modified.
considerations are important. First, fracture Inspection of other CRD return line nonles toughness increases with increasing temperature.
which incorporated thermal sleeves indicated Second, fracture toughness decreases with in-that these sleeves may not be effective in pre-creasing load rates. Third, fracture toughness venting this cracking phenomenon. The Georgia decreases with neutron irradiation.
Power Company found a crack in the CRD in recognition of these consid rations, power return line nonle at its Hatch Plant, Unit 1, reactors are operated within restrictions imposa 3 which did have a thermal sleeve. (The crack was by the Technical Specifications on the pressure removed, the nonle capped, and the return line during heatup and cooldown operations. These rerouted to the reactor water cleanup system.)
restrictions assure that the reactor vessel will not The NRC staff efforts related to the resolu-be subjected to that combination of pressure and tion of these two similar issues regarding nonle temperature that could cause brittle fracture of cracking in boiling water reactors were con-the vesselif there were significant flaws in the solidated into a single staff effort, Generic Task vessel material. The effect of neutron radiation A-10, in 1977. Under Generic Task A-10, the on the fracture toughness of the vessel material staff issued interim guidance to operating plants is accounted for in developing and revising these in a report entitled, " Interim Technical Report Technical Specification limitations over the life on BWR Feedwater and Control Rod Drive of the plant.
Return Line Nonle Cracking," in July 1977.
For the service times and operating conditions The staff is often requiring in-service inspection typical of current operating plants, reactor vessel using liquid penetrant examinations at operating fracture toughness provides adequate margins of reactors in accordance with the frequency, pro-safety against vessel failure. Further, for most cedures and acceptance criteria described in the plants the vessel material properties are such that adequate fracture toughness can be main-above report.
Additional efforts under Generic Task A-10 tained over the life of the plants. However, include following and reviewing advancements in results from a reactor vessel surveillance pro-(1) the development and testing of effective feed-gram indicate that up to 20 older operating water noule thermal sleeves and spargers, (2) pressurized water reactors were fabricated with life-cycle testing of certain CRD system valves, materials that will have marginal toughness after (3) the development of various feedwater system comparatively short periods of operation.
and CRD system modifications, and (4) the The objective of Task A-11 is to evaluate deselopment of viable ultrasonic system tech-material degradation mechanisms resulting from neutron irradiation and determine appropriate niques by the nuclear industry to allow reliable and consistent early determination of cracking licensing criteria and corrective action for low from positions exterior to the reactor vessel.
toughness reactor vessel materials in these cur-Generic Task A-10 is scheduled for comple-rently licensed plants. Task A-11 is currently scheduled for completion in July 1979. This tion in late 1979.
completion date is wellin advance of the date needed to assure that adequate fracture toughness is maintained in these older plants.
A-10
will be used to supplement the heat derived nom the reactor coolant loop to obtain the required perating temperature of the support materials.
Fracture Toughness and Potential For Lamellar Tearing of PWR Steam B*'""S* sinular materials and designs gag been used m. other plants and therefore similar Generator and Reactor Coolant Pump pr blems may exist, review of this issue was in-Sup@
cluded in the NRC Program for Resolution of Generic issues as Generic Task A-12.
During the course of licensing review for a A consultant was engaged to reassess the frac-specific Pressurized Water Reactor (PWR) a ture toughness of the steam generator and reac-number of questions were raised as to (1) the tor coolant pump support materials for all adequacy of the fracture toughness properties of operating PWR plants and those in the later the material used to fabricate the reactor coolant stages of operating license review. The staff
~
pump supports and steam generator supports, thereafter completed a review of the materials and (2) the potential for failure due to lamellar utilized in the supports of 34 potentially affected tearing of these same supports. The safety con-PWRs. Based on the consultant's preliminary cern is that, although these supports are de-evaluation,it was determined that there are ap-signed for worst-case accident conditions, poor proximately 15-20 plants whose supports are of fracture toughness or lamellar tearing could questionable toughness. We expect that these cause the supports to fail during such accidents.
plants may be required to utilize in-service in-Support failure could conceivably impair the ef-spection or auxiliary heating if adequate fectiveness of systems designed to mitigate the toughness properties cannot be demonstrated.
consequences of the accident. (An example of a Upon completion of the generic study, the postulated event sequence of potential concern generic phase of the fracture toughness program would be a large pipe break m the reactor win be documented and the results implemented coolant system which severely loads the sup-on, o plant-specific basis. Lessons derived from ports, followed by a support failure of sufficient the generic solution will be incorporated into the magnitude that a major component such as a Standard Review Plan for use in future license steam generator is severely displaced resulting in reviews.
failure of the emergency core cooling system The staff has concluded that continued opera-piping which is needed to provide cooling water tion (and licensing) of PWRs is justified pending to the core.)
completion of this task and implementation of Two different steel specifications (ASThi the task results because support failure is not ex-A36-70a and ASTN1 A572-70a) covered most of pected to occur except under the unlikely com-the material used for the supports of the PWR bination of:
in question. To address the fracture toughness (1) The occurrence of an...imtiating event question (lamellar tearing is discussed separately (e.g., a large pipe break) which has been below), tests not originally specified and r.ot in determined to be of low probability (nor-the relevant ASTN1 specifications were made on mal operatmg stresses on pipmg are very those heats of steel for which excess material I *)
was available. The toughness of the A36 steel was found to be adequate, but the toughness of (2) The existence of non-redundant and the A572 steel was relatively poor at an operat-critical support structural member (s) with ing temperature of 80'F. In the case of the low fracture toughness (many supports PWR in question, the applicant agreed to a contain redundant members).
license condition which stated that he would (3) The existence of support structural raise the temperature of the ASThi A572 beams members at operating temperatures low in the steam generator supports to a minimum enough that the fracture toughness of the temperature of 225'F-prior to pressurizing the support material is reduced to a level at reactor coolant system above 1,000 which brittle failure could occur if a large psig-thereby assuring adequate toughness in the flaw existed.
event of an accident. Auxiliary electrical heat A-11
(4) The existence of a flaw of such size that teractions between and among systems. Such the stresses imparted during the initiating adverse events might occur, for example, event could cause the flaw to rapidly because designers did not assure that redundancy propagate, resulting in brittle failure of and independence of safety systems were provid-the member (s).
ed under all conditions of operation required, which might happen if the functional teams were The second potential concern, lamellar tear-n t adequately coordinated. Simply stated, the ing*, may also be a problem in those support left hand may not know or understand what the structures which are similar in design to those of right hand is doing in all cases where it is the aforementioned PWR. However, continued necess ry for the hands to be coordinated.
operation of PWRs during the continuing The NRC staff believes that its current review generic review of this concern was judged ac-pr cedures and safety criteria provide reasonable ceptable, based on a review of approCmately assurance that an acceptable level of redundancy 400 relevant technical documents which revealed and independence is provided for systems that on!) one instance of known failure from are required for safety. Nonetheless, in lamellar tearing. This failure occurred in often-mid-1977, this task (Task A-17) was im,tiated to stressed truck brakes. In addition, the factors confirm that present procedures adequately take considered above for the fracture toughness con-into account the potential for undesirable in-cern-such as low stresses during normal opera-teractions between and among systems.
tion and the low probability of an initiating The NRC staff's current review procedures event-apply equally to thi concern.
assign primary responsibility for review of The generic fracture tougimess program is ex-vari us technical areas and safety systems to pected to be completed in August 1979. The srecific organizational units and assign secon-lamellar tearing evaluation is a longer term ef_
d ry responsibility to other units where there is a fort and is expected to be completed in 1981.
functional or interdisciplinary relationship.
Designers follow somewhat similar procedures
.tamellar tearing is a cracking phenomenon which occurs beneath welds and is principally found in rolled steel plate and provide for interdisciplinary reviews and fabrications. The teanng always lies within the parent plate, analyses of systems. Task A-17 will provide an IIZIan is genera paral e i e weld fu undary.
independent investigation of safety func-Lamellar tearing occurs at certain critical joints usually with-tions-and systems required to perform these in large welded structures invohing a high degree of stiffness functions-in order to assess the adequacy of and restraint. Restraint may be defined as a restriction of the 8
movement of the vanous joint components that would nor-mally occur as a result of expansion and contraction of weld be conducted by Sandia Laboratories under con-metal and adjacent regions during weldmg ("Lamellar Tear-tract assistance to the NRC staff.
ing in Welded Steel Fabrication." The Weldmg Institute).
p began in May 1978 and is expected to be com-Systems Interactions pleted in September 1979. The Phase I investiga-In Nuclear Power Plants tion is structured to identify areas where interac-tions are possible between and among systems and have the petential of negating or seriously In November 1974, the Advisory Committee degrading the performance of safety functions.
on Reactor Safeguards requested that the NRC The investigation will then identify where NRC staff give attention to the evaluation of safety review procedures may not have properly ac-systems from a multi-disciplinary point of view, counted for these interactions. Finally, in a in order to identify potentially undesirable in-follow-on Phase 11 of the task, specific correc-teractions between plant systems. The concern tive measures will be taken in areas where the in-arises because the design and analysis of systems vestigation shows a need.
is frequently assigned to teams with functional As noted above, the NRC staff believes that engineering specialties-such as civil, electrical, its review procedures and acceptance criteria cur-mechanical, or nuclear. The question is whether rently provide reasonable assurance that an ac-the work of these functional specialists is suffi-ceptable level of system redundancy and in-ciently integrated in their design and analysis ac-dependence is provided in plant designs and this tivities to enable them to identify adverse in-A-12
task is expected to confirm this belief.
themselves), (2) control power to motor Nonetheless, because adverse systems interac-operators for certain valves (e.g., ECCS and tions are potentially of large significance to containment ; solation valves located inside con-plant safety, this issue has been identified as an tainment), and (3) fan cooler motors for those
" Unresolved Safety Issue." If no significant plants that utilize fan coolers for containment system interactions are identified in the Phase I heat removal.
investigation described above, as is expected, The current NRC safety review process for this issue will not be treated in subsequent nuclear power plants applies certain criteria for reports as an " Unresolved Safety issue."
confirming the capability of electrical equipment important to safety to function in the environ-ment that might result from various accident Environmental Qualification of c...ditions. Although such criteria have been ap-Safety-Related Electrical Equipment wied to varying degrees since the early days of commercial nuclear power, they have come to be defined in clearer detail over the years, in addition to the conservative design, nn-The process of clarifying the criteria has given struction and operating practices and quality rise to certain questions regarding: (1) the degree assurance measures required for nuclear power to which electrical equipment used in older plant plants, safety systems are installed at nuclear designs (those now operating)is capable of plants to mitigate the consequences of postulated withstanding the environmental conditior.s accidents. Certain of these postulated accidents (pressure, temperature, humidity, steam, could create severe environmental conditions in-chemicals, vibration, and radiation) of various side the containment. The most serious of these accident conditions under which it must be able accidents would be a high energy pipe break in to function (i.e., the " qualification of equip-the reactor coolant system pipmg or m a main ment" in these older piants), and (2) the ade-steam line. In either case, the release of hot quacy of test or analyses conducted for electrical pressurized water and steam to the containment equipment in newer plants to " qualify" such would create a high temperature environment equipment as capable of withstanding the condi-(250 to 400*F) at high humidity (including tions of the environment created by various ac-ster.m) and pressure (as high as c. 50 psig). For cidents during which the equipment must func-some applications, chemicals are added for fis-tion (i.e., the " adequacy" of qualification tests).
sion product removal to the containment sprays With regard to older plants, the following ac-that are used to reduce the pressure in the con-tions have taken place in recent months, tainment. Additionally, some electrical equip-As a result of a Sandia testing program being ment is predicted to be submerged following a conducted for the NRC Office of Nuclear large pipe break. Thus, the safety equipment is Regulatory Research, a generic safety concern exposed to such environmental conditions and with the adequacy of environmental qualifica-needs to remain operable during this period, as tion of certain electrical equipr--nt was iden-well as for the long-term post-accident period.
w a November tified. This issue was hight, v In order to assure that electrical equipment in 4,1977 petition from the U; -
Concerned safety systems will perform its function under Sc entists which requested iran t aate action by accident conditions, the NRC requires that such the NRC regarding operating power reactors and equipment-principally equipment associated licensing actions for other proposed plants. (See with the emergency core cooling system and con-
" Abnormal Occurrences-1978," in Chapter 7 tainment isolation and cleanup systems-be "en-for extended discussion of specific actions vironmentally qualified. Specific electncal following the Sandia tests.) Subsequent NRC equipment of concern during postulated accident staff investigations in response to this issue led, conditions meludes: (1) the instrumentation as of June 30,1978, to seven plant shutdowns needed to imtiate the safety systems and provide 7
g g
diagnostic information to the plant operators two other plants to make modifications. These (e.g., electrical penetrations into containment, ions were taken for the most part as a result any electrical connectors to cabling which gg, g g7 transmits signals, and the instrunants g
- 7; g
A-13
Having identified the problems associated with of them has resulted in a diversity of methods in qualification of electrical equipment, the NRC use and different levels of documentation of the conveyed that information to the licensees of all extent to which equipment is qualified.
operating reactor facilities through an Inspecti'
&veral aspects of equipment qualification are and Enforcement Circular which was issued c being pursued at this time by the NRC staff and May 31,1978. The purpose of this Circular was the nuclear industry on a generic basis, in order to ensure that the knowledge gained by the NRC to achieve a more uniform implementation of re-staff would be appropriately factored into future quirements established in IEEE Standard actions by licensees. The NRC staff also im-323-1974. One such activity is the development tiated an augmented inspection effort, t of nterim NRC staff positions regarding how become part of the normal inspection activities, the requirements of IEEE Standard 323-1974 can which will concentrate on the inspection of in-be met. This activity is a part of Generic Task stalled safety-related electrical equipment and on A-24, " Environmental Qualification of Safety-an audit of the records for environmental Related Electrical Equipment,"in the NRC Pro-qualification.
gram for the Resolution of Generic Issues and is in addition, a resiew of the environmental scheduled for completion in 1979.
qualification of safety-related electncal equip-Further efforts under Generic Task A-24 in-ment has been imtiated for II operating reactor volve the review of the environmental qualifica-facilities in the Systematic Evaluation Program tion programs of reactor vendors and ar-(SEP).
chitect/engincers as a basis for qualifying safety-With regard to the second question above-
~
related electrical equipment, pursuant to the re-the adequacy of qualification tests for newer quirements of IEEE-Standard 323-1974. Per-plants-the NRC staff has worked with the in-forming these reviews on a generic basis rather dustry to develop standards for equipment g
j;
,;g qualification and documentation which will time and resources for the NRC staff and the in-assure the high level of equipment reliability re-dustry. This follow-on portion of the generic quired for nuclear applications. This effort has task will be scheduled following completion of culminated in the development of IEEE Stan-the development of the interim NRC staff posi-dard 323, "lEEE Standard for Qualifying Class tions referred to above.
IE Equipment for Nuclear Power Generating Stations." This standard and its ancillary stan-dards have provided the focal point for the Reactor Vessel Pressure Transient development of environmental qualification re-quirements in recent years.
IEEE Standard 323 was first issued as a trial Over the past several years, incidents known use standard (IEEE Std. 32F1971) in 1971 and as " pressure transients" have taken place at later, after substantial revision, as a final stan-various PWR facilities. A pressure transient oc-dard (IEEE Std. 323-1974) in 1974. Both ver-curs when the pressure-temperature limits includ-sions of the standard set forth basic re-ed in the technical specifications for the facility quirements for environmental qualification of have been exceeded. As of the close of the electrical equipment but do not provide details report period, there had been a total of 33 such for implementation of these requirements.
events. Half of them occurred before the plant Specific qualification techniques have been achieved initial criticality (i.e., before initial reviewed and approved by the NRC staff on a operation of the reactor); the majority occurred case-by-case basis as a part of individual licens-during startup or shutdown operations. In all of ing actions. These licensing actions include ini-these incidents fracture mechanics and fatigue tial construction permit and operating license ap-calculations indicated that the reactor vessels plication reviews and requalification actions for were not damaged and continued operation of operating reactors, where documentation of the the vessels was acceptable. Nevertheless, the initial qualification was not available.
staff concluded that appropriate regulatory ac-The evolutionary nature of the process of tions were necessary (1) to reduce the frequency developing environmental qualification re-of pressure transient events, and (2) to provide quirements and the case-by-case implementation equipment which would restrict future transients A-14
to acceptable pressures. This action was substaatially lower than their hot-standby condi-necessary because reactor vessel safety margins tion values.
would be reduced over the lifetime of the vessel Even though it may generally be considered by neutron irradiation, which reduces material safe to maintain a reactor in a hot-standby con-toughness.
dition for a long time, experience shows that The NRC staff's review of this safety issue there have been events that required eventual was incorporated in the NRC Program for cooldown and long-term cooling until the reac-Resolution of Generic Issues as Generic Task tor coolant system was cold enough to perform A-26. Task A-26 was completed in September inspection and repairs. For this reason the abili-1978 with the issuance of the final report, ty to transfer heat from the reactor to the en-NUREG-0224, " Reactor Vessel Pressure Tran-vironment after a shutdown is an important sient Protection for Pressurized Water safety function for both PWRs and BWRs. It is Reactors."
essential that a power plant be able to go fcom Upgraded procedural controls were imple-hot-standby to cold-shutdown conditions (when mented at operating PWR facilities which this is determined to be the safest course of ac-significantly reduced the occurrence of pressure tion) under any accident conditions.
transient events. The few events which have oc-This issue was designated as Task A-31,"RHR curred were not significant and were of the type Shutdown Requirements," in 1977, and included that will be precluded by equipment changes, in the NUREG-0410 Report to Congress. In ac-Most of the equipment changes carried out at cordance with the Task Action Plan for this operating PWR facilities involve the addition of task, the staff's views on requirements for a second lower set point on existing power residual heat removal systems were translated in-operated relief valves, the addition of new to proposed changes to Standard Review Plan spring-loaded relief valves, or modifications to Section 5.4.7. These proposals were considered allow use of existing spring-loaded relief valves.
by the Regulatory Requirements Review Com-A few newly licensed facilities must complete mittee (RRC) during its 71st meeting on January similar design changes by their first refueling 31,1978.
shutdown. The extended equipment implementa-The RRC recommended approval of the pro-tion schedule for new facilities was based upon posed changes and further recommended that (1) the reduced frequency of occurrence of pressure the changes be applied on a case-by-case basis to transient events, a result of improved procedural all operating reactors and all other plants controls and the large safety margins for new (custom or standard) for which the issuance of pressure vessels.
the operating license is expected before January 1,1979, and (2) the changes be backfitted to all plants (custom or standard) for which construc-Residual Heat Removal tion permit or preliminary design approval ap-Shutdown Requirements plications were docketed before January 1,1978, and for which the operating license issuance is The safe shutdown of a nuclear power plant expected after January 1,1979. These recom-following an accident not related to a loss-of-mendations were approved by the Director of coolant accident (LOCA) has been typically in.
NRR and are being implemented. Accordingly, terpreted as achieving a " hot-standby" condi.
Task A-31 has been completed.
tion (i.e., the reactor is shutdown, but system Subsequently, the staff positions on design re-temperature and pressure are still at or near nor, quirements for residual heat removal systems mal operating values). Considerable emphasis were incorporated into Regulatory Guide 1.139, has been placed on the hot-standby condition of
" Guidance for Residual Heat Removal", which a power plant in the event of an accident or ab.
was issued for public comment in May 1978.
normal occurrence. A similar emphasis has been Comments were received during the latter part placed on long-term cooling, which is typically of 1978 and it is expected that this Regulatory achieved by the residual heat removal (RHR)
Guide can be issued in its final form in late 1979 system. The RHR system starts to operate when or carly 1980.
the reactor coolant pressure and temperature are A-15
not complete, movement of shielded casks over or near spent fuel has been prohibited.
Control of Loads Near Spent Fuel Concurrent with the NRC review, licensees have examined their current procedures for the movement of heavy loads over spent fuel to Overhead cranes are used to lift heavy objects, assure that the potential for a handling accident sometimes in the vicinity of spent fuel, in both that could result in damage of spent fuel is PWRs and BWRs. If a heavy object, such as a minimized while the generic evaluation proceeds.
spent fuel shipping cask or shielding block, were Most of the licensees'submittals of their reviews to fall or tip onto spent fuel in the storage pool have been received and were under review at the or in the reactor core during refueling and end of 1978.
damage the fues, there could be a release of Generic Task A-36 is expected to be com-radioactivity to the environment and a potential for radiation over-exposures to in-plant person-pleted in early 1979. The Task will result in the nel. If the dropped object is large, and is assum-deselopment of generic criteria, but implementa-tion of these criteria will be dependent on plant ed to drop on fuel containing a large amount of design characteristics and the specific procedures fission products with minimal decay time, in effect at each particular plant.
calculated offsite doses could exceed the siting guideline values in 10 CFR Part 100.
The NRC staff's review of tha safety issue has been incorporated in the NRC Program for Seismic Design Criteria Resolution of Generic Issues as Generic Task A-36. The objective of the task is to develop a NRC regulations require that nuclear power revision to the Standard Review Plan (SRP) plant structures, systems and components impor-based on a reevaluation of current NRC re-tant to safety be designed to withstand the ef-quirements and procedures currently utilized at fects of natural phenomena such as earthquakes.
operating plants. If necessary, the revision will Detailed requirements and guidance regarding provide criteria to further reduce the potential the seismic design of nuclear plants is provided for heavy loads causing unacceptable damage t in the NRC regulations and in Regulatory spent fuel in a storage pool or m the reactor Guides. Ilowever, there are a number of plants core during refueling. The revised SRP will pro-with construction permits and operating licenses vide the basis for implementmg addit onal re-ssued before the NRC's current regulations and quirements and procedures in existing plants regulatory guidance were in place. For this where warranted and can be used in future reason, re-reviews of the seismic design of reviews of new plants.
various plants are being undertaken (principally it is the NRC staff s view that continued as part of the Commission's Systematic Evalua-operation during review of this generic issue tion Program) to assure that these plants do not presents no undue risk to the health and safety present an undue risk to the public.
of the public. Operating facilities use a variety The NRC staff is conducting Generic Task of design and administrative measures to minimize the potential for dropping a heavy ob-440, as part of the NRC Program for Resolu-tion of Generic Issues. Task A40 is, in effect, a ject over the reactor core or over the spent fuel pool. These design and administrative measures compendium of short-term efforts to support the have been effective since no heavy load handling reevaluation of the seismic design of operating accidents resulting in damaged fuel have occur-reactors. The objective of the task is, in part, to red in over 300 reactor years of U.S. operating investigate selected areas of the seismic design experience. For facilities that have requested in-sequence to determine their conservatism for all creases in spent fuel pool storage capacity, the types of sites, to investigate alternate approaches NRC has prohibited the inovement of loads over to parts of the design sequence, and to quantify the overa!! conservatism of the design sequence.
fuel assemblies in the spent fuel pool that weigh more than the equivaler.t weight of one fuel In this manner the program will aid the NRC assembly. And for those plants where the review staff in performing its reviews of the seismic of the cask drop or the crane handling system is design of operating reactors.
A-16
Generic Task A-40 is separated into ten In response to these occurrences of BWR separate subtasks. The subtasks are described in primary system cracking, the NRC has taUn a the Task Action Plan for Task A-4, which is in-number of measures These actions included:
cluded in NUREG4371. Most of the subtasks issuance of Regulatory Guide 1.44 on e
are scheduled for completion in September 1979.
" Control of the Use of Sensitized However, three of the subtasks-related to Stainless Steel."
developing state-of-the-art methodology in order Issuance of Regulatory Guide 1.45 on to better define earthquake ground motion near earthquake sources-are longer term efforts.
". Reactor Coolant Boundary Leak Detec-tion Mems.,,
These three subtasks are scheduled for comple-Closely following the incidence of crack-tion in 1981.
ing in BWRs, including foreign ex-perience.
Pipe Cracks At Encouraging replacement of furnace-Boiling Water Reactors sensitized safe ends.
Requiring augmented in-service inspection Pipe cracking has occurred in the heat-(additional more frequent ultrasonic ex-affected zones of welds in primary system piping amination) of " service sensitive" lines, in boiling water reactors (BWRs) since the i.e., those that have experienced cracking.
mid-1960's. These cracks have occurred mainly Requiring upgrading of leak detection in Type 304 stainless steel, which is the type systems.
used in most operating BWRs. The major prob-lem is recognized to be intergranular stress cor-Pipe cracking and furnace sensitized safe end rosion cracking (IGSCC) of austenitic stainless cracking has been recently reported.n larger steel components that have been made suscepti-(24-inch diameter) lines in a BWlR (designed by ble to this failure by being " sensitized," either the General Electric Company) m Germ'my with by post-weld heat treatment or by sensitization ver 10 years of service. Because the safe ends of a narrow heat affected zone near welds.
n that facility had been furnace-sensitized dur-
" Safe ends" (short transition pieces between ing fabrication, IGSCC was suspected. One of the safe ends was removed for destructive ex-vessel nozzles and the piping) that have been highly sensitized by furnace heat treatment while amination. During laboratory examination of attached to vessels during fabrication were very the removed safe end, and also a small section early (late 1960's) found to be susceptible to f attached pipe, cracks were discovered at various locations in the safe end and m the weld IGSCC. Because of this, the Atomic Energy Commission took the position in 1969 that fur-heat affected zone of the pipe. The cracks m the nace sensitized safe ends should not be used on pipe weld area were very shallow, with the max-new applications. Most of the furnace-sensitized imum depth less than about 5 mm (about safe ends in older plants have been removed or 1/8-inch). Crackm, g m the furnace-sensitized safe clad with a protective material, and there are on-end was somewhat deeper. The German ex-ly a few BWRs that still have furnace-sensitized perience was the first known occurrence of IGSCC m, pipes as large as 24-inch m diameter.
safe ends in use. Most of these, however, are in smaller diameter lines.
In June 1978, a through-wall crack was Earlier reported cracks (prior to 1975) occur-discovered in an inconel recirculation riser safe red primarily in 4-inch diameter recirculation end (10-inch diameter) at the Duane Arnold loop-bypass lines and in 10-inch diameter core facility (see discussion under " Abnormal Occur-spray lines. More recently cracks were dis-rences-1978," in Chapter 7). The crack has covered in recirculation riser piping (12-inch to been attributed to IGSCC although the material 14-inch)in foreign plants. Cracking is most in this instance is different from the Type 3M often detected during inservice Inspection using stainless steel that has been historically found to ultrasonic testing techniques. Some piping cracks crack. Subsequent ultrasonic examination have been discowred as a result of primary discovered indications in some of the other seven coolant leaks.
A-17
The potential for stress corrosion crack-safe ends. Following their removal, cracking was discovered in all eight safe ends. The cracking ing in PWRs.
appeared to have originated in a tight crevice The significance of the safe end cracking between the mside wall of the safe end and an at Duane Arnold relative to similar mternal thermal sleeve. Such crevices are known material and design aspects at other to enhance IGSCC. Differences m matenals, fg; geometry, stress lesels and crevices appear to The Study Group is scheduled to complete its make the problem at Duane Arnold unique to a evaluation and report in January 1979. In addi-particular type of recirculation riser safe end tion to the Study Group effort, the NRC has (Type 1). As a result of this event, ultrasonic ex.
underway several generic technical review efforts amination of the other Type I safe ends in U.S.
which are aimed at improving piping inspection BWRs (i.e., at the Brunswick I and 2 facility) techniques and requirements. These generic ef-was conducted. No significant indications were f rts and any follow-on efforts resulting from found in Unit 2, and one indication was iden-the Stud" Group's evaluation will be incor-tified at Unit 1. Although this indication is p r ted into a new generic task, Task A-42, relatively minor and is not " reportable" pur-
" Pipe Cracks at Boiling Water Reactors.'
suant to the NRC Regulations, evaluation of it is continuing. The ultrasonic indication which was found was to be reevaluated at another Containment Emergency plant shutdown scheduled for later m 1978, after Sump Reliability the close of the report penod.
In discussions with General Electric (the reac-tor vendor) regarding recent pipe cracking ex-Following a postulated loss-of-coolant acci-perience, the company was asked by the NRC to dent, i.e., a break in the reactor coolant system provide an in-depth report on the significance of piping, the water flowing from the break would recent events regarding current inspection, be collected in the emergency sump at the low repair, and replacement programs. They were point in the containment. This water wou:d be also asked to address any new safety concerns recirculated through the reactor system oy the related to the occurrence of cracking in large emergency core cooling pumps to maintain core main recirculation piping. Based on information cooling. This water would also be circulated presented by the sendor and extensive staff through the containment spray system to remove evaluation, it was concluded that the recent oc.
heat and fission products from the containment.
currences did not constitute a basis for im.
Loss of the ability to draw water from the mediate concern about plant safety, nor require emergency sump could disable the emergency any new immediate actions by licensees.
core cooling and containment spray systems.
The staff briefed the Commission on pipe The consequences of the resulting inability to cracking in BWRs on August 31,1978, and on cool the reactor core or the containment at-September 14,1978, re-established an NRC Pipe mosphere could be melting of the core and/or Crack Study Group. The Study Group will breaking of the containment.
specifically address the following issues:
One postulated means of losing the ability to draw water from the emergency sump would be The significance of the cracks discovered blockage by debris. A principal source of such in large diameter pipes relative to the con-debris could be the thermal insulation on the clusions and recommendations set forth reactor coolant system piping. In the event of a in the referenced report and in its im-piping break, the subsequent violent release of piementation document NUREG-0313.
the high pressure water in the reactor coolant Resolution of concerns raised over the system could rip off the insulation in the area of ability to use ultrasonic techniques t the break. This debris could then be swept into detect cracks in austenitic stainless steel, the sump, potentially causing damage.
The significance of the cracks found in Currently, regulatory positions regarding large diameter sensitized safe ends, and sump design are presented in Regulatory Guide any recommendations regarding the cur-1.82, " Sumps for Emergency Core Cooling and rent NRC program for dealing with this Containment Spray Systems," which addresses matter.
A-18
the question of debris (insulation). The Regulatory Guide recommends that, in addition elevated to the highest priority as Generic Task to providing redundant separated sumps, two A-43, under the more general title of "Contain-protective screens be installed. A low approach ment Emergency Sump Reliability." Because th,s i
velocity in the vicinity of the sump is required to action has only recently been taken, a Task Ac-allow insulation to settle out before reaching the tion Plan and schedule for this task have not yet been developed.
sump screening; it is also required that the sump remain functional assuming that one-half of the screen surface area is blocked. The NRC staff Station Blackout believes that sump designs in accordance with this regulatory guide acceptably resolve this issue. Nonetheless, the NRC staff is continuing Electrical power for safety systems at nuclear to study the behavior of insulation under pipe power plants is supplied by two redundant and break conditions to gain a better understanding independent divisions. The systems used to of how it might behave.
remove decay heat to cool the reactor core A second postulated means of losing the abili-following a reactor shutdown are included ty to draw water from the emergency sump among the safety systems that must meet these would be abnormal conditions in the sump or at requirements. Each electrical division for safety the pump inlet-such phenomena as air entrain-systems includes an offsite alternating current ment, vortices, or excessive pressure drops.
(a. c.) power connection, a standby emergency These conditions could result in pump cavita-diesel generator a. c. power supply, and direct tion, reduced flow and possible damage to the current (d. c.) sources.
pumps.
The issue of station blackout was originally Currently, regulatory positions regarding included as Generic Task B-57 in the NRC Pro-sump testing are contained in Regulatory Guide gram for Resolution of Generic Issues. The task 1.79, " Pre-Operational Testing of Emergency involves a study of whether or not nuclear Core Cooling Systems for Pressurized Water power plants should be designed to accom-Reactors," which addresses the testing of the modate a complete loss of all a. c. power, i.e., a recirculation function. Both in-plant and scale loss of offsite a. c. sources and both onsite model test.s have been performed to demonstrate emergency diesel generator sources. Loss of all that circulation through the sump can be reliably
- a. c. for an extended period of time in pressuriz-accomplished. The NRC staff believes that ed water reactors, accompanied by loss of the sumps tested in accordance with this Regulatory auxiliary feedwater pumps (usually one of two Guide acceptably resolve this issue. As sup-redundant pumps is a steam turbine driven plemental guidance, the staff, through a contrac-pump that is not dependent on a. c. power for tor, is studying whether further guidance for the actuation or operation), could result in an in-design and review of emergency sumps to assure ability to cool the reactor core, with potentially adequate hydraulic design can be developed.
serious consequences. If the auxiliary feedwater The NRC staff initially planned to study the pumps are dependent on a. c. power to func-issue of containment emergency sump blockage tion, then a loss of all a. c. power for an extend-from insulation as part of Generic Task C-3, ed period could of itself result in an inability to
" Insulation Usage Within Containment." In ad-cool the reactor core. Although this is a low dition, initial plans were to study the vortex for-probability event sequence, it could be a signifi-mation issue as part of Generic Task B-18, cant contributor to risk.
" Vortex Suppression Requirements for Con-Current NRC safety requirements require as a tainments.' Howeser, containment emergency minimum that diverse power drives be provided sump operabHity is fundamental to the suc-for the redundant auxiliary feedwater pumps. As cessfu: operation of both the emergency core noted above, this is normally accomplished by cooling system (needed to cool the core) and the utilizing an a. c. powered electric motor driven containment spray system (needed to assure con-pump and a redundant steam turbine driven tainment integrity), following a loss-of-coolant pump. One concern is the design adequacy of accident. For this reason, these portions of plants licensed prior to adoption of the current Tasks C-3 and B-18 have been combined and requirements.
A-19
An initial survey of operating plants has been completed which indicates that all operating pressurized water reactors have either steam tur-bine driven or diesel driven auxiliary feedwater pumps (neither of which are dependent on a. c.
power). This assures at least that some capability exists for accommodating an extended loss of all
- a. c. power. Further review of older plants in this regard will be conducted as part of the NRC's Systematic Evaluation Program (see earlier discussion in this chapter). Further study will melude determining if any requirements beyond providing diverse power drives for the auxiliary feedwater pumps are needed-such as specific time requirements for the period during which the plant must be capable of accom-modating a station blackout.
This safety issue was previously included in the NRC Program for the Resolution of Generic issues as Generic Task B-57, but has recently been elevated to the highest priority as Generic Task A-44. Because this action has only recently been taken, a Task Action Plan and schedule for this task have not yet been developed.
A 20
APPENDIX B GENERIC ISSUE GROUPING BY TYPE OF ACTIVITV Group 1 - Analyzirg generic technical problems related to plant safety that have arisen from new information such as experience from operating reactors, identified design or construction defi-ciencies, or research and test results, for the purpose of (1) determining whether existing safety requirements or review procedures require upgrading and developing new requirements or procedures if needed, and/or (2) developing technical po-sitions regarding the acceptability of long term solutions for operating plants.
Lead Task No.
Title Di vision A-1 Water Hammer DSS A-2*
Asymmetric Blowdown Loads on the Reactor Vessel DDR A-3 Westinghouse Steam Generator Tube Integrity DDR A-4 Combustion Engineering Steam Generator Tube Integrity DDR A-5 Babcock & Wilcox Steam Generator Tube Integrity D0R A-6 Mark I Short Term Program (Complete)
DDR A-7 Mark I Long Term Program DOR A-8 Mark II Program DSS A-10 BWR Nozzle Cracking DDR A-ll Reactor Vessel Materials Toughness D0R A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports D0R A-13 Snubbers DSS A-16 Steam Effects on BWR Core Spray Distribution DOR A-26 Reactor Vessel Pressure Transient Protection (Overpressure) (Complete)
DOR A-35 Adequacy of Offsite Power Systems DDR This task includes Task B-52, Fuel Assembly Seismic and LOCA Responses.
Group 1 (Continued)
B-2 Lead Task No.
Title Division A-39 Determination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containments 055 B-9 Electrical Cable Penetration of Containment DSS B-10 Behavior of BWR Mark III Containment DSS B-18 Vortex Suppression Requirements for Containment Sumps DSS B-32 Ice Effects on Safety-Related Water Supplies DSE B-34 Occupational Radiation Exposure Reduction DSE B-47 Inservice Inspection of Supports and Bolting of Class 1, 2, 3 and MC Components DOR B-48 BWR Control Rod Drive Mechanical Failures DOR B-55 Improved Reliability of Target-Rock Safety Relief Valves DOR B-56 Diesel Reliability DOR C-7 PWR System Piping DOR C-8 Main Steam Line Leakage Control Systems DDR C-12 Primary System Vibration Assessment DDR
B-3 Group 2 - Evaluating exist'r; safety requirements and review procedures in areas related to plant safety that could reqJire upgrading.
Lead Task No.
Title Division A-9 ATWS DSS A-21 Main Steam Line Break Inside Containment DSS A-29 Design Features to Control Sabotage DOR A-31 RHR Shutdown Requirements (Complete)
DSS A-36 Control of Heavy Loads Near Spent Fuel DOR B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containments DSS B-8 Locking-Out of ECCS Power Operated Valves D0R B-12 Containment Cooling Requirements (Non-LOCA)
DSS B-21*
Core Physics DSS B-22*
LMFBR Fuel DSS B-26 Structural Integrity of Containment Pentrations DSS B-57 Etation Blackout DSS B-63 Isolation of Low Pressure Systems Connected to the RCPB (Complete)
D0R B-68 Pump Overspeed During a LOCA DSS B-69 ECCS Leakage Ex-Containment DSE B-70 Power Grid Frequency Degradation and Effect on Primary Coolant Pumps DSS B-73 Monitoring for Excessive Vibration Inside the Reactor Pressure Vesse!
DSS Tasks B-21, B-22 and B-23 contain activities that fall into a number of the groups.
B-4 Group 3 - Performing studies to confirm the adequacy of current staff safety requirements.
Lead Task No.
Ti tle Division A-14 Flaw Detection DSS A-17 Systems Interaction in Nuclear Power Plants DPM A-18*
Pipe Rupture Design Criteria DSS A-22 PWR Main Steam Line Break - Core and Primary Coolant Boundary Response (MSLB Outside Con-tainment)
DSS A-30 Adequacy of Safety-Related DC Power Supplies DSS A-32 Evaluation of Overall Effects of ' Missiles DSS B-4 ECCS Reliability DSS B-14 Study of Hydrogen Mixing Capability in Contain-ment Post-LOCA DSS B-17 Criteria for Safety-Related Operator Actions DSS B-54 Ice Condenser Containments DSS B-58 Passive Mechanical Failures DSS B-61 Analytically Derived Allowable ECCS Equipment Outage Periods D0R B-66 Control Room Infiltration tieasurements DSE C-1 Assurance of Continuous Long Term Capability of Hermetic Seals on Instrumentation and Electrical Equipment DPM C-2 Study of Containment Depressurization by Inad-vertent Spray Operation to Determine Adequacy of Containment External Desige 0 essure DSS This task includes Task B-16, Protection Against Postulated Diping Failures in Fluid Systems Outside Containment.
Group 3 (Continued)
B-5 Lead Task No.
Title
- Divisior, C-3 Insulation Usage Within Containment DSS C-4 Statistical Methods for ECCS Analysis DSS C-6 LOCA Heat Sources DSS C-9 RHR Heat Exchanger Tube Failures DOR C-10 Effective Operation of Containment Sprays in a LOCA DSE Call Assessment of Failure and Reliability of Pumps and Valves D0R D-1 Advisability of a Seismic Scram DPM D-3 Control Rod Drop Accident (Complete)
B-6 Grc.. 4 - Performing studies for the purpose of quantif ing safety margins provided by current requirements or detcrmining whether or not current safety requirements car. be relaxed.
Lead Task No.
Title Division A-25 Non-Safety Loads on Class IE Power Sources DSS A-38 Tornado Missiles DSE A-40 Seismic Design Criteria - Short Term Program DPM B-3 Event Categorization DSS B-6 Loads, Load Combinations, Stress Limits DSS C-5 Decay Heat Update
B-7 Group 5 - Developing, maintaining or irproving staff capabilities to perform independent calculations and audits for safety and environmental reviews. Also performing generic audit re-views and calculations.
Lead Task No.
Title Division A-24 Qualification of Class IE Safety-Related Equipment DSS B-2*
Forecasting Electricity DSE B-7 Secondary Accident Consequence Modeling DE B-ll Subcompartment Standard Problems DSS B-13 Marviken Test Data Evaluation DSS B-15 CONTEMPT Computer Code Maintenance DSS B-19 Thermal-Hydraulic Stability DSS B-20 Standard Problem Analysis DSS B-24 Seismic Qualification of Electrical and Mechanical Equipment DSS B-25 Piping Benchmark Problems DSS B-28 Radionuclide/ Sediment Transport Program DSE B-31 Dam Failure Model DSE B-33 Dose Assessment Methodology DSE B-35 Confirmation of Appendix I Models for "Calcula-tion of Releases of Radioactive Materials in Gaseous and Liquid Ef fluents from Light-Water-Cooled Power Reactors" DSE B-41 Impacts on Fisheries DSE B-44 Forecasts of Generating Costs of Coal and Nuclear Plants DSE B-46 Costs of Alternatives in Environmental Design DSE B-65 Iodine Spiking DSE This task includes Task B-45, Need for Power-Energy Conservation.
Group 5 (Continued)
B-8 Lead Task No.
Ti tl e Division B-72 Health Effects and Life-Shortening from Uranium and Coal fuel Cycles DSE C-14 Storm Surge Model for Coastal Sites DSE C-16 Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection DSE
B-9 Group 6 - Improving guidance to applicants, licensees and/or staff reviewers regarding staff safety and environmental require-ments or developing documentation describing the basis for staff safety and environmental requirements.
Lead Task No.
Title Division A-15 Decontamination DDR A-19 Digital Computer Protection Systems DSS A-23 Containment Leak Testing (Complete)
DSS A-27 Reload Application Guide D0R A-28 Increase in Spent Fuel Storage Capacity DOR A-34 Instruments for Monitoring Radiation and Process Variables During Accidents DSE A-37 Turbine Missiles DSS B-1 Environmental Technical Specifications DSE B-27 Implementation and Use of Subsection NF DSS B-29 Effectiveness of Ultimate Heat Sinks DSE B-30 Design Basis Floods and Probability DSE B-36 Develop Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems D3E B-38 Reconnaissance Level Investigations DSE B-39 Transmission Lines DSE B-43 Value of Aerial Photographs for Site Evaluation DSE B-49 Inservice Inspection Criteria and Corrosion Pre-vention Criteria for Containments D0R B-50 Post-0perating Basis Earthquake Inspection DDR B-51 Assessment of Inelastic Analysis Techniques for Equipment and Components D0R B-S3 Load Break Switch DSS
Group 6 (Continued)
B-10 Lead Task No.
Title Division B-59 N-1 Laop Operation in BWRs and PWRs 00R B-60 Loose Parts Monitoring Systems B-62 Re-examination of Technical Bases for Establishing SLs, LSSSs, and Reactor Protection System Trip Functions D0R B-64 Decommissioning of Reactors D0R B-67 Effluent and Process Monitoring Instrumentation DSE C-15 NUREG Report for Liquid Tank Failure Analysis
- DSE C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes DSE
B-11 Group 7 - Performing studies related to staff environ ental reviews as necessary to (1) address new infomation, (2) determine whether or not current staff requirements can be relaxed, and (3) con-firm the adequacy of current staff requirements.
Task No.
Title A-20 Impacts of Coal Fuel Cycle DSE A-33 NEPA Reviews of Accident Risks DSE B-37 Chemical Discharges to Receiving Waters DSE B-40 Effects of Power Plant Entrainment on Plankton DSE B-42 Socioeconomic Environmental Impacts DSE
B-12 Group 8 - Other tasks that do not fit the definitions of Categories 1-7.
Lead Task No.
Ti tle Division D0R B-71 Incident Response (This task is not related to individual li-censing actions.
It involves improving NRC's capabilities to respond to emergencies at operating facilities.)
C-13 Non-Random Failures DSS (This task is included within the scope of Tasks A-9, A-17, A-30, A-35, B-56 and B-57.)
D-2 ECCS Capability for Future Plants DSS (This task is included within the scope of the RES p ogram to develop improv.ed safety features.
APPENDIX C RISK-BASED CATEGORIZATION OF GENERIC TASKS Category I:
Potential High Risk Items NRR Category A A9 ATWS A6/A7/A8/A39/855 Mark I, Il programs, SRV pool dynamic loads and temperature limits, Target Rock Valve Reliability A17 Systems Interactions A40 Seismic Design Criteria A29 Design Features to Control Sabotage A10 BWR Nozzle Cracking A24 Qualification of Class IE Safety-Related Equipment NRR Category B B57 Station Blackout Requirements B63 Isolation of Low Pressure Systems connected to RCPB B30 Design Basis Floods and Probability B34 Occupational Radiation Exposure Reduction NRR Category C C3 Insulation Usage within Containment (suni, blockage)
Category II:
Potential Low Risk Items NRR Category A A3/4/5 W, B&W, CE Steam Generator Tube Integrity Al Water Hammer A12 Fracture Toughness of SG/RCP Supports A2 Asymmetric Blowdown Loads A30 DC Power Supplies A15 Decontamination
C-2 Category III Negligible Risk Potential ( )
NRR Category A A13 Snubbers All Reactor Vessel Toughness A14 Flaw Detection A26 R.V. Pressure Transient Protection A18 Pipe Rupture Design Criteria A22 MSLB PWR-Core and RCS Response (material toughness)
A32 Overall Missile Effects A37 Turbine Missiles A38 Tornado Missiles A28 Increase in Spent Fuel Storage Capacity A21 MSLB inside containment A36 Heavy Load Control Near Spent Fuel A33 NEPA Review of Accident Risks (e.g., 3-8 accidents)
A31 RHR Shutdown Requirements A16 Steam Effects on Core Spray Distribution NRR Category B B68 Pump Overspeed during LOCA B70 Power Grid Frequency Degradation and Effect on Primary Coolant Pumps 84 ECCS Reliability B5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containment B6 Loads, Load Combinations, Stress Limits B7 Secondary Accident Consequence Modeling B8 Locking out of ECCS Power-0perated Valves B9 Electrical Cable Penetrations of Containment B13 Marviken Test Data Evaluations B18 Vortex Suppression Requirements for Containment Sumps B19 Ther sl-Hydraulic Stability B28 Radionuclide/ Sediment Transport Program 829 Effectiveness of Ultimate Heat Sinks B31 Dam Failure Model B32 Ice Effects on Safety-Related Water Supplies B48 BWR CRD Mechanical Failures 849 Inservice Inspection Criteria for Containment B54 Ice Condenser Containments B59 Review of (N-1) Loop Operation in BWRs and PWRs B60 Loose Parts Monitoring Systems B61 Allowable ECCS Equipment Outage Periods B62 Re-examination of Technical Bases for Establishing SLs, LSSSs, Etc.
B65 Iodine Spiking
C-3 NRR Category B (Continued)
B66 Control Room Infiltration Measurements B67 Effluent and Process Monitoring Instrumentation B69 ECCS Leakage Excontainment B73 Monitoring for Excessive Vibration Inside the PWR and BWR Reactor Vessel Bl4 Study of Hydrogen Mixing Capability in Containment Post-LOCA B50 Requirements for Post-0BE Inspection B12 Containment Cooling Requirements (non-LOCA)
NRR Category C Cl Assurance of Continuous Long-Term Integrity of Seals on Instrumentation and Electrical Equipment C2 Adequacy of Containment External Design Pressure C7 PWR System Piping C8 Main Steam Line Leakage Control Systems C9 RHR Heat Exchanger Tube Failures C10 Effective Operation of Containment Sprays in LOCA Cl2 Primary System Vibration C15 NUREG Report for Liquid Task Rupture Analysis NRR Category D DI Advisability of Seismic Scram D3 CRD Accidents Category IV:
Not Directly Relevant to Risk (e.g., Procedural)
NRR Category A A34 Instruments for Monitoring Radiation and Process Variables During Accidents A35 Adequacy of Offsite Power Systems A19 Digital CPS A25 Non-Safety Loads on Class IE Power Sources A27 Reload Application Guide A23 Containment Leak Testing A20 Coal Fuel Cycle NRR Category B B1 Environmental Technical Specifications B2 Forecasting Electricity Demand by State in the United States on an Annual Basis
C-4 NRR Category B (Continued)
B3 Event Categorization B10 Behavior of BWR Mark III Containment Bll Subcompartment Standard Problems B15 CONTEMPT Computer Code Maintenance B17 Criteria for Safety-Related Operator Actions B20 Standard Problem Analysis B21 Core Physics B22 LWR Fuel B23 LMFBR Fuel B24 Seismic Qualification of Electrical and Mechanical Components B25 Piping Benchmark Problems B26 Containment Penetrations B27 Implementation and Use of Subsection NF B33 Dose Assessment Methodology B35 Confirmation of Appendix I models for " Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors" B36 Develop Design, Testing and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units for ESF Systems and for Normal Ventilation Systems B37 Chemical Discharges to Receiving Waters B38 Reconnaissance Level Investigations B39 Transmission Lines B40 Effects of Power Plant Entrainment on Plankton B41 Impact on Fisheries B42 Socioeconomic Enivronmental Impacts B43 Value of Aerial Photographs for Site Evaluation B44 Forecasts of Generating Costs of Coal and Nuclear Plants B46 Costs of Alternatives in Environmental Design B47 Inservice Inspection Criteria for Supports and Bolting of Class 1, 2, 3, and MC Components B51 Assessment of Inelastic Analysis Techniques B5i Load Break Safety Switch B56 Diesel Reliability B58 Passive Mechanical Failures B71 Incident Management B72 Development of Models for Assessing Risk of Health Effects and Life Shortening from Uranium and Coal Fuel Cycles NRR Category C C4 Statistical Methods for ECCS Analysis C5 Decay Heat Update C6 LOCA Heat Sources Cll Assessment of Failure and Reliability of Pumps and Valves
C-5 NRR Category C (Continued)
Cl4 Storm Surge Model for Coastal Sites C16 Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection C17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes NRR Category D D2 ECCS Capability for Future Plants
APPENDIX D EVENTS REPORTED TO CONGRESS AS ABNORMAL OCCURRENCES Event Description Affected Plants Calender Year 1975 Steam Generator (S/G) Tube Failure Fire Point Beach 1 in Electrical Cable Trays Loss of Main Browns Ferry 1&2 Coolant Pump Seals Improper Control H.B. Robinson 2 Rod Withdrawals Dresden 2 Quda-Cities 1 Cracks in Pipes at Boiling Water Various BWRs Reactors Fuel Channel Box Wear at Boiling Various BWRs Water Reactors S/G Feedwater Flow Instability at Various PWRx Pressurized Water Reactors Calender Year 1976 Occupational Whole Body Overexposure Zion 1 Indian Point 2 Failure of Undervoltage Tri-Logic and Millstone 2 Consequent Loss of Safeguary AC Power Core Power Distribution Anomaly St. Lucie 1 S/G Tube Integrity Various PWRs Inadvertent Criticality Millstone 1 Feedwater Nozzle Cracking in BWRs Various BWRs
D-2 Event Description Affected Plants Calender Year 1977 Breach of Physical Security System Ft. St. Vrain Fuel Rod Failures Lacrosse Management and Procedural Control Zion 1&2 Deficiencies Generic Design Deficiency (Net North Anna 1&2 Positive Suction Head)
Surry 182 Beaver Valley 1 Environmental Qualifications All Plants Safety-Related Electrical Equipment Inside Containment Calender Year 1978 Insulation Failures in Containment Millstone 2 Electrical Penetrations Fuel Assembly Control Rod Guide Millstone 2 Tube Integrity Overexposure of Two Radiation Trojan Protection Technicians Degraded Primary Coolant Boundary Duane Arnold in a BWR
U.S. NUCLE AR REGUL ATORY COMMISSION (7 77)
BIBLIOGRAPHIC DATA SHEET NUREG-0510
- 4. TITLE AND SUBTITLE (Add Volume No., d moropnate)
- 2. fleave blank)
Identification of Unresolved Safety Issues Relating 3 RECIPIENT S ACCESSION NO.
to Nuclear Power Plants -- A Report to Congress 7 AUTHOR (S)
- 5. D ATE HE PORT COMPLE TED MONTH YEAR
- 9. PE RFOHMING ORGANIZATION N AME AND M AILING ADDHFSS (include lip code)
DATE REPORT ISSUED l YEAR MONTH Program Support Staff Office of Nuclear Reactor Regulation 6 ILeave blan* /
Washington, D. C.
20555 8 (Leave blank)
- 12. SPONSORING ORGANIZATION N AME AND M AILING ADDRESS (loclude 2,0 Codel 10 PHOJE CT/TASA/ WORK UNIT 40.
- 11. CONTRACT NO.
- 13. TYPE OF REPORT PE RIOD COVE RE D (/nclusive dates)
Congressional Report
- 15. SUPPLEMENTARY NOTES 14 (Leave 6/ank)
- 16. ABSTR ACT (200 words or less)
In accordance with Section 210 of the Energy Reorganization Act of 1974, as amended, the NRC transmitted a report, NUREG-0410 to Congres.s on December 30, 1977, describing the NRC program for the resolution of generic issues. The report, however, pointed out that the NRC program was considerably broader than the " Unresolved Safety Issues Plan" re-quired by Section 210.
The NRC program, included plans for the resolution of generic environmental issues, for instance. Over the past year, the NRC has undertaken a review of the generic issues addressed in the NRC program to determine which issues qualify as " Unresolved Safety Issues" for future reporting to Congress. This report describes this review and identifies the " Unresolved Safety Issues" discussed in the 1978 NRC Annual Report.
- 17. KEY WOHDS AND DOCUMENT AN ALYSIS 17a. DESCRIPTORS 17tt IDENTIFIE RS/OPEN ENDE D TERMS
- 18. AV AILABILITY STATEMENT
- 19. SE CUHITY CLASS (Th,s report)
- 21. NO. Of P AGE S Non-classified Unlimited
- 20. SE CU RI TY CLAS3 (Th,s pace) 22 PRICE S
NRC FORM 335 (7-77)
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