ML20217M109

a Prioritization of Generic Safety Issues

ML20217M109
Person / Time
Issue date:	03/31/1998
From:	Emrit R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-0933, NUREG-0933-S22, NUREG-933, NUREG-933-S22, NUDOCS 9804080010
Download: ML20217M109 (398)

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SUPPLEMENT 22 TO NUREG-0933 "A PRIORITIZATION OF GENERIC SAFETY ISSUES" REVISION INSERTION INSTRUCTIONS Remove Insert introduction: pp.1 to 28, Rev. 5 pp.1 to 28, Rev. 6 pp. 29 to 62, Rev. 21 pp. 29 to 62, Rev. 22 '

i Section'1: pp.1.0-1, Rev.1 pp.1.0-1, Rev. 2 i pp.1.1.A.1-1 to 2, Rev. 2 pp.1.l.A.1 'i to 2, Rev. 3 '

pp.1.l.A.2-1 to 23, Rev. 5 pp.1.1.A.2-1 to 22, Rev. 6 pp.1.1.A.3-1 to 9, Rev. 5 pp.1.1.A.3-1 to 9, Rev. 6 pp.1.1.A.4-1 to 11, Rev. 5 pp.1.l.A.4-1 to 11, Rev.6 l pp.1.1.B.1-1 to 9, Rev. 3 pp.1.1.B.1-1 to 9. Rev. 4 )

pp.1.1.B.2-1 to 5 pp.1.1.B.2-1 to 5, Rev.1 f^N pp.1.1.C-1 to 7, Rev. 3 pp.1.1.C-1 to 8, Rev. 4 l Q pp.1.l.D-1 to 14, Rev. 7 pp.1.1.E-1 to 8, Rev. 2 pp.1.1.D-1 to 14, Rev. 8 pp.1.1.E-1 to 8, Rev. 3 l

pp.1.1.F-1 to 7, Rev. 2 pp.1.1.F-1 to 7, Rev. 3 l pp.1.1.G-1 to 5, Rev. 2 pp.1.1.G-1 to 5, Rev. 3 pp.1.ll.A-1 to 3, Rev.1 pp.1.ll.A-1 to 3, Rev. 2 pp.1.II.B-1 to 14, Rev. 3 pp.1.II.B-1 to 14, Rev. 4 pp.1.II.C-1 to 9, Rev. 2 pp.1.ll.C-1 to 10, Rev. 3 l pp.1.11.D-1 to 3, Rev.1 pp.1.ll.D-1 to 4, Rev. 2 I pp.1.ll.E.1-1, Rev.1 pp.1.ll.E.1-1 to 2, Rev. 2 l pp.1.ll.E.2-1 to 7, Rev.1 pp.1.II.E.2-1 to 7, Rev. 2 pp.1.II.E.3-1 to 3, Rev.1 pp.1.ll.E.3-1 to 3, Rev. 2 pp.1.11.4.4-1 to 11, Rev.1 pp.1.ll.E.4-1 to 11, Rev. 2 Section 3: pp. 3.78-1 to 5, Rev. 2 pp. 3.78-1 to 5, Rev. 3 pp. 3.119-1 to 6, Rev. 2 pp. 3.119-1 to 7, Rev. 3 pp. 3.142-1 to 9, Rev. 3 pp. 3.142-1 to 9, Rev. 4 i pp. 3.163-1 pp. 3.166-1, Rev.1 pp. 3.166-1 to 2, Rev. 2 pp. 3.173-1 to 4 pp. 3.173-1 to 4, Rev.1  !

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pp. A.F-1 to 2 Glossary pp. G 1 to 6, Rev. 2 pp. G-1 to 7, Rev. 3 O

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INTRODUCTION II. BACKGROUND .

dia100t On October 8,1976, the Commission directed the staff to develop "a program plan for resolution of generic issues and completion of technical projects." The Commission .

1 further requested that "this plan should include: task schedules ... task priority and ,

manpower requirements (with proportions of staff contract efforts explicitly identified)."

On December 12,1977, the Energy Reorganization Act of-1974 was amended by Congress through Public Law 95-209 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 action 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 thereafter.

In order to meet both Commission and Congressional directives, the staff developed a generic issues program that provided for.the-identification of generic issues, the assignment of priorities, the development of detailed action plans, projections of dollar and manpower costs, continuous high level management oversight of progress, and y public dissemination of information related to the issues as they progressed. This program 'was published in NUREG-0410"7 in January 1978 and, shortly thereafter, the Commission issued a Policy Statement"" on the NRC " Program for Resolution of Generic issues Related to Nuclear Power Plants."

The NRC generic issues program published in NUREG-0410"7 was considerably broader

. than the " Unresolved Safety issues Plan" required by Section 210. It included plans for the resolution of generic environmentalissues, for the development of improvements in the reactor licensing process, and for consideration' of less conservative design criteria or operating limitations in areas where existing requirements might be unnecessarily restrictive' or costlyk U f Ths 'first'inttempts by~ the~ staff. to implement the generic issues program stated in

, 3 ' NUREG-0410"7 were based largely on engineering judgments. This qualitative' effort to ji ,

rank" unresolved generic issues continued through two phases:

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Revision 6 (1) in 1977, alt issues were classified into four categories according to importance, from "significant" to "little or no importance."

(2) In the early part of 1978, the issues were reclassified into Groups 1 through 8 by type rather than by order of importance.

Later in 1978, the staff began to take a quantitative approach by using risk assessment to place the issues into four categories ranging from I (potential high risk items) to IV (items not directly relevant to risk)." With increased confidence in this risk assessment l approach, the staff introduced a more comprehensive quantitative system in early 1979.

Points were assigned to each issue based on an assessment of safety significance, environmental significance, licensing effectiveness, deadline pressure, and retrofit versus forward-fit. Although the point system was still quite subjective,it was nevertheless a major improvement over the previous methods used.

In the aftermath of the Three Mile Island Unit 2 (TMI-2) accident, many new generic issues were raised and the staff came to the conclusion that the point system was too subjective to be used for ranking the issues. One of the TMI Action Plan" items. IV.E.2, called for the staff to develop a plan for the early resolution of safety issues. It was in resolving this issue that the staff developed a quantitative "prioritization" methodology whereby a numerical priority score could be assigned to each generic safety issue. With this approacn, priorities were to be based on an evaluation of the estimated risk reduction associated with the potential change in requirements that could result from resolution of an issue and the estimated costs to the NRC and the industry in implementing such a change. This methodology was submitted to the Commission for information in SECY-81-513.' in April 1983, this approach was refined and resubmitted to the Commission for approvalin SECY-83-221."** After Commission review, approval to use the methodolo0y was given in December 1983."" l In April 1993, after approximately ten years of experience with the methodology, adjustments were made in the numerical thresholds, while retaining the basic features of the method. These adjustments involved raising risk thresholds and simplifying the way in which costs entered the priority rankings. What motivated the raising of risk thresholds was the observation"" that, of the issues resolved, only 3 of the 27 MEDIUM-priority and about half of the HIGH-priority issues resulted in decisions to take regulatory action, i.e., in retrospect, it appeared that resources had been devoted to resolving a large number of issues with no resulting safety improvement. This outcome must be interpreted with the qualification that generic issue resolution efforts that have not led to regulatory action have, nevertheless, in many instances, produced safety benefits through licensee actions taken voluntarily, in consideration of the issues raised, or in response to interim guidance. However, the extent of these benefits, when they occurred, was generally in proportion to the priority rank and MEDIUM-priority issues usually resulted in marginal improvements. The proposed revisions were submitted to the Commission in SECY-93-108""; in July 1993, Commission approval was obtained."

The threshold adjustments were intended to cause the prioritization process to model the resolution process without the earlier, apparently excessive margin for initial uncertainties, to reduce resolution efforts that do not produce safety improvements, 12/31/97 2 NUREG-0933 k.

R; vision 6 (m) while still ensuring attention to issues that require it. The raising of the numerical safety thresholds is accompanied by strengthened attention to uncertainties and special

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considerations, to help recognize instances when a priority rank higher than the indication from the new numerical formula is warranted, the objective being to improve the efficiency of the prioritizations without impairing their prudence.

The priority ranking chart and risk thresholds used in prioritization analyses completed before July 24,1993, are shown in Appendix C.

The r,implification of the way in which costs entet .eflects the confirmation from experience that risk significance is indeed the priniary factor in priority ranking, with a more bounded role for safety-cost trade-offs.

,Operatina Plan The initial work in prioritizing issues was essentially done by various Staff Working Groups. Following a reorganization of the Office of Nuclear Reactor Regulation (NRR) in April 1980, the lead responsibility for prioritization was assigned to the Safety Program Evaluation Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation (SPEB/ DST /NRR).

The 1983 NRC Policy and Planning G 4 dance (NUREG-0885, issue 2),2' in addressing the area of Coordinating Regulatory Requirements (Planning Guidance, item 5, p.6) n called for "...a priority list of generic safety issues including TMI-related issues based on V)

( the potential safety significance and cost of implementation of each issue..." to be submitted to the Commission for approval. Using the prioritization methodology outlined below, this list was developed by SPEB in response to the Planning Guidance and forwarded to the Commission in SECY-83-221.""

After another NRR reorganization in November 1985, the task of preparing and maintaining the list of generic safety issues and their priority was assigned to the Safety Program Evaluation Branch, Division of Safety Review and Oversight (SPEB/DSRO/URR).

Following an NRC reorganization in April 1987, this responsibility was assigned to the Advanced Reactors and Generic issues Branch, Division of Regulatory Applications, Offica of Nuclear Regulatory Research (ARGIB/DRA/RES). In July 1991, this responsibility was transferred to the Division of Safety issue Resolution (DSIR) in RES.

With the elimination of DSIR in December 1994, this function was transferred to the Generic Safety issues Branch (GSIB), Division of Engineering Technology (DET), RES.

The prioritization of generic issues is an ongoing staff function that has been reflected annually in the NRC Policy and Planning Guidance.2' This document was superseded in 1987 by the NRC Five-Year Plan.

11. GENERIC ISSUES PROGRAM After issuance of the Policy Statement"' in 1978, the NRC program to resolve generic issues underwent many reviews and changes. As a result, the Commission concluded

'~'N in April 1989 that the 1978 Policy Statement no longer reflected the NRC's generic j issuedprogram and withdrew it from the public record." The current generic issues 12/31/97 3 NUREG-0933

R3 vision 6 program consists of six separate and distinct steps: identification, prioritization, resolution, imposition, implementation, and verification (See Exhibit A). An explanation of each of these six steps is given below.

Identification Generic concerns may be identified by individuals or organizations within the NRC staff or by the Advisory Committee on Reactor Safeguards (ACRS), the nuclear power industry, or the public. RES Office Letter No. 7 (OL #7)"" provides a procedure and suggested content for individuals or organizational units within the NRC to request consideration of a concern as a new generic issue. This procedure may also be used by parties outside the NRC to express their concerns to the staff for consideration as potential generic issues. Sources of potential generic issues are many and varied and include, but are not limited to, the following: evaluation of safety-related research, risk assessment analyses, and public and industry concerns.

Prioritiration This report focuses on the prioritization step of the generic issues program which is explained in detail in Paragraph lli below.

Resolution After an issue has been prioritized and approved for resolution, the first task is the development of a plan to delineate the work to be done, assignment of major responsibilities, identification of project resource needs, and scheduling of milestone dates. These activities vary in scope and depth in accordance with issue priority and the depth of information on a given issue. The second task involves development of a technical solution. Typically, the information used to resolve an issue comes from experience data, experiments, tests, analyses, and probabilistic risk assessments (PRAs).

The results of such work or the technical findings may be published in contractor and staff NUREG reports which are made available through the NRC Public Document Room j (PDR), Washington, D.C., or the National Technical Information Service, Department of Commerce, Springfield,. Virginia.

In the final stage of resolution, the technical findings are used as a basis to develop a proposed resolution for the issue involving a change to NRC requirements or guidance.

Several alternatives may be considered. A regulatory analysis, including a detailed cost / benefit analysis of each practical alternative, and consideration of the best methods of imposition, implementation, and verification are used in selecting a proposed i resolution. If a backfit is proposed, first, a determination is made as to whether the i backfit is required to provide adequate protection to the health and safety of the public or simply provides for enhancement of public health and safety. If it is determined that the backfit is necessary to provide an adequate level of protection, the backfit will be imposed regardless of the costs to achieve it. If it is determined that the backfit provides for enhancement of public health and safety, a generic analysis is required that treats the nine factors specified in 10 CFR 50.109(c).

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Rsvisi:n 6 j Exhibit A ff ERIC ISSUES PROGRAM IDENTIFICATION 1P PRIORITIZATION 1P RESOLUTION

,f IMPOSITION 1f IMPLEMENTATION

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Revision 6 Once the cognizant NRC Office Directors have agreed to a proposed resolution, it is then forwarded to the Committee for the Review of Generic Requirements (CRGR), the ACRS, the Executive Director for Operations (EDO), and the Commission for review and approval as appropriate. Changes to regulations, Policies, the Standard Review Plan (SRP), and Regulatory Guides are published in the Federal Register for public comment.

Comments received are then incorporated, as appropriate, with the final product published in the Federal Register, Resolution of a generic issue can take from several months to a few years depending on the length of time required by the deliberations involved at each of the above steps.

OL #7'8 describes the procedure to be followed in the resolution of a generic issue, denotes the required elements of the resolution plan and resolution package, and identifies review procedures and organizational responsibilities for the approval of the resolution of a generic issue. [ Prior to June 2,1994, this procedure was issued separately in RES Office Letter No. 3 (OL #3)""; however, OL #3 became obsolete'"'

when it was merged with OL #1.] Milestone information and reporting requirements as well as organizational responsibilities for the tracking of generic issue resolution are also required by OL #7 (Prior to June 16,1996, these functions were outlined in RES OL

1. 1.) Allissues scheduled for resolution are tracked through the resolution process by the Generic Issue Management Control System (GIMCS) which is updated quarterly and placed in the PDR. Guidance for the preparation, review, and required content of the regulatory analysis portion of the resolution package is provided in RES Office Letter No.

3C.""(Prior to February 23,1996, these procedures were outlined in RES Office Letter l

No. 2."")

Imoosition f imposition is the step in the generic issues program where each affected licensee and/or applicant is required or guided to prepare a schedule for implementing the generic issue resolution consistent with a Rule, Policy, Regulatory Guide, generic letter, bulletin, f and/or licensing guidance developed during the resolution stage. Normally, NRC requirements, policies, and/or guidance will not provide for NRC consideration of a licensee's modifications prior to their implementation at an affected plant. This facilitates completion of plant modifications to enhance safety within two refueling outages, not to exceed three years after issuance of NRC requirements, policies, and/or guidance. However, in a few exceptional cases, licensees may be expected to submit (normally for NRC approval) their plans (including schedules) for plant modifications prior to their implementation. In all cases, licensees will be expected to certify in writing to '

the NRC that plant modifications have been completed.

i For the exceptional cases, the staff reviews each app:icant's and/or licensee's submittal with regard to proposed modifications to site, equipment, structures, procedures, technical specifications, operating instructions, etc., and schedules proposed for the accomplishment of the rnodifications. For backfits, imposition is complete when each affected licensee has committed to compliance actions and schedules for imp;ementing these actions. For forward-fitr,, the imposition of a generic issue resolution is complete when the new requirement or guidance becomes effective as an integral part of NRC regulations, policies, and/or guidance.

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Revision 6 i

pq During this stage, a resolved GSI is identified as a Multiplant Action (MPA) for licensee (V) action. The imposition status of all MPAs is tracked in the Safety issue Management System (SIMS).

Imolementation l

implementation is the step in the generic issues program where the affected licensees perform the actions on existing plants to satisfy the commitments made during the imposition stage. These may include modifications / additions to equipment, structures, procedures, technical specifications, operating instructions, etc. No later than 30 days after each affected licensee has completed all of the actions required for a particular generic issue resolution, and the modified / additional system is fully operational, the licensee is required to certify in writing to the NRC that plant modifications have been completed in accordance with NRC requirements, policies, and/or guidance. When all affected licensees have officially notified the NRC of completion of all l required / committed actions, the implementation stage is complete, unless it is l determined by the staff from subsequent verification inspection that additional licensee actions are needed for compliance. l Verification The verification step consists of three parts. First, the portions of a licensee's actions, l if any, that warrant NRC inspection must be determined. This decision is made during O the resolution stage based on the judgment of the safety significance of the issue i

) relative to other matters in the inspection program, licensee performance, and the resources needed to accomplish a meaningful inspection. Next, as necessary, inspection instructions are prepared to ensure that the NRC inspection is performed in a consistent and appropriate manner at all affected plants; the inspection, by its very nature, is an audit. Therefore, carefully thought-out instructions must be provided to the NRC inspectors so that the maximum safety benefit is achieved for the limited resources devoted to this effort. The third part of the verification process is the actual verification i and documentation of the results in an inspection report. Physical inspections are l performed on an sudit basis in a manner consistent with generalinspection procedures which involve a sampling of changes made by licensees or applicants, as opposed to a 100% inspection of all actions. Verification of licensee implementation of generic issue resolution is reported by the staff in SIMS.

111. PRIORITIZATION i

Pornose and Scone The primary purpose of prioritization is to assist in the timely and efficient allocation of resources to those safety issues that have a high potential for reducing risk and in decisions to remove fiom further consideration issues that have little safety significance and hold little promise of worthwhile safety enhancement. However, issues of such gravity that consideration of immediate action is called for are excluded from prioritization because of the compressed time scale in which decisions for such issues e

j must be made. Generally, immediate action takes the form of a Bulletin or Order. Both

~\ ,/ operating and future plants are considered in the priority ranking process.

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Revision 6 Prioritization focuses on generic safety issues (GSis) i.e., safety concerns that may affect the design, construction, or operation of all, several, or a class of nuclear power plants and may have the potential for safety improvements and promulgation of new or revised requirements or guidance. However, the method can be used to identify changes in current requirements that could significantly reduce the impact (usually cost) on licensees without any substantial change in public risk. Issues of this type a e classified as Renulatorv Imnact issues (RI) to clearly differentiate them as not improving the safety of nuclear power plants but, nevertheless, possibly worthwhile.

In order to identify GSis, allissues originated in accordance with OL #1"'8 are reviewed to determine their safety significance. Issues that primarily concern environmental protection or the licensing process and do not involve significant safety improvement elements are classified accordingly and noted for separete consideration outside the GSI priority ranking scheme. These issues are classified as either environmental issues or licensing issues. Environmental issues (El) involve impacts on the human environment and the values sought to be protected by the National Environmental Policy Act (NEPA).

Ucensing issues (LI) are not directly related to protecting public health and safety or the environment, but relate to: (1) increasing the staff's knowledge, certainty, and understanding of safety issues in order to increase its confidence in assessing levels of safety; (2) improving or maintaining the NRC capability to make independent assessments of safety; (3) establishing, revising, and carrying out programs to identify and resolve GSis: (4) documenting, clarifying, or correcting existing requirements and guidance; and (5) improving the effectiveness or efficiency of the review of applications.

The list of issues subjected to prioritization contains the following groups:

(1) TMI Action Plan items identified for development in NUREG-0660"; these issues are covered in Section 1. The priority recommendations in this report exclude ]

those issues that were designated for implementation in NUREG-0737.**

(2) Task Action Plan items identified in NUREG-03718 and NUREG-0471,' plus the subsequently added issues A-42 through A-49 that were designated as Unresolved Safety issues (USis); these issues are covered in Section 2. However, l

issues designated as USIs were excluded from prioritization because of the high-priority attention they were given based on priority decisions previously made.

In the future, USls will come from issues that have been prioritized.

(3) New Generic issues identified by the staff, ACRS, or others; these issues are covered in Section 3. All new issues identified will be prioritized and included in Section 3 and published in future supplements to this report.

(4) Human Factors Program Plan (HFPP) items identified for development in NUREG-0985' 3; these items are covered in Section 4.

(5) Chernobyl issues identified in NUREG-1251"'5; these issues are covered in Section 5.

A comprehensive listing of all issues in the above five groups is given in Table 11 which includes the following information for each issue: (1) the NRC person responsible for the l

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Rsvision 6 prioritization evaluation; (2) the lead NRC office, division, and branch responsible for V , reviewing the prioritization analysis and/or resolving the issue; (3) the priority ranking or status; (4) the latest version of the evaluation; (5) the issuance date'of the latest

, version of the evaluation; and (6) the MPA number for those issues that have been resolved and require licensee actions. .A summary of the number of inues in each

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category is shown in Table 111. A cross-reference listing of reports prepared by the Office for Analysis and Evaluation of Operational Data (AEOD) and their corresponding generic issues is provided in Table IV.

How tha Work la Done The work is rione, in accordance with the criteria described below, by the responsible

'NRC Branch in consultation with others in the NRC with knowledge of the issues or expertise in the technical disciplines involved. In a number of instances, technical or cost information is obtained from industry and other outside sources. The Battelle Pacific Northwest Laboratories (PNL), under a technical assistance contract, developed detailed methods to quantify safety benefits and costs and provided safety benefit analyses and cost information for many of the issues. The responsible NRC Branch, with internal consultations as necessary, reviews and applies the PNL-supplied technical factors, in conjunction -with additional factors, in developing the priority rankings and recommendations.

Systematic peer review of each prioritization evaluation within the NRC contributes to the assurance that the analysis is complete and accurate and that the judgments are soundly based. This review is done in two stages. First, each analysis is reviewed by the NRC organizational unit or units whose area of responsibility or specialized knowledge

~is substantially involved. Second, any comments made are then resolved, where practical, and factored into the analysis, as appropriate. Upon completion of peer review, the analysis is then finalized and prepared for approval by the responsible Office Director. Once approved, it is placed in the PDR and published in a future supplement to this report, after which, additional comments from the ACRS, the industry, and the public are considered in any further reassessment of the issue's priority.

Priority Cataoorian Their Meanina and Pronomad Uma Four priority rankings are used: HIGH, MEDIUM, LOW, and DROP. They are intended for

. use in guiding allocation of NRC resources and scheduling of efforts to resolve the various issues, in conjunction with other pertinent factors such as:-(1) the nature, extent,' and availability of manpower and material resources estimated to be required;

. (2) length of time needed to resolve; (3) conflicts in resource allocation and scheduling among. items of comparable priority; (4) status of affected reactors; and (5) budget constraints.

A Hlfati' priority' ranking means that strong efforts to achieve the earliest practical resolution are appropriate. This is because: (a) an important safety concern may be

involved (though generally the concern is not severe enough to require prompt plant shutdown); or (b) the uncertainty'of the safety assessment is unusually large and an i

^ upper-bound risk assessment would indicate an important safety ' concern. All unresolved 1

HIGH priority issues are' periodically reviewed in accordance with the criteria stated in-4

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NUREG-0705 f" possible designation as ult.. A USlis defined as a rnatter affecting a number of nuclear power plants that poses im;artant 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 accordance with Section 210 of the Energy Reorganization Act of 1974, progress on the resolution of USis is reported to Congress in each NRC Annual Report.

A MEDIUM priority ranking means that no safety concern demanding high-priority attention is involved, but there is believed to be potential ix safe +y improvements or reductions in uncertainty of analysis that may be substantial and worthwhile. Efforts at resolutN should be planned, perhaps over the ensuing years, but on a basis of not interfering with pursuit of HIGH-priority generic issues or other high-priority work.

A IflE priority ranking means that no safety concerns demanding at least MEDIUM-priority attention are involved and there is little or no prospect of safety improvements that are both substantial and worthwhile. When the prioritization process results in a LOW priority ranking for an issue, approval of this aanking by the responsible Office Director signifies that the issue has been eliminated from further pursuit. Hoever, in accordance with Staff Requirements Memorandum (SRM) 871021 A,"88 the staff

) conducis a periodic review of existing LOW-priority GSis to determine whether there is any new information that would necessitate reassessment of the original prioritization evaluations.

The DROP category covers proposed issues that are without merit or whose significance is clearly negligible. Issues are also DROPPED from further consideration if it is determined that their safety concerns have been addressed in previously prioritized or resolved issues. When the prioritization process results in a DHOP priority ranking for an issue, approval of this ranking by the responsible Office Director signifies that the issue has been eliminated from further pursuit.

An issue is considered PESOLVED, indicated by NOTE 3 in Table 11, when its resolution has resulted in either: (a) the establishment of regulatory requirements or guidance (by Rule, SRP" change, or equivalent); or (b) a documented authoritative decision that no change in requirements is warranted. Priority rankings are not assigned to issues that have been resolved. However, in those cases where issues were resolved after having been identified for further pursuit by the prioritization process, the related calculations have been retained in the text of this document for future use.

Priority rankings are not assigned to issues that are neartv-resolved (denoted by NOTES 1 and 2 in Table 11) because approval of changes to requirements, based on the resolution of an issue, requires that a detailed value/ impact evaluation of the safety bcnefit. implementation costs, and other relevant factors be made. Prioritization would duplicate this value/irnpact analysis, but in a less comprehensive manner. Therefore, the effort that would be needed to prioritize an issue is devoted to completing the final evaluation of the issue, rather than making a tentative judgment as to the importance and value of the issue, Possible resolution of an issue is considered to be identified, indicated by NOTE 1 in Table 11, when a possible technical resolution is under evaluation and the evaluation is nearing completion. Further work may be required as part of the 12/31/97- 10 NUREG-0933

r Revisior: 6 m review and approval process before a change in requirements or guidence is issued.

/ i

(,) Resolution of an issue is considered ava!:abee, i.@wted by NOTC ^* 'o Table 11, when

' proposed or recommended changes to requirements oc guidance are documented in a NUREG report, NRC memorandum, Safety Evaluation Report (SER), or equivalent.

Priority rankings are also not assigned to those issues whose safety concerns are deti ' mined to be covered (at the time of prioritization) in other issues of broader scope l that ce being prioritized or are being resolved. Issues in this category are integrated into l

th: issues of broader scope. A detailed listing of all such issut 4 is given in Table V. <

Criteria For Assionino Priorities

1. Basic Anoroach The method of assigning priority rank involves two primary elements: (i) the estimated safety importance of % ksue; and (ii) the estimated cost of developing and implementing a resolution. Special considerations may influence the proper use of the estimates. These eternents are applied as follows:

(a) The issue is identified and defined. Since issues are often complex and interrelated with other issues, careful definition of an issue's scope and l bounds is essentialin arriving at a sound and applicable assessment.

n (b) A quantitative estimate is made of the safety importance of the. issue, measured in terms of the risk (the product of accident probabilities and (V) radiological consequences) attributable to the issue and the decrease in that risk that may be attainable by resolving the issue.

(c) A quantitative estimate is made of the cost of resolution.

(d) A numerical impact /value ratio is calculated by dividing the estirr.ated cost entailed by the estimated potential risk reduction. The do measures the safety value received in return for the cost impact incurred.

(e) A priority rank (HIGH, MEDIUM, LOW, or DROP) is obtained by application of criteria in which both the safety significance of the issue and the impact /value ratio are taken into account. The ratio is not always directly applied to determine the priority rankings. In some cases, the safety '

l significance of the issue is so great that it demands a HIGH priority, or so minor inst only a LOW priority (or a decision to DROP) is wuranted irrespective of the impact /value assessment.

o (f) The prioity ranking is reviewed and modified, if appropriate, in light of any w Lial factors (discussed below) that: (i) might bring int 3 question the ppi!c:.bility of the necessarily simplified calculation technique; and (ii) call for special consideration of NRC management decisions or large uncertainties in the quantitative estimates.

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R3 vision 6 in summary, while the method has a quantitative emphasis, the edcv!sted numerical velues are used as an aid to judgment and not as ,

deurminative of the ranking results. The nature of the specific issue, the ]

quality of G. date base, and the scope of the necessarily limited analysis j determine in each w the dependability of the numericalindications as a judgment aid.

2. Safetv Sionificance The safety significance of an issue is represented by the reduction in risk that resolution could effect. Risk is ordinarily expressed here in terms of the product of the frequency of an accident occurrence and the public dose (in person-rem) that would result in the event of the accident. If more than one accident scenario is important within the necessarily rough risk estimates, the risks are summed.

The potential risk reduction calculated in this way is used in calculating the impact /value ratio as part of the simplified impact /value analysis, discussed in Paragraph 111.3 below. It is also used directly as a measure of safety significance, as discussed in Paragraph 111.4 below, in arriving at a priority rank that is influenced by the safety significance of an issue as well as by the estimated value/ impact relation of a projected solution, or is determined on the basis of safety significance alone.

The person-rem-based risk reduction estimate may not be the only appropriate measure of an issue's safety significance in all cases. For example, when a possible core damage is involved but release outside containment would be minor or highly improbable, contribution to the core-damage probability may well be more indicative of safety significance. Provision is made, as described in Paragraph 111.4 below, for use of alternative measures of safety significance in determining a priority ranking when such alternative measures are useful.

3. Imonct/Value Relation (a) The Imonct/Value Ratio Formula To the extent reasonably possible, quantitative estimates are made of the possible solutions to a GSI by calculating an impact /Value Ratio that reflects the relation between the risk reduction value expected to be achieved and the associated cost impact. The formula for the impact /value ratio (R) is:

R = Cost Safety Benefit wi ere the safety benefit is the estimated risk reduction (event frequency x Sublic dose averted) that may be achieved, and the cost is that thought necessary to develop and implement a resolution in the number of plants involved. The scoring computation for any issue is then:

12/31/97 12 NUREG-0933 w ___ - -__-__-___-_ _ __ . -

R , vision 6

()/

[ R = _C._.

NFTD where, N = number of reactors involved T = average remaining life (years) of the affected plants, based on an original license period of 40 years F = the accident 1,equency reduction (event / reactor-year)

D = public dose from the isdioactive material released from condnment (person-rem)

C = total cost 'of developing end implementing the lasciution of the toue for all plants affected (dollars).

The total cost (C) includes both the cost of developing the generic solution, typically NRC cost, and the cost of implementing the possible solution at all affected plant, typically industry cost, including design, equipment, installation, test, operation, and maintenance. The priority ratio (R) has the units of dollars per person-rem.

Simplified calculations usually suffice, since only an approximate impact /value ratio is required. Reference should be made to the current version of the Value-impact Handbook

• where necessary to supplement g the general guidelines provided below.

(b) Rationale for the Formula The qualitative diversity of factors entering impact /value analyses in support of GSI prioritization, together with inevitable quantitative uncertainties, make any of various possible impact /value score formulas necessarily imperfect. Accordingly, provisions are made to compensate  ;

for those imperfections to the extent practical (as discussed in Paragraph j 111.5 below).:

The formula selected measures a total-cost / total-safety-benefit relation.

As dir, cussed herein, it is applied within limits set by other possible considerations where a safety issue is either too important to depend on safety-cost tradeoffs or too trivial to merit attention at all. Two principal arguments favor a formula of this type-1 (1) The' denominator is designed as a direct measure of the safety -

values that it is NRC's primary mission to protect. The. numerator

'is designed to measure the overall cost impact, including industry as well as NRC costs, and should thus reflect the entire public i interest in economy. The resulting impact /value ratio should, subject to the stated caveats, reasonably approximate measuring ,

L [,. <' '

the overall public interest'in safety value received for. total A _)j resources expended.

12/31/97.. 13 NUREG-0933 l 1

i e  :

R: vision 6 L

(2) The allocation of national resources, which in most cases are primarily industry resources, is optimized.

(c) Risk Estirnates The risk estimates developed for GSis are useful as rough approximations for comparative purposes, but are not necessarily applicable to the assessment of absolute levels of risk attributable to particular issues.

Similarly, the impact /value ratios provide, for the limited purpose of prioritization, tentative assessments of relative potential for cost-effective resolution. They are not intended to be applied as impact /value determinations for any regulatory proposal that may ultimately result from efforts to resolve en issue. In addition, the assumed resolutions are not intended to prejudg > tiic final resolutions, but are only assumptions that I are necessary t pr.rform quantitative analyses.

l The basis of frequency estimates generally involves the following:

(1) Identificatinn of the specific events which are the basis for the concern, for which the consequences are to be established, and which are to be eliminated or ameliorated by a proposed technical solution (2) Use of event sequence diagrams, fault trees, or decision trees, if possible (3) Identified references and calculations, or stated assumptions for l the numbers used l (4) Consideration of the probability of common mode as well as random independent failures.

Where possible, numerical estimates are made based on operating i

experience, usually Licensee Event Reports (LERs). Other sources include prior PRAs and other risk and reliability studies. Some numbers are based on engineering judgment;in such cases, the basis for that judgment is

! stated. ]

l i

l For the identified end event (s), the expected radiological consequences are expressed in person-rem generally based on the radioactive release categories described in WASH 1400 (Appendix VI, pp. 2-1 to 2 5), j reproduced as Appendix A to this report. Exhibit B gives estimated Curies released and approximate population doses for each release category. The computer proDram CRAC2, applied to a typical midwest site (Braidwood) meteorology, was used for the dose calculations. However, the calculated l doses were adjusted to reflect the mean of the population density within L a 50-mile radius of U.S. nuclear power plants." Assumptions and

! parameters used for the calculations at this stage (Step (b) described ,

under " Basic Approach")were as follows:

12/31/97 14 NUREG-0933

i R::visi:n 6 i i

f" (d

) - Consequences are represented by the whole body population dose (person-rem) received within 50 miles of the site. i; i

An exclusion area of 1/2 mile was assumed with a uniform j population density of 340 persons per square mile beyond ]

1/2 mile. This is the mean 50-mile radius population density projected for the year 2000 (NUREG-0348, p.T52)."

Evacuation of people was not considered because of the possible large variations in evacuation capability for each plant site.

!1

- All exposure pathways were included in the basis of the tabulated numbers except ingestion pathways, i.e., interdiction of contaminated foods was assumed. (Farmland usage parameters for the State of Illinois were used for separate ingestion pathway calculations where made.)

Meteorological data was taken from the U.S. National Weather l Service station at Moline, Illinois. l ll The person-rem factors for each release category are given in Exhibit B.

Although generally used, consequence estimates were not solely based on these factors. Other factors were used in some cases when more  ;

appropriate.

]

\ )

I An estimated occupational dose of 20,000 person-rem from postaccident cleanup, repair, and refurbishment is also considered.

Where significant occupational radiological exposure (ORE) is incurred or averted in implementing current requirements or the proposed resolution of a GSI, such exposure is taken into account but stated separately.

Where more direct issue-specific ORE information is lacking, dose estimates are obtained by assuming an average dose rate of 2.5 millirem / hour (based on the PNL analysis" cited above) and multiplying by the estimated number of man-hours involved.

A second factor is that the risk associated with an issue is more likely to be overestimated than underestimated. Where risk estimates are widely uncertain, a reasonably conservative value of risk reduction is generally selected to help assure adequate priority to issues that may warrant attention.

The sum of the estimated risks of all the separate issues wi!! likely exceed the present estimate of the total risk of nuclear power plants because of two factors. First, individual accident sequences can be affected by more than one ist,ue. The resolution of one issue would reduce the probability or consequences of a certain set of accident sequences. Some or even all of these sequences could be the same as some or even all of the

/ sequences affected by another issue. However, issues are assessed 12/31/97 15 NUREG-0933

I Revision 6 l

l independently and this interaction of their risk significance is not ordinarily considered. This interaction is strongest for issues related to human factors, since human error affects almost all sequences. The sum of the reductions in core-melt frequency estimated for all of the human factors-related issues may be as much as twice as great as the total human factors contribution to total risk. However, most of the issues not related to human factors are much less strongly interrelated.

Exhibit B Estimated Release Release Public Dose **

Category (Curies) (Person-rem)

PWR-1 1.2 x 10' 5,400,000 PWR-2 9.3 x 10' 4,800,000 PWR-3 5.2 x 10' 5,400,000 PWR4 2.8 x 10 8 2,700,000 PWR-5 1.3 x 10 8 1,000.000 PWR-6 1.0 x 10' 150,000 PWR-7 2.1 x 10' 2,300 PWR-8

• 7.7 x 10 5 75,000 PWR9* 1.1 x 10' 120 BWR-1 1.1 x 10' 5,400,000 BWR-2 1.1 x 10' 7,100,000 BWR-3 5.0 x 10' 5,100,000 BWR-4 2.1 x 10 8 610,000 BWR-5' 1.7 x 105 20

• Non-core-melt (Other release categories involve core-melt).

The Release value (Curies) and Estimated Public Dose (Person-rem) will be updated in the future to be consistent with the ongoing evaluation to revise the Source Term following a postulated severe accident.

(d) Cost Estimates Because cost estimates are used here only in relation to risk estimates which are generally subject to more or less wide uncertainties, only approximate costs are needed.

No separate estimates are generally made for offsite property damage; reasonably conservative use of the public dose estimates is an adequate 12/31/97 16 NUREG-0933

_ __O

1 .

Ravision 6

/ -

. surrogate in this application. Furthermore, there is no readily-available

data on offsite damage that is realistic and detailed enough to make estimates meaningful, reasonably accurate, and generically applicable. If unusual or special offsite effects are not adequately represented by the public dose in some issues, this fact will be considered separately and explicitly in evaluating such issues.

The expected technical solution on which the cost estimate is based is identified. Estimated costs are established by collecting available data regarding engineering, procurement, installation, testing, and periodic inspection and maintenance. Where data are non-existent, estimates are based on judgments by the experts involved. Assumptions and estimated uncertainties are identified. Costs are estimated in 1982 dollars.

NRC costs include the following: (1) issue identification, analysis, resolution, and report issuance; (2) research to establish proposed I specific changes to licensing requirements (or to determine that no change ~ is required); (3) technical assistance contracts (including associated NRC effort); (4) discussions and correspondence with industry owners' groups; (5) plant reviews; and (6) preparation and review of SERs and . requirement documents. The estimated ecst of NRC professional time is based on $100,000 per person-year.

The costs to industry generally consist of some combination of the following: (1) licensing; (2) design; (3) equipment procurement; (4) installation; (5) testing, inspection, monitoring, and periodic maintenance;

.[ .L ' and (6) plant downtime to effect a change, taken as the cost of 1 replacement power at $300,000/ day. Industry manpower costs are ordinarily taken as $100,000 per person-year.

Averted plant damage costs may affect the priority of a GSI. Estimates for such averted costs are multiplied by the accident frequency and used as negative costs, i.e., subtracted from the (positive) costs of implementing the resolution of the issue."" The averted costs may include those of averted equipment failures, limited time plant outage, or limited plant-contamination cleanup. In the extreme, they can also include averted permanent loss of the plant, estimated at approximately $2 billion present worth. This' estimate for a " generic" plant includes the costs of both plant-wide cleanup and permanent loss of use of the plant, discounted to present worth based on a 7% real discount rate. This figure is multiplied in'ench case by the reduction in frequency of such events

.. that would be brought about by resolution of the GSI.' The plant loss estimate includes allowance for typical plant. age at the time of the 3 accident as well as replacement power costs together with apportioned J g cost of'a replacement plant. The plant-wide cleanup estimate reflects -

cleanup to the point at which the plant is ready for decommissioning or

. refurbishing' for ' restart.'" . Refurbishing costs, when restart is more A) d v

a v4 economical than decommissioning, would depend on the nature of the t

12/31/979 ,

17- NUREG-0933

, 4

~ ' '

. . . . . , , . . , , , . . . , . . . . , . . . , . . . . . . ..,m. .. . . , , . . . . ,

3 R; vision 6 accident and could range from a fraction of the total plant loss figure to a cost approaching that figure.

Some fixed costs are one-time, initial costs; others may occur at future times. Future costs are discounted to present worth at a 7% rate. Where costs are continuous or periodically recurring throughout a plant's remaining life, the periodic cost is taken into account using an approximation of the present worth of the continuing (or repetitive) costs for plants with remaining operating lives of 20 years or longer.

(e) Uncertaintv Bounds Major sources of uncertainty in the priority score are identified and judgments as to their quantitative significance are indicated as information warrants. Where data warrant, the method described in NUREG/CR-2800," Section 5, for the general case of combining uncertainties for random variables with unknown distributions (as well as some special cases) are used. [See also Paragraph ill.5(a)). Most often, however, a rigorous uncertainty analysis is not warranted, in most cases, the uncertainty in the point estimates of risks and costs is known to be large. However, sufficient information is not usually available to make a meaningful quantitative analysis of the uncertainty bounds of these point estimates. Decisions are tempered by the knowledge that the uncertainty is generally large. This knowledge was also used in developing the chart of tentative priority rankings (Figure 1). The wide spread between a level of risk, for example, at which an issue would be ranked as having a high priority and the level at which an issue would be ranked as low priority (a factor of 100) is partially based on the recognition that the uncertainties are large. In cases where uncertainty has a special character or importance, this is discussed and considered in the conclusion of the analysis of the GSI.

4. Priority Ranking (a) Priority Rankina Chart A chart showing how the tentative priority rankings are derived from the safety significance of an issue and its impact /value ratio is presented in Figure 1. The thresholds on the chart are discussed in Paragraphs Ill.4(b)  !

and Ill.4(c) below. A conversion factor of $1,000/ person-rem was used  !

until September 18,1995, when an increase to $2,000/ person-rem was approved by the Commission.'

(b) Preliminarv Screening for Safety Significance The determination of a priority rank starts with a triage based on safety significance, i.e., the incremental risk associated with the issue. For a 4

reduction in core damage frequency (ACDF) greater than 10 per reactor-year (RY), a HIGH priority is assigned on the basis of safety importance 12/31/97 13 NUREG-0933 j

i

. c.

Rsvision 6 1

alone, regardless of other considerations, such as an initially estimated

. high cost, which might result in a low priority score.

1 At the other extreme, an issue's safety significance could be too minor I to warrant diversion of attention from more important safety issues even if it has a low impact /value ratio because an inexpensive solution is

believed to be available. Below a minimal safety significance threshold, -

the priority would always be DROP; where the potential risk reduction is 3

- trivial,'there can be no basis for regulatory action on safety grounds.

'In between, there may be issues of less . extreme importance or unimportance, for which a HIGH, MEDIUM, LOW, or DROP priority may be appropriate, based on consideration of the impact /value relation as well as safety significance. As indicated in Figure 1, a HIGH priority may be assigned to an issue exclusively on the basis of a high safety significance; the threshold shown on the chart is ACDF= 104 /RY. For an issue with a safety significance lower than the threshold for an always-HIGH priority but at least 10% of that threshold (ACDF= 10 /RY), the chart indicates a HIGH or MEDIUM priority based on cost trade-offs. At the low-risk end of the abscissa, the priority rank indicated is always DROP for ACDF<103 /RY. Cost trade-offs enter in the 104 to 10 d/RY ACDF range, as discussed in Section 4(c) below.

The abscissa in Figure 1 provides a measure of an issue's estimated safety significance in terms of the change (A) in CDF attributable to resolution of the issue. This is often the most useful safety significance measure in GSI prioritization, though for some issues other measures may be required or appropriate. For example, a measure based on radiological consequences (probability-averaged over the remaining reactor life) is  ;

used when the issue under consideration involves containment bypass or -

relates to containment performance or other features or actions to mitigate the radiological consequences of a core damage. Also, the thresholds may need to accommodate the possible influence of the number of reactors. affected on the appropriate priority ranking.

Therefore, Figure 1 is repeated in Figure 2, with auxiliary abscissae providing additional measures of safety significance. These are used when the principal abscissa is inapplicable, or when an auxiliary abscissa leads to a higher priority indication. Thus, the abscissae for total effect on all plants are considered when more than 30 plants are affected.

'(c)- ImnactNalue Ratio Thresholds When the safety. significance is in the intermediate range discussed above,i.e., ACDF between 104 and 104/RY, or between 0.1 % and 100%

of the threshold for an always-HIGH priority, the impact /value ratio (R) is

~

' taken into' account in the ranking indicated by the chart (Figure 1). This

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RGvision 6 (1) in the range of 10% to 100% of the threshold for an always-HIGH priority, the indicated priority is HIGH if R is below $2,000/ person- l rem; otherwise, the indicated priority is MEDIUM.

(2) In the range of 1% to 10% of the threshold for an always-HIGH priority, the. indicated priority is MEDIUM if R is below

$2,000/ person-rem; otherwise, the indicated priority is LOW. l (3) in the range of 0.1% to 1% of the always-HIGH threshold, the indicated priority is LOW or DROP, depending on whether R is below or above $2,000/ person-rem. l

5. Other Considerations The formula-based rankings represent the primary concern of the NRC: public safety. The secondary concern is the impact on licensees, evaluated in terms of cost. However, the tentative priority rankings are subject to the limitations of an often incomplete and imprecise data base and to possible distortions due to the nature of the necessarily highly simplified quantitative formula underlying them.

Special situations with respect to some issues niay cause added difficulty in priority assignment. While the formula-based tentative rankings ger.arally indicate that the safety significance is sufficient to justify NRC action, other considerations not adequately reflected, or not reflected at all, in the numerical formula are often needed to corroborate or adjust the results. Decision-making is helped by explicit identification of such other considerations and explanation of how they bear on the resulting final priority ranking, whether the effect is one of corroborating or of changing the estimates.

Listed below are some factors that rnay be important in arriving at a sound priority ranking and may lead to adjustment of a tentative, formula-derived ranking. Possible effects of occupational doses and uncertainty bounds [(a)(1),

(a)(2), and (b)(1) below] require particularly careful consideration for all issues.

The factors listed are not considered all-inclusive. Others thought significant are discussed and, when practical, quantified appropriately in the overall risk significance measure and impact /value ratio along with their associated uncertainties. Sometimes, there are special considerations that are quite specific to an issue or some aspect of it. However, it should be noted that, in determining an issue's priority, those factors that relate to safety are given the most consideration. The following is a partial list of other factors to be considered:

(a) Special risk and cost aspects not included in or potentially masked by the numerical formulas:

(1) The additional risk associated with a license renewal period of 20 years for the affected plants. GSis prioritized and resolved up to March 31,1994 were evaluated for license renewal implications; these evaluations were documented in NUREG/CR-5382*' and an 12/31/97 22 NUREG-0933

Revision 6 k

CN RES report." All other GSis prioritized and resolved after March

(,) 31,1994 were required to consider the impact of license renewal.

(2) The net change in occupational doses entailed by implementing the current versus the proposed requirements.

(3) Any significant non-radiation-related occupational risk affected by the proposed resolutions.

(4) Loss or severe degradation of a layer in the defense-in-depth concept (e.g., one mode of core cooling or containment cooling)

(5) Issues for which solutions of widely differing costs may be applicable to different classes of plants or various plants are otherwise affected in vastly different ways.

(b) Factors related to uncertainties stemming from an incomplete or imprecise data base for the priority formula:

)

a (1) Uncertainty bounds, imbalance in uncertainty factors, certainty of  ;

cost to fix versus uncertainty that safety is really improved and )

the true extent of such improvement.

n (2) Situations where uncertainty is extraordinarily large (in accident

[ ) probability, consequences, or cost, or any or all of these). If there V are large uncertainties in either the numerator or the denominator, the mean of the impact /value ratio (mean ratio) should be used with caution in assigning a priority ranking. The ratio of the means is a good approximation to the mean ratio provided only that the uncertainty in the denominator is small. However, if the uncertainty in the denominator is large, then the ratio of the 7

means is a poor estimate of the mean ratio.

(3) Problems which are ill-defined and problems for which solutions are not evident so that at least the resources necessary to understand the problem are assigned.

(4) The potential for a proposed change to affect more than one accident or transient sequence, thus affecting risk to a greater or lesser degree than assessed in the description of the issue; notably, the potential for a new safety decrement, or increase in risk, due to unidentified effects of a proposed change, or added complexity, or for other reasons.

(5) Circumstances imparting unusual significance to accident consequences (such as ingestion pathway effects) or mitigating measures (such as evacuation) that are not directly included in the

, , public dose calculations.

y

]

12/31/97 23 NUREG-0933 L _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -

l Rxvision 6 ;

(6) Potential for human intervention, using available equipment. l (7) Acute knowledgeable professional controversy concerning the importance of an issue or modes of dealing with it.

(c) Change with passage of time:

(1) The effect of license renewal should be considered in every prioritization. The effect, if any, on the priority rank of an additional 20 years of operation should be separately stated.

(2) Potential substantial deterioration of the impact /value ratio while awaiting regulatory resolution (e.g., a potential design fix that is inexpensive to apply before construction, much more expensive after the plant is largely built, and extremely expensive and problematical to apply to an operating plant).

(3) The amount of resources already spent on an issue, and how close to completion it may be; the value of continuity in efforts to resolve an issue.

(4) The span of time predicted to resolve an issue and implement the resolution.

(5) The clarity of an " issue" and the objectivity with which it is currently defined. (Perhaps additional research effort is necessary to identify and define a specific risk reduction of interest.)

(6) Change of perceptions (of safety importance or impact /value relation or some special issue-peculiar factor) in the course of time.  ;

Generally, in situations of large doubt or conflicting indications, the highest priority rank reasonably consistent with the nature of an issue is assigned. Thus, where no solution is evident, assignment of a priority consistent with the safety significance of the issue may lead to a search for resolution or mitigation at an acceptable cost. Generally, should uncertainties narrow or perceptions change in the course of time, the priority rankings can be reexamined in the light of new developments and retained or changed. When different classes of plants are expected to be very differently affected by a potential resolution, the piiority assignment is governed by the class of plants for which resolution is most worthwhile and urgent. (Resolution in such cases can involve a new requirement for some class of plants and no action for others.) Where resolution differs for different classes of plants, differing priorities may be assigned.

6. Concludino Remarks The criteria and estimating process on which the priority rankings are based are neither rigorous nor precise. Considerable application of professional judgment, 12/31/97 24 NUREG-0933

w - gr Revision 6 sometimes guided by good information but often tenuously based, occurs at a

..b; number of stages in the process when numerical values are selected for use in the formula calculations and when other considerations are taken into' account in corroborating or changing a priority ranking. What is important in the process is that it is systematic, that it is guided by analyses that are as quantitative as the situation reasonably permits, and that the bases and rationale are explicitly stated, providing a " visible" information base for decision. The-impact of imprecision is blunted by the fact that only approximate rankings (in only four broad priority categories) are necessary and sought.

IV. RESULTS OF PRIORITIZATION

~ The results of the prioritization and resolution of all issues contained in this report are summarized and tabulated by group in Table 111. In addition, a listing of those issues that

- affect operating and future plants is given in Appendix B. This appendix reflects the results of prioritization and resolution and only includes: (1) issues that have been resolved with new requirements (NOTE 3(a)l; (2) USI, HIGH, and MEDIUM priority issues that are being resolved; (3) nearly-resolved issues (NOTES 1 and 2); (4) issues that are scheduled for prioritization and whose impact is not yet known (NOTE 4); and (5) issues that were resolved without requirements for operating plants but with staff requirements for future plants under development.

REFERENCES

1. SECY-81513, " Plan for Early Resolution of Safety issues," August 25,1981.

2. NUREG-0371, " Task Action Plans for Generic Activities (Category A)," U.S. Nuclear Regulatory Commission, November 1978.

p 3. NUREG 0471, " Generic Task Problem Descriptions (Categories B, C, and D)," U.S.

Nuclear Regulatory Commission, June 1978.

11. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," U.S. Nuclear Regulatory Commission (1st Edition) November 1975, (2nd Edition) March 1980, (3rd Edition) July 1981.

16. . WASH-1400 (NUREG-75/014), " Reactor Safety Study, An Assessment of Accident Riskr.

s in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission,

- October 1975.

' 44.  :

NUREG-0705, " Identification of.New Unresolved Safety issues Relating to Nuclear Power Plant Stations," U.S. Nuclear Regulatory Commission, March 1981.

48. . NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S.

Nuclear Regulatory Commission, May 1980, (Rev.1) August 1980.

, .64. NUREG/CR-2800, " Guidelines for Nuclear ~ Power Plant Safety issue Prioritization

Inforrration Development," U.S. Nuclear Regulatory Commission, February 1983, t

~ 12/31/97 . ~25; NUREG-0933

.y

Revision 6 (Supplement 1) May 1983, (Supplement 2) Decernber 1983, (Supplement 3) September l 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.

I

70. NUREG-0348, " Demographic Statistics Pertaining to Nuclear Power Reactor Sites," U.S.

Nuclear Regulatory Commission, November 1979.

98. NUREG-0737, " Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory Commission, November 1980, (Supplement 1) January 1983.

140. Memorandum for R. Frahm from R. Emrit, " Summary Report on a Risk Based Categorization of NRC Technical and Generic issues," June 30,1989.

186. NUREG-0510, " Identification of Unresolved Safety issues Relating to Nuclear Power Plants," U.S. Nuclear Regulatory Commission, January 1979.

210. NUREG-0885, issue 2, "U.S. Nuclear Regulatory Commission Policy and Planning Guidance," U.S. Nuclear Regulatory Commission, January 1983.

387. NUREG-0410, "NRC Program for the Resolution of Generic issues Related to Nuclear Power Plants," U.S. Nuclear Regulatory Commission, January 1978.

1 393. "TMI-2 Recovery Program Estimate," Rev.1, General Public Utilities Corp., July 1981.

603. NUREG-0985, "U.S. NRC Human Factors Program Plan," U.S. Nuclear Regulatory Commission, August 1983.

970. NUREG/CR-3568, "A Handbook for Value-impact Assessment," U.S. Nuclear Regulatory Commission, December 1983.

f I

1188. SECY-83-221, "Prioritization of Generic Safety issues," June 7,1983.

1189. Memorandum for W. Dircks from S. Chilk, "SECY-83-221 - Prioritization of Generic Safety issues," December 9,1983.

{

1190. Federal Register Notice 43 FR 1565, " Program for Resolution of Generic issues Related to Nuclear Power Plants," January 10,1978, 1191. Federal Register Notice 54 FR 24432, " Program for Resolution of Generic Issues Related to Nuclear Power Plants; Policy Statement," June 7,1989.

1192. RES Office Letter No.1, Revision 4, " Procedure for Identification, Prioritization, Resolution and Tracking of Generic issues," June 2,1994.

1193. RES Office Letter No. 2, " Procedures for Obtaining Regulatory Impact Analysis Review and Support," November 18,1988.

1194. RES Office Letter No. 3. Revision 2, " Procedure and Guidance for the Resolution of Generic issues," March 27,1989.

4 12/31/97 26 NUREG-0933

h ..

Revision 6 t

s O H 1195. NUREG-1251, " implications'of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States," U.S. Nuclear Regulatory Commission, (Vol.1) April 1989, (Vol. 2) April 1989.

. $ 1 A73. Memorandum for J. Taylor from S. Chilk, "SECY-89-102 - Implementation of the Safety  !

Goals," June .15,1990.

l 1479. SECY-93108, " Revised Guidelines for Prioritization of Geaeric Safety issues," April 28, 1993.

11493. Memorandum for V. Stello from S. Chilk, "Sta'ff Requirements - Briefing on Status of 1 Unresolved, Safety / Generic issues,10:00 a.m. Wednesday, October 21, 1987, Commissioner Conference Room D.C. Office (Open to Public Attendance," November 6,-1987..

1505. Memorandum for.J. Taylor from S. Chilk, "SECY-93-108 - Revised Guidelines for PriorItization of Generic Safety issues," July 23,1993.

c1563.- NUREG/CR-5382, " Screening of . Generic Safety issues for License Renewal Considerations," U.S. Nuclear. Regulatory Commission, December 1991.

156' 4. Memorandum for.W. Russell from E. Beckjord, " License Renewal implications of Generic Safety issues (GSis) Prioritized and/or Resolved Between October 1990 and March

^ 1994," May 5i1994.

i

'1689. Memorandum to J. Taylor from J. Hoyle, "COMSECY-95-033 - Proposed Doller per Person-Rem Conversion Factor: Response to SRM Concerning; issuance of Regulatory

- Analysis Guidelines of the U.S. Nuclear Regulatory Commission and SRM Concerning the

Need for. a Backfit Rule for Materials Licensees (RES-950225) (WITS-9100294),"

September- 18i 1995.

1690. Memorandum for All RES Employees from D. Morrison, "RES Office Letter No. 3C-Procedures for Obtaining Regulatory impact Analysis Review and Support," February 23, 1996.

4 s

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Y-SECTION 1 TMI ACTION PLAN ITEMS This'section contains the TMI Action Plan items that were documented in NUREG-0660." All items in Chapters I,11, lil,- and IV that were identified for prioritization and listed in this section follow the numbering system established in NilpEG-0660." Items found to be closely related have been combined where possible to form single issues for prioritization purposes. As a result, some of these combined issues contain items with the lead responsibility assigned to 1 several offices. However, the lead responsibility and a summary of the findings for each item

-lis,ted can be found in Table ll of the Introduction items clarified in NUREG-0737" are' listed -i in this section for accounting purposes only.

Chapters 1, Ililli, and IV presented a detailing of plans for NRC staff or licensee action whereas Chapter V addressed NRC policy, organization, and management and originally' called for 17

. specific actions to be taken by the Commissioners, in recognition of the interrelationships that

- required correlated planning, these 17 ' items were later grouped into 7 subject areas by the staff and for. warded to the Commission in'SECY.-80-2308."8 This revision to Chapter V was agreed

} "upon by the Commission and was published as Rev.1 to NUREG-0660"in July 1980. All items

/ . .of Chapter V listed in this section follow the numbering system established in NUREG-0660," -

Rev.1.

,s .!

H

+

e .,

l

( );

y ,

. .L12I31/97?<' -

1.0 1 . NUREG-0933 -

ws <- ; ;

y. ,

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w A +

y:

e -

, - Revisicn 3 -

+

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.y TASK LA: OPERATING PERSONNEL 2

LL k ' TALK 'I A 1' OPERATING PERSONNFt AhlO STAFFING-

! Complex transients iti ndclear power ' plants place high ' demands on the operators in the control

' room. The objective of the actions described in this task'was to increase the capability of the i

, . shift crews in the' control room to operate the facility in a safe. and competent manner, by '

assurin0 that a proper number of individuals with the proper qualifications and fitness are on

- shift at all times. The work to improve the design of control rooms is described elsewhere in -

this plan.

ITEM l.A.1.1: SHlFT TECHNICAL ADV1SOR

$Thistitem was clarified in NUREG-0737," requirements were issued, and MPA F-01 was established by DUNRR for implementation purposes.

LITEM l_'A:1.2: SHIFT SUPERVISOR ADMINISTRATIVE DUTIES This item was clarified in NUREG-0737" and requirements were issued.

jO '

ITFM l.A.1.3r SHIFT MANNING This item was clarified in' NUREG-0737," requirements were issued, and MPA F-02 was

. established by' DUNRR for implementation purposes, k

ITEM I.A.1.4 LONG-TERM UPGRADING DESCRIPTION f

The purpose of this item was to develop changes to 10 CFR 50.54 concerning shift staffing with~ licensed operators and working hours of licensed operators. As described in NUREG-0660," the staff was expected to " develop proposed changes to 10 CFR 50.for consideration by the Commission to effect appropriate changes concerning plant staffing,' including shift

manning,' control room presence, and working hours."

c CONCLUSIOh!

_ SIECY-81e440*!was prepaied by the st'aff in July 1981. and resulted in a Commission policy p fl ~ Tstatement on working houilimitations which was issued in the Federal Register on February 17.

j $1982. Working hour limitations were to be incorporated into Regulatory Guide'1.33 8 [see jW c issue.75). The. specific issues to be considered were: (1) the number of licensed operators, f( )? 1 based on the number of reactors', control room configuration, and operating model (2) whemir

.ge ~ .;

g w. -

m3 . ,.

1-12/31/97/

1.1.A.1-1. - NUREG-0933

t. 1

g. ,

i

a {y a s 1

f Revision 3 f

I f existing rulemaking should be expanded to include non-licensed operators; (3) whether existing rulemaking should be expanded to include " position titles," in addition to the type of NRC license: (4) whether shift technical advisors (STAS) or shift engineers (SEE .:.ould be required on shift; and (5) whether shift supervisors (SSs) should be licensed.

A proposed rule was published on August 30,1982 and, after the comment period expired, the 5

final rule was submitted to the Commissioners in SECY-83-52A " on March 14,1983. In response to the TMl Action Plan, licensing has required, through technical specification (TS),

the great majority of the substantive features of the expected changes to regulation. Therefore, adoption of the rule was expected to have the effect of codifying existing requirements with minimal impact on licensees. The final rule amending 10 CFR 50.54 was approved

• by the Commission on April 28,1983. Thus, this issue was RESOLVED and new requirements were established."

l REFERENCES

48. NUREG-0660, "NRC Action Plan Developed As a Result of the TMI-2 Accident," U.S.

Nuclear Regulatory Commission, May 1980, (Rev.1) August 1980.

98. NUREG-0737, " Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory Commission, November 1980, (Supplement 1) January 1983.

225. Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)," U.S.

Nuclear Regulatory Commission, November 1972, (Rev.1) February 1977, (Rev. 2)

February 1978.

250. SECY-81-440, " Nuclear Power Plant Staff Working Hours," July 22,1981.

595. SECY-83 52A, " Final Rulemaking Concerning Licensed Operator Staffing at Nuclear Power Units and Draft Policy Statement on Shift Crew Qualifications," March 14,1983.

596. Memorandum for W. Dircks from S. Chilk, " Staff Requirements - Affirmation / Discussion and Voto, 3:35 p.m., Thursday, April 21,1983, Commissioners' Conference Room (Open to Public Attendance)," April 28,1983.

956. Memorandum for V. Stello from H. Denton, "Close-out of the Division of Human Factors Technology TMI Action Plan items," January 6,1987.

l O'

12/31/97 1.1.A.1-2 NUREG-0933

R:visi:n 6

.n/

~ TASK 1.A.2 TRAINING AND QUALIFICATIONS OF OPERATING PERSONNEL 5

The objectives of this task were to: (1) improve the capability of operators and supervisors to understand and control complex reactor transients and accidents; (2) improve the general capability of an operations organization to respond rapidly and effectively to upset conditions; and (3) increase the education, experience, and training requirements for operators, senior '

operators, supervisors, and other personnel in the operations organization to substantially improve their capability to perform their duties. )

l 1

ITEM I.A.2.1 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING j AND QUAllFICATIONS This item required all operating plant licensees and all license applicants to provide specific improvements in training and qualifications of senior operators and control room operators. The  !

three parts of this item are listed below. j ITEM l.A 2.111h OUALIFICATIONS - EXPERIENCE DESCRIPTION t

( This NUREG-0660" item set specific experience requirements that were to be met by applicants for senior. operator licenses by May 1,1980. Applicants for senior operator licenses were required to have been a licensed operator for one year effective December 1,1980.

CONCLUSION This item was clarified in NUREG-0737,88 new requirements were issued, and MPA F-03 was established by DL/NRR for implementation purposes.

ITEM l. A.2.1(2h TRAINiliG DESCRIPT1Qht .

This'NUREG-0660" item set the following specific requirements:

(1) Effective August 1,1980, senior operator applicants were required to have 3  !

. months of continuous on-the-job training as an extra person on shift.

~(2) Effective August 1,1980, control room operator applicants were required to i have 3' months training on shift as an extra person in the control room. I r y ;(3) . Training, programs were to be modified to provide: (a) training in heat transfer, l W ~

fluid flow, and thermodynamics; (b) training in the use of installed plant systems

% e ).

L^ '

12/31/97. 1.i.A.2-1 -NUREG-0933

l R3 vision 6 i i

to control or mitigate an accident in which the core is severely damaged; and (c) increased emphasis on reactor and plant transients.  !

]

CONCLUSION This item was clarified in NUREG-0737," new requirements were issued, and MPA F-03 was established by DL/NRR for implementation purposes.

ITEM l.A.2.1(3h FACILITY CERTIFICATION OF COMPETENCE AND FITNESS OF A OLICANTS 1 FOR OPERATOR AND SENIOR OPERATOR LICENSES DESCRIPTION This NUREG-0660 item required all applicants for operator and senior operator licenses, pursuant to 10 CFR 55.10(a)(6),10 CFR 55.33(a)(4), and 10 CFR 55.33(a)(5), to be certified by the liighest level of the corporate management of their respective plants. This requirement was effective May 1,1980.

CONCLUSION This item was clarified in NUREG-0737," new requirements were issued, and MPA F-03 was established by DL/NRR for implementation purposes.

ITEM l.A.2.2: TRAINING AND QUAllFICATIONS OF OPERATIONS PERSONNEL DESCRIPTION Under the TMI Action Plan,** the NRC could require reactor licensees to review their training and qualification programs for all operations personnel. This was interpreted to include licensed and auxiliary operators, technicians, maintenance personnel, and supervisors. The review was to be conducted to examine existing practices in light of the safety significance of the duties of the operations staff. If the review determined that the existing practices adequately assured proper safety-related staff conduct, then documentation of the justification for this determination was required; this documentation did not require submittal to the NRC but was required to be maintained on site. If the review uncovered inadequacies, the licensee was required to upgrade the training and qualification practices to ensure adequate performance of l operations personnel. The evaluation of this issue included the consideration of item I.A.2.6(3).

PRIORITY DETERMINATION To estimate the effect of training reviews on operator-error contributions to plant risk, a panel of PNL experts was assembled with considerable experience in reactor operations, utility training programs, and reactor plant systems. The panel included members with utility field l experience and reactor operator licensing examiners. The judgments of the panel, as detailed l below, were based on the two following considerations:

O 12/31/97 1.1.A.2-2 NUREG-0933

Rsvision 6 ,

L .<

Q ~(1)< ;The potential'ehfect of this issue was limited by its semi-voluntary nature, i.e.,

Q ?thei judgment = of adequacy. was in : the' hands of the individual utilities.

Furthermore, the existing institute .for Nuclear Power Operations (INPO) and NRC J

- research work . In task analysis. dealt; with_ generic routine operations.

~ Plant-specific operation and operation under upset conditions were left to the

. Individual utilities. This diluted the effectiveness of the task analysis efforts in providing the basis for the training and qualifications review.

1 Related issues which were supported by and, in turn, supported this issue were I

the conduct of plant drills anti accreditation of training programs. While neither -

E of these were directly required by the training and qualifications review, both could have been a part of the response.and both would have had a positive

.effect on personnel performance.-

(2) There was a wide variation among utilities in both the training programs and the -

performance of operations staff. In many facilities, there was much room for improvement. Therefore, while the potential effect. of the training . and qualifications review effort was limited, a significant- overall reduction in safety-related human error for operations personnel was expected because of the wide margin available for improvement.

Assumptions The PNL panel divided licensees into three groups:

(1) Minimally-Affected: These utilities had a good, effective training and qualifbation ..

program and good operations personnel performance. They were to be minimally affected by this issue.' The fractional population of this group was estimated to be 15% of the reactor licensees.  !

-(2) Intermediatalv-Affectedi These utilities' training and qualification programs and/or operations performance had room for improvement. This group, estimated

, to be 60% of the population, had to undergo improvements and, therefore, were affected by the issue.

(3) Maximally-Affected: These utilities had deficiencies in their training and qualification programs and in operations personnel performance. They would be significantly affected by this issue and major restructuring of programs were

. expected. This group was estimated to contain 25% of reactor licensees.

~

. From the; estimates for _thesel groups, weighted composite estimates were derived.

NUREG/CR-2800" shows the safety benefit estimated by the panel for each of the groups and

,r

'also gives the weighted averages.

1The values given in NUREG/CR 2800" are in terms of percent changes. Foi inclusion into the U value/ impact score formula, they.were converted to other measures. The reduction in human error' was; transformed into1the resulting.' reduction'in risk, as measured by change in  !

probabilistic risk exposure (man-rem /RY) The change in annual ORE was also transformed from  :

1. n)

L./;

percent improvement into man-rem /RY.

L 12/31/9L NUREG-Og33

. 1.1.A.2-3 an__ _.

Rnvision 6 The reduction in risk was developed by examining the quantitative impact on accident event frequencies of human error rates in key scenarios. The reduction in human error translated into a reduction in accident frequency. No additional reduction due to accident mitigation was assumed. The values given in NUREG/CR-2800" were used for the best estimate of im-provement: 17% for operator error and 28% for maintenance.

Freauency Estimate This issue centered around operator and maintenance training programs to improve personnel performance. The issue related generically to both BWRs and PWRs and, ideally, the risk reduction attributable to its resolution was estimated by selecting a representative plant of each type. However, maintenance and operator performance essentially impacted accident sequences in the risk equations. The calculations were performed for one representative PWR and inferences drawn for all reactors. The Oconee-3 (a RSSMAP PWR) plant risk equations developed in NUREG/CR-1659," Vol. 4, were used for this analysis.

h .as assumed that the 17% reduction in operator error could be applied directly to elements containing an operator error frequency and the 28% reduction could be applied directly to maintenance variables. This assumption introduced some error in the maintenance contribution because some maintenance operations on nuclear systems have fixed times associated with cooldown and preparation, etc., in addition to the actual hands-on time for maintenance that would be subject to improvement through training. Maintenance done properly the first time also reduces the frequency of maintenance outage and downtime for proper repairs at some future date. Thus, fixed time periods in maintenance outages are indirectly reduced over the long run with improved maintenance performance simply because the need for maintenance may be reduced, except for systems that undergo preventive maintenance at set intervals.

Consecuence Estimate it was assumed that the resolution applied to all plants existing and planned, as given in NUREG/CR-2800, Appendix C." This represented a total of 4,000 RY of operation (143 plants with an average remaining life of 28 years). Implementation of the solution would provide a reduction of 9 man-rem /RY. For all plants, assuming a typical midwest-type meteorology and an average population density of 340 people per square-mile at U.S. reactor sites, the total public risk reduction was estimated to be 122,400 man rem.

Cost Estimate Industrv Cost: In estimating the ::osts to industry of implementing and operating under the resoiution of this issue, the PNL panel divided the industry once again into three categories; these groups and their estimates are shown in NUREG/CR 2800." The cost for implementation was the product of the number of plants and the cost / plant: (143)($0.335M) or $48M. The operation cost was the product of the number of plants, the average remaining life, and the annual plant cost: (143)(28)($0.16M) or $640M. Thus, the total industry cost was $(640 +

48)M or $688M.

NRC Cost: The NRC cost to implement the resolution was taken from NUREG-0660." This called for 1.1 man-years of NRC effort which was equivalent to $110,000. The annual NRC offort through ole to review the justification documentation and new training programs was e

12/31/97 1.1.A.2-4 NUREG-0933

R: vision 6 A

([ ' estimated.to' be one man-year ort $100,000/ year. Over the lifetime of the completed and planned reactors, this cost was estimated to be $2.8M. Therefore, the total NRC cost was

$[0.11 + 2.8]M or $2.9M.

Total Cost:.The totalindustry and NRC cost associated with the possible solution was $1688

- + 2.9]M or approximately $691M.

1 Valuellmonet A==ansment Based on an estimated public risk reduction of 122,400 man-rem and a cost of $691M for a possible solution, the value/ impact score was given by:

S~= 122.400 man-rem

$691M

= 177 man-rem /$M -I l

Other ConsideratiDES It was estimated that, with improved training, the operational doses could be reduced by 2.4  ;

x 10' man-rem for 143 plants over the average remaining plant life. Including the occupational i dose reduction (2.4 x 10' man-rem)in the above equation would increase the value/ impact I score to 524 man-rem /$M. PNL calculated *' the occupational risk reduction for accident-related i g  !

t ORE to be 880 man-rem.

CONCLUSION Because of the extensive number of sequences considered to be affected by this issue, the base case risk was high with a calculated range from 60 to 73 man-rem /RY. Bascd on the potential reduction in public risk and ORE, the issue was given a high priority ranking (see Appendix C).

However, in June 1985, the Commission recognized that the industry had made progress in

. developing programs to improve nuclear utility training and personnel qualification. As a result, the Commission adopted a Policy Statement on Training and Qualifications which made the training accreditation program managed by INPO the focus of training improvement in the industry.* Thus,'this item was RESOLVED and no new requirements were established.

ITEM l.A.2.3: ADMINISTRATION OF TRAINING PROGRAMS DESCRIPT1QR

' This NUREG-0660 item required the staff to develop criteria and procedures to be used in auditing training programs, including those provided by reactor vendors, and to increase the amount of auditing. Specifically, NRR was expected to: (1) audit training programs to ensure training was formalized and, eventually, in conformance with accreditation; (2) conduct cold operator licensing certification at simulators; and (3) pending accreditation, require certain

,n . _ instructors to be'SRO-certified.

'( )

' Lj.

12/31/97/' 1.1.A.2-5 NUREG-0933

Revision 6 CONCLUSION Elements (2) and (3) were implemented and were incorporated into the Examiner Standards and inspection Procedures. The issue of training audits was addressed by the Commission's Policy Statement'" on Training and Qualification of Nuclear Power Plant Personnel which endorsed the INPO-managed accreditation program." Thus, this item was clarified in NUREG-0737" and new requirements were issued.

ITEM l.A.2.4: NRR PARTICIPATION IN INSPECTOR TRAINING DESCRIPTION Based on NUREG-0660," the NRR licensing and human factors staff was required to provide supplementalinstruction to the ole inspectors as an addition to the previously established OIE l inspector training program. The purpose of such instruction was to focus the inspectors' j attention on problems associated with human factors. With such training, it was expected that the inspectors would become more sensitive to such problems and, hence, more apt to initiate corrective action and thereby improve plant safety in this area. This would provide a means of responding to the TMI-related concern on human factors prob!em. for plant operations staff.

The principal benefit to be derived from NRR participation in OIE inspector training was the improvements the inspectors would gain from enhanced training. This training would increase inspector awareness in human factors and personnel-related problems. In areas such as emergency procedures reviews, routine operational practices and hardware-to-human interface deficiencies could be found by inspectors and corrected. The potential significance of this issue was explored by a panel of PNL experts that included three reactor operator license examiners with utility field experience in training as well as general reactor safety.

The panel envisioned that the solution to this issue would be the addition of one week of instruction in human factors to the ole inspector training course. The staff from NRR would participate in the instruction but would probably rely on a qualified consultant to conduct the majority of the instruction. It was assumed that the principal target of the training would be the resident inspectors. The potential effect of the training upon the OIE review of emergency procedures, plant hardware, and routine practices could be significant, but the overall effect was thought to be limited because of two factors: the short exposure of the inspector to human factors training, and the indirect nature of the safety benefit. That is, a marginal improvement in inspector awareness could result in some corrective actions which would result in some safety improvement. The separation between initial action and the safecy benefit complicated assessment of the effectiveness of the proposed resolution of the issue.

L PNL estimated" a human-error rcte scduction of 2% for operators and maintenance personnel (operations staff assumed most likely to affect plant safety). This was an overall industry-wide estimate: some isolated actions could be highly significant. PNL estimated the cost for this additional training to % about $1,000.

Capabilities of insp<setors could clearly be improved through the proposed training. There would be an indirect effect on risk, since better trained inspectors would identify more cost-effective improvements in plant operations. However, there was no reasonable way that the magnitude 12/31/97 1.1.A.2-6 NUREG-0933

, Rsvision 6

, 1 ,

g - of the safety' significance and cost of the improvements could be estimated quantitatively. This .

additional training would enhance the capabilities and thus contribute to the effectiveness and -

efficiency 'of the NRC in performing its regulatory safety mission. Thus, this training proposal -

was determined to be a Ucensing lasue.

CONCLUSION

. This Ucensirig issue was resolved in September 1983 with the regionalization of the operator licensing function which provided for training and guidance nf the regional operator licensing personnel."' -

l

' ITEM l.A.2 5 PLANT DRH I R

,m s-DESCRIPTION .

' The intent of this TMl Action Plan" item was to upgrade operator training by requiring. 'I F , L operating personnel to. conduct plant drills during shifts. Normal and off-normal operating maneuvers would be simulated for walk-through drills on a plant-wide basis. Drills would also be required to test the adequacy of reactor and plant operating procedures. This was an effort to reduce the risk of off-normal operating conditions by improving the capability of operators I and supervisors to understand and control complex reactor transients and accidents, and also i to improve the general capability of an operations organization to respond-rapidly and  ;

effectively to upset conditions, j PRIORITY DETERMINATION Assumptions it 'was assumed that the frequency of a core-melt incident was 5 x 10 4/RY, based on ,

WASH-1400. Also, it was_ assumed that operator error accounted for 50% of these events, i but plant drills would improve operator performance by 2%. In addition, it was assumed that e the release associated with a core-melt was the value averaged over the probabilities of the  !

. WASH-1400 accident categories for PWRs and BWRs and weighted by the number of PWRs  !

(95) and BWRs (48). This resulted in a total of 2.4 x 10' man-rem / accident. The average i remaining plant life was assumed to be 28 years. I Frenuancv Entimata ,

i Based on the assumptions above, the reduction in the core-melt frequency resulting from pisnt l 4 4 1 drills was calculated to be (0.02)(0.50)(5 x 10 )/RY.or 5 x 10 /RY.

J Consanuanca Estimate Forf143 'affected plants with'an average remaining life of 28 years, the public risk reduction Lwasl estimated .to be- (5 x 104 /RY)(2.4 x'10' ' man-rem)(28 years)(143 reactors)'or '4,805 e '~  : man-rem. -# ,.

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Pevision 6 Cost Estimate Industrv Cost: The industry resources required for implementation were estimated to be one man-month / plant. This was the estimated personnel requirement associated with the utility staff time for attendance at the drill, preparation by staff and management, and staff time dedicated to the dissemination of insights gained from the drills. At a cost of $100,000/ man-year and with 4.33 weeks / month, this yielded a cost of $8,333/ plant. For the 143 affected plants, the cost was estimated to be $1.2M.

The industry resources required annually to participate in the plant drills were estimated to be 2 man-months / plant and included drill attendance, preparation before the drill, and dissemination of information afterward; this cost was $16,660/RY. For the 143 affected plants, this cost was $2.38M/ year. Over the average remaining life of 28 years, this cost was estimated to be $67M.

Thus, the total industry implementation and operational cost was $(1.2 + 67)M or approximately $68.2M.

NRC Cost: The total NRC cost to implement the resolution of this issue included NRC staff labor and services of a contractor. Since the activities of the NRC staff and the contractor were to some degree interchangeable, no attempt was made to provide separate estimates so that the totalimplementation cost was estimated to be $300,000. The annual cost to the NRC was also estimated to be $300,000; again,this was assumed to contain some mixture of staff and con-tractor expenses. Over the average remaining life of 28 years, the operational cost was estimated to be $8,4M. Therefore, the total NRC implementation and operation cost was $(8.4

+ 0.3)M or $8.7M.

Total Cost: The totalindustry and NRC cost associated with the possible solution was $(68.2

+ 8.7)M or $76.9M.

Value/imonet Assessment Based on an estimated public risk reduction of 4,805 man-rem and a cost of $76.9M for a possible solution, the value/ impact score was given by:

S = 4.805 man-rem

$76.9M

= 62 man-rem /$M 3

CONCLUSION Based on the above value/ impact score, the ranking of this issue would have been low to medium. Because the risk may have been estimated to be well on the conservative side, the J issue was given a low priority ranking (see Appendix C). However, ongoing work by DHFS/NRR on the subject was completed in July 1985 and published' for information only as NUREGICR-4258." Thus, this item was RESOLVED and no new requirements were established.*

J 12/31/97 1.1.A.2-8 NUREG-0933 i

4 P ,

Revision 6 s

mmMNG AND QUALIFICATIONS' jlTEM l_A 2 Blih REVISE REGULATORY GUIDE 1.8

' Items 1.A.2.6(1), l.A.2.6(2), l.A.2.6(3), and I.A.2.6(5) were combined and evaluated together.

DESCRIPTION Historical Background item I.A.2.6 of-the~TMl Action Plan" called for the long-term upgrading of training and qualifications of operations personnel. The specific paragraphs of this item in NUREG-0660"

- called for a revision of Regulatory' Guide 1.8,88' (ANSI /ANS 3.1),8 in order to incorporate short-term requirements into this issue and any other changes resulting from a national

' standards effort. Also, it was stated that more. explicit guidance regarding exercises in simulator requalification programs would be included in the regulatory' guide (Recommendation 8 of

.SECY-79-330E*) as would qualifications of shift supervisors and senior reactor operators

![NUREG-0585,"' Recommendations 1.6(1) and (2)]. In addition, based on the NRC staff review

' of' NRR 80-117,8*8 recommendations were to be made to the Commission and Commission decisions factored into the regulatory guide or regulation changes. Moreover, appropriate revisions to '10 CFR 55,' Operator Licenses, were to be recommended for action by the Commission in order to incorporate the applicable short-term changes plus requirements based S

on Commission action on SECY-79-330E for mandatory simulator training for applicants for

licenses 1 Recommendation 4); mandatory simulator training in requalification programs

_ / (Recommendation 7); NRC administration of requalification examinations (Recommendation 9, as modified by the Commission); and mandatory operating tests at simulators (Recommendation 11).

Finally, the Nuclear Waste Policy Act of 1982 (Public Law 97-425, Section 306) authorized and directed NRC to promulgate regulations or guidance for the training and qualification of civilian nuclear power plant personnel. A task force was formed within NRC as a result of this bill. As part of the task force objectives, items 1.A.2.6(1,2, and 3) were to be addressed.

Safety Significance

A public risk reduction was anticipated as a result of a reduction in core-melt frequency which follows from a reduction in operator error rates. Reduction in operator errors was expected to

' result from the upgraded training and qualifications which formed the assumed resolution of

' this issue. -

Possible Solutions

The upgrades were assumed to include an increase in time spent in simulator operation, both Lin training and in requalification. The simulator time was assumed to. improve in ' quality as well as quantity. Emphasis on improvements on the operators diagnostic capability was felt to be lespecially important in contributing to a reduction in core-melt frequency. Furthermore, the p;  : enforcement activities in terms of NRC-administered examinations and ole inspection of training

(,Vj l programs were likely to emphasize the~value of this long-term training and qualification of reactor operators.'

11,2/31."M

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. 1.1.A.2-0 ~ NUREG-0933 -

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R: vision G PRIORITY DETERMINATION The numerical assessment of this issue was conducted by PNL staff ** with experience in reactor operator licensing, reactor operation, and general reactor safety, in consultation with General Physics Corporation. General Physics Corporation provided utility training services and had significant experience in reactor simulators, providing procurement and startup assistance, operation and maintenance services, and simulator modifications.

Assumotions it was assumed that the resolution of this issue would take the form of upgrading utility training and qualification programs that would represent a major enhancement of the training and qualification programs.

It was noted that many of the TMI Action Plan'* items associated with operator training were interrelated and it was, therefore, difficult to assess them independently. For example, this issue was related to l.A.4.1 which addressed the improvement of simulators and provided for more realistic modeling of a plant, whereas this issue, [l.A.2.6(1,2,3,5)], dealt with training

. Improvements, including the enhanced use of existing simulators. Either issue, by itself, would improve operator performance; however, there could have been significant overlaps in improving operator performance if both items were implemented. Even though it was recognized that the total improvement would be less than the sum of the individual contributions when each is assessed separately, the extent of any overlap was not identified here.

Based on engineering judgment,it was estimated by the PNL panel that the resolution of this issue would result in a 30% reduction in operator error rates. The number of plants to which this issue was applicable was assumed to be 95 PWRs and 49 BWRs with average remaining lives of 28.5 years and 27 years, respectively.

For the PNL analysis, Oconee-3 was selected as the representative PWR plant. It was assumed that the fractional risk and core-melt frequency reductions for the representative BWR (Grand Gulf-1) would be equivalent to those for the representative PWR. Therefore, the analysis was conducted only for a PWR, but the fractional risk and core-melt frequency reductions were also applicable to a BWR. The dose calculations were based on a reactor site population density of 340 people per square-mile and a typical midwest meteorology was assumed.

Frecuency Estimate Based on the a.. acted accident sequences and the parameters affected by the possible solution, the original core-melt frequencies of 8.2 x 10/RY for PWRs and 3.71 x 10~'/RY for BWRs were calculated to be reduced by about 16%.

Consecuence Estimate

'The associated reduction in public risk was 31 man-rem /RY for PWRs and 37.4 man rem /RY for BWRs, resulting in a total public risk reduction of 132,600 man-rem for all plants.

12/31/97 - 1.1.A.2-1 u NUREG-0933

T 4-L Revision 6 Coat Estimata industry Cost: The resolution of this issue was assumed to be a major enhancement of the' training and qualification programs. The programs would have to be upgraded in order to' meet the requirements of INPO accreditation. These requirements were assumed to be far-reaching and required significant effort on the part of utility training staffs. The amount of effort would

vary among the utilities, depending on.the existing state of their programs. The effort to implement the program 'was estimated by the PNL panel to require 10 to 20 man-years of effort :i at each plant. The.mean value was expected to be shifted toward the lower and since, at the j time of this evaluation, many utilities were improving their training programs. A 12 man-year L '!

~effort was taken as the mean estimate.

Operation under.the' upgraded programs would require enhanced training activities and more operator time in training; the training staff was estimated to require three additional people. It-t was assumed the major cost of additional operator time could be estimated from increased time 1st simulators. It was estimated that 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of simulator time would be added to operator ~ q training and requalification. For 20 operators / year passing through these' programs, this was l a . equivalent to 800 additional hours. It was further assumed that operators could be trained three

~a t a time on the simulator and that simulator time could be acquired for $600/ hour. T"'

additional simulator cost was estimated to be ' $160,000/ year. The industry costs were )

I estimated as follows:'

(1) Imolamentation (12 man-years / plant)(143 plants)($100,000/ man-year) = $173M i (2) Onaration and Maintenance 4.) .  ;

I Training Staff = (3 man-year /RY)(52 man-weeks / man-year)

= 156 man-weeks /RY Operators = (800 man-hour /RY)/(40 man-hours / man-week)

= 20 man-weeks /RY '

i Th'us, the total labor was 176 man-weeks /RY.  !

(b) sirnulatnr Time (Onarator 1 : l l

'(800' man-hours /RY)/(3 man-hours / simulator-hour) = 267 simulator-hour /RY Therefore, the industry cost / plant-year for operation and maintenance was given by:

. [176 man-wk1[' 4100.OOO/ man-vr 1 +. [.267 aimulatar-hr if $800 1 f,' RY- c52 man wk/ man-yr RY simulator-hr.

( '

'~

' = 500,000/RY -

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/-

112/31/97j .

, 1.112-11 'NUREG-0933

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I i

R: vision 6 For all affected plants, the total industry cost for operation and maintenance war. 1

($500,000/RY)[(49)(27) + (95)(28.5)]RY or $2,000M. ]

The totalindustry cost for implementation, operation, and maintenance of the solution was then

$(173 + 2,000)M or $2,173M.

NRC Cost: The NRC effort to implement the resolution of this issue would be significar.t. It was estimated in NUREG-0660"that 5.4 man-years plus $259,000 would be required. At the time of the evaluation of the issue, some of the development activities had been completed; however, much work remained to be done. The remaining effort was estimatr,d to be 4.5 man-years and $100,000.

The operational activities of the NRC would include reviews of training progiam, increased inspection, and additional examination. The annual labor for reviews and inspections was estimated to be equivalent to 3 man-years. The principal addition in examinations was assumed to be NRC cunduct of a portion of requalification examinations. It was assumed that the NRC would conduct 25% of the requalification examinations and that 20 operators would be requalified at each p! ant every year. It was estimated that one man-month was required for each plant based on the assumption that the five (25% of 20) operators se.ceted for NRC examination at each plant would be tested at the same time. NRC ccsts were estimated as follows:

(1) Imolementation Staff Labor + Other Costs

= (1.4 man-week / plant)($1,600/ man-week) + ($100,000/144 plants)

= $3,386/ plant Total cost for all affected plants was ($3,386/ plant)(144 plants) or $488,000.

(2) Review of Maintenance and Ooeration (a) Review and Inspection = (3 man-year /yr)(52 man-wk/ man-yr)/144 plants

= 1.08 man-wk/RY (b) Examination = (1 man-rnonth/RY)(3.7 man-wk/ man-month)

= 3.7 man-wk/RY Thus, the total time spent was 4.78 man-wk/RY.

The cost / plant-year for the review of operation and maintenance was (4.78 rnan-week /RY)($1,900/ man-week) or $9,088/RY. For the 144 affected plants, this cost was (49,088)[(49)(27) + (95)(28.5)) or $36.6M.

Thus, the total NRC cost for implementation, operation, and maintenance was $(0.488 +

36.6)M or $37.1M.

Tottd cost: The total industry and NRC cost associated with the possible solution was estimated to be $(2,173 + 37.1)M or $2,210M.

12/31/97 1.1.A.2-12 NUREG-0933

-fm S h-P -

Rsvision 6.

n I. s,

'i.L , . f' g Valuallmpact Annanament M

• • hk Based on an estimated public risk reduction of 132,600 man-rem and a cost of $2.210M for.

~

- c a possible solution, the value/ impact. score was given by:

v S = 122 800 man-ram 42,210M

= 60 man rem /$M Other Considerations t The total occupational' risk reduction was associated only with accident avoidance inasmuch as there was no dose associated with implementation or maintenance of the solution..With a dose of 20,000 man-rom associated with accident cleanup and with the calculated reductions 4

E ;in core-melt frequencies of 1.3 x 104 /RY and 5.9 x 10 /RY for PWRs and BWRs, respectively, I

'the total occupational dose reduction associated with accident avoidance was calculated to be

' 860 man-rem.

p CONCLUSION Nthough the Oslue/ impact score was iow, this issue was given a high priority ranking' because of the large potential public risk reduction (see Appendix C). Resolution of the issue included the consideration of items 1.B.1.1(6,7) regarding changes to Regulatory Guide 1.8.88'i in November 1986, SECY-86-348** was submitted to the Commission with recommended f ' revisions to' Regulatory Guide 1.B* to endorse ANSI /ANS 3.1-1981 for the positions of shift

- supervisor,' senior operator, licensed operator, shift technical advisor, and radiation protection 888 manager. These revisions to Regulatory Guide 1.8 were subsequently approved by the 1

' Commission. and published in. May 1987.** Thus, this issue was RESOLVED and new l requirements were established.*' l

= lTEM f A 2.812h STAFF REVIEW OF NRR 80-117 This Litem was evaluated in item I.A.2.6(1). above and, in accordance with , an RES memorandum,'" was RESOLVED. No new requirements were established.

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(ITEM l.A.2[813b REVISE 10 CFR 55 'l gi

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,m This item was evaluated in item I.A.2.6(1) above and, as a result of the Nuclear Waste Policy .

' Act of 1982 (Public' Law 97-425), was determined to be covered in item I.A.2.2.

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Revision 6 ITEM 1.A.2.6(4h OPERATOR WORKSHOPS DESCRIPTION Historical Backaround On the basis of NUREG-0660,*8 NRR was required to develop a Commission paper on training workshops for licensed personnel. NUREG-0585,"* the source of this issue, states that the Intent of the issue was to conduct seminar-type workshops to exchange information on operations experience between the NRC and licensees and among licensees. This would assist in the improvement of operator performance and in improvements to reactor regulation, both resulting in ,mproved safety. The proposed requirements would have one representative for each shift at each unit attend such a workshop annually.

Safety Sionificance it was expected that there would be two potential pathways to improved safety benefit that would emerge from this issue: (1) improved operator performance through the sharing of safety-related experiences; and (2) the effect of improved regulation arising out of interaction between the operators and the NRC attending the workshops. The second pathway was considered to be a second-order effect and very difficult to quantify. Therefore, it was assumed that all the benefit would be derived through the reduction in operator-error rates.

PRIORITY DETERMINATION Assumntions It was assumed that major gains in reactor safety would come through the improvement in operator performance, i.e., a reduction in their error rates. There was also a pathway to improve safety by means other than human performance through improved regulations developed from operator input at the workshops. The latter would be extremely difficult to quantify so that only the human error rate-reduction pathway to improved safety was treated.

A panel of PNL experts was assembled and included staff that conduct operator licensing examinations, staff with experience in reactor operations, reactor safety and risk assessment, and the staff responsible for the conduct of the operator feedback workshops for NRR. This panel produced the estimates that formed the basis of this analysis. The analysis was based on the following additional assumptions:

(1) Acolicable Plants: 95 PWRs and 48 BWRs.

(2) Selected An3! vsis Plant: Oconee representative PWR. It was assumed that the fractional risk and core 2 melt frequency reductions for the representative BWR (Grand Gulf-1) _would be equivalent to those fer the representative PWR.

Therefore, the analysis was conducted only for a PWR, but the fractional risk and core-melt frequency reductions were also applied to a BWR, 0

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1 F . Revision 6 J3) Affected Accident Saguances and Bana Case Frequencias: Most sequences were.

'~a ffected. The affected sequences and the base case frequencies are shown in .

NUREGICR-2800."

(4)' Affected Releasa Categories and Base Case Frequencias: All release categories.

were affected by the resolution.' The original base case frequencies were used ' j

, . as given below. <

Oconaa-3 Grand Gulf-1 '

IPWR 1 = 1.10 x 104/RY .BWR-1 = 1.09 x 104/RY PWR-2' = 1.00 x' 10 4/RY . BWR 2 = 3.35 x 10 4/RY-PWR-3 = 2.86 x 104/RY - BWR-3 = 1.44 x 104/RY Frannancy Entirnata .

. The PNL panele ' stimated" the most likely reduction in human error rates for operators due to

' the' conduct of the proposed workshops would be 3%, assuming that the workshops were

- conducted in the manner perceived, i.e., to focus on data-gathering for the NRC. This reduced '

the amount of time that could be devoted to inter-licenses sharing of operational experiences

. which would have had c more direct effect on safety-related operational performance in the plants. The possible reduction ranged from 1 % to 10%. If the focus could have been shifted -

toward the inter-licensee exchange of operational experiences, the most likely reduction in error rate would shift upward; however, it was not expected to exceed 10%.

Conseouance Estimata Based on. the PNL estimates and calculations" and assuming a typical midwest-type

' meteorology and an average population density of 340 people per square-mile at U.S. reactor

~

sites, the public risk reduction was estimated to be 7,140 man-ram for 143 plants with an average remaining life of 28 years. The occupational dose reduction was minor at a calculated value of 46 man-rem.

Cost Estimate industrv Cost: The' industry resources required for implementation were estimated to be one man-month / plant. This was the_ estimated personnel requirement associated with the trial

. workshops that were being conducted, it included utility staff time for attendance at the j' workshop, preparation b'y staff and management, and staff time dedicated to the dissemination F

of Linsights gained, at .the : workshop. At a cost of $100,000/ man year and with 4.33 J weeks / month; this yielded a cost'of $8,333/ plant. For the 143 affected plants, this cost was estimated to be $1.19M.

' (The industry resources l required: annually. to participate in the training workshops were _

. estimated to be the same as those for implementation, i.e.,-~one man-month / plant; which -

Eincludediworkshop attendance,' preparation before the- workshop, and dissemination of finformation. afterward.1This was equivalent to-48,333/RY. For 143 ' plants, this costLwas S life of the As); ; estimated to be 143 man-months / year or;41.19M/ year. Given the average'remainirig\

y . , -

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Rsvision 6 plants, the operational cost was $33.3M. Therefore, the total industry cost associated with the solution to this issue was $34.5M.

NRC Cost: The NRC cost to implement the resolution of this issue was estimated to be $0.3M and included NRC staff labor and services of a contractor. Since the activities of the NRC staff and the contractor were to some degree interchangeable, no attempt was made to provide separate estimates. The annual cost to the NRC was also estimated to be $0.3M; again, this was assumed to contain some mixture of staff and contractor expenses. Over the average remaining plant life, the operational cost was estimated to be 88.4M. While not specific, these estimates for implementation and operation were firmly based on the experience of conducting the trial workshops. Therefore, the NRC implementstion and operation cost was estimated to

. be $8.7M.

Total Cost: The totalindustry and NRC cost for the possible solution was estimated to be

$(34.5 + 8.7)M or $43.2M.

Value/imoact Assessment Based on the estimated public risk reduction of 7,140 man-rem and a cost of $43.2M for a possible sslution, the value/ impact score was given by:

S = 7.140 man-rem

$43.2M

= 165 man-rem /$M Other Considerations The accident avoidance cost was estimated by calculating the product of the change in accident frequency (AF) and the estimated cost to the utility of a major accident (A); the latter term was estimated" to be $1.65 Billion. Thus, the cost / plant-year was estimated to be:

PWRs: (AF)(A) = (7 x 10 4)($1,650M)/RY = $1,200/RY BWRs: (AF)(A) = (3.2 x 10 4)($1,650M)/RY = $530/RY The total cost for all plants was the product of the cost / plant-year, the number of plants (N),

and the average remaining life (T) of each type of plant:

I(NT)(AF)(A) = $(95)(28.5)(1,200)M + $(48)(27.0)(530)M = $3.9M CONCLUSION Because of the extensive number of sequences considered by PNL to be affected by this issue, the base case risk was high at a calculated range from 60 to 73 man-rem /RY. With a value/ impact score of 165 man-rem /$M and an estimated risk reduction of 7,140 man-rem, this issue was given a medium priority ranking (see Appendix C).

The staff conducted three workshops and a mail survey in order to evaluate the effectiveness of both rnechanisms for obtaining feedback to the NRC from utility operating staffs. The results '

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of thess'two approaches were-documented in NUREG/CR-3739"*.'and NUREG/CR-4139,"*

? respectively.;The staff concluded that both feedback mechanisms proved to be effective imethods of gathering data from operations personnel and did not recommend conducting workshops or surveys on an annual basis; it was preferable to use'such mechanisms judiciously

-when a real need existed."' Thus, this item was RESOLVED and no new requirements were

' established.

ITEM l.A.2.6(Sh DEVELOP INSPECTION PROCEDURES FOR TRAINING PROGRAM

. This : item was evaluated : in ? ltem I.A.2.6(1) above 'and, in accordance - ~with an ole

' memorandum," was RESOLVED. No new requirements were established.'

ITEM l.A.2.6(6h NUCLEAR POWER FUNDAMENTALS DESCRIPTION Historical Background This NUREG-0660*? item called for.NRR to develop requirements for the inclusion of nuclear power fundamentals in the instruction given to reactor operators. This arose out of a concern"*

that the.12 weeks of fundamentals training that were being. given to operators were insufficient.

\

Safety Significance Safety issues that deal with operator training can affect public risk by improvements in the operator safety-related performance. This can lead to a reduction in core-melt frequency and' a reduced probabilistic risk.

Possible Solution The' additional nuclear power fundamentals training would add 4 weeks to the training period.

PRIORITY DETERMINATION

' in order to assess this issue,'a panel of experts was assembled from the PNL staff. This panel

? was comprised of members experienced in rea: tor operator licensing, reactor operations, utility field work, and general reactor safety areas. The results of the PNL assessment are contained in NUREG/CR-2800."

The' PNL panel felt there had been significant progress.across the industry in the area of instruction in nuclear power. fundamentals since the issuance of NUREG0585"' in 1979. j

' Further increase ~ in emphasis on fundamentals was felt to be unlikely to improve operator.

' performance. The trend in operator licensing examinations was to stress operational knowledge

}fyy and ~de-emphasize fundamentals.1 This supported the view that further fundamental training Lwould not add to plant' safety.

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l R:: vision 6 Frecuenev/Consecuence Estimate f The PNL panel felt that the existing level of instruction in nuclear power fundamentals was adequate. Further emphasis on fundamentals was viewed as not likely to improve operator safety performance. Therefore, there would be no measurable public risk reduction associated with the possible solution to this issue. The PNL panel also saw no reduction in occupational dose associated with the implementation of the solution.

Cost Estimate Industrv Cost: It was assumed that 20 operators would complete the training course each year at every plant. In addition, one full-time instructor was assumed to be required. This yielded 80 man-weeks for the operators and 44 man weeks for the instructors, or 124 man weeks / plant overall each year. To implement this practice, an effort equivalent to one year of operation (124 man-weeks) was estimated to be required.

NRC Cost: Implementation of the solution was estimated to take 0.4 man-year or approximately 18 man-weeks: no added costs were estimated for operation. The review of the additional instruction could be contained in the existing routine function thereby causing no bdded expense.

Valuellmonet Assessment Based on the judgmeit that there would be no risk reduction resulting from this issue, the value/ impact score w is zero.

CONCLUSION in view of the fact that it was believed that the existing level of instruction in nuclear power fundamentals was adequate for reactor operators, further emphasis on fundamentals as r, squired by this issue was viewed as not likely to improve operator safety performance. Vlith the resulting value/ impact score of zero, this issue was DROPPED from further consideratbn.

l ITEM l.A.2.7: ACCREDITATION OF TRAINING INSTITUTIONS I

DESCRIPTION i Historical Backoround Based on the requirements of NUREG-0660, this item required NRR to complete a study to establish the procedures and requirements for NRC accreditation of reactor operator training l

' programs. The resulting study was to be developed into a Commission paper describing the  ;

various options for accreditation. l l

Safetv Sionificance There were two aspects to the safety benefit of this issue: (1) the reduction of public risk through the improvement of operator performance, which was expected from the improved ]

12/31/97 1.1.A.2-18 NUREG-0933 l 1

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Revision 6 i

(,)' training accreditation; and (2) a reduction in occupational exposure, primarily for operators who often supervise maintenance or perform other duties in radiation zones. However, some reduction in routine occupational exposure could also be expected for other operations personnel as a result of the increased awareness by the operators.

Possible Solution in order to assess this issue, a panel of experts was assembled from the PNL staff. This panel was comprised of members experienced in reactor operator licensing, reactor operations, utility field work, and general reactor safety areas. The panel envisioned the resolution of this issue as the formation of an accreditation board consisting of representatives from the NRC, industry, and academia. This board would develop and apply criteria for accreditation including training programs of utilities, university-related programs, and independent training institutions.

While theoretically applying to training for all operations staff, the PNL panel felt the existing thrust was focused on reactor operators. Therefore, this assessment was made assuming only operators would be affected."

PRIORITY DETERMINATION Assumotions The views of the PNL panelincluded an awareness of the fact that, at the time this issue was o evaluated, some training programs were very near to accreditation. Either through association V)

( with the universities or through other means of providing high quality instruction, these programs would be likely to acquire accreditation from the board easily. Other training programs were not so well prepared for accreditation and may have required significant effort and expense to upgrade them. Some savings may have been gained for multi-unit sites by sharing costs.

Therefore, the resolution of this safety issue was assumed to be an improvement in operator performance. For some utilities (approximately 10% of the total), this issue essentially had no I effect because: (1) their existing training programs would be accredited with little effort; and (2) the quality of their programs was sufficiently high that accreditation would result in no discernible improvement in their operators' performance. Other utilities were expected to see varying degrees of improvement. Those with training programs that were below the accreditation standards were to be brought closer to the high quality enjoyed by the outstanding utilities. Overall, the effect on operator human error was estimated to be a reduction of 10% across the affected portion of the industry. The detailed assumptions for this analysis were as follows:

(1) Anolicable Plants: 90% of all plants - 43 BWRs and 86 PWRs, or 129 plants.

(2)- Selected Analvsis Plant: Oconee 3 - representative PWR. It was assumed that the fractional risk and core-melt frequency reductions for the representative BWR (Grand I Gulf 1) would be equivalent to those for the representative PWR. Therefore, the analysis

]

was conducted only for a PWR, but the fractional risk and core-melt frequency 1

(' reductions were also applied to a BWR.

= Q}

.12/31/97 1.1.A.2-19 NUREG-0933

Revision 6 Frecuenev/Consecuence Estimate Based on the PNL analysis ** and assuming a typical midwest type meteorology and an average population density of 340 people per square-mile at U.S. reactor sites, the estimated public risk reduction was 26,180 man-rem.

Cost Estimain Industrv Cost: The PNL panel estimated" the one-time industry cost to implement the change initially to be in the range of $0.1M to $1M/ reactor. Those plants with training programs closer to accreditable status would enjoy the smaller costs. The best estimate for the average plant was taken to be $0.3M. Operation under the accreditation program was estimated to cost between $0.05M and $0.25M/ plant annually for additional funding to maintain an accredited training program; the best estimate was $0.1M/ plant annually. The following is a breakdown of the industry cost:

(1) . Imolementation: Approximately 3 man-years ($300,000/ plant) to: (1) review

' accreditation standards; (2) compare the existing utility practices with the developed standards; and (3) plan the necessary upgrades and implement the program upgrades to fulfill the accreditation requirements. For 129 affected plants, this cost was estimated to be $39M.

(2) Ooeration and Maintenance: $100,000/ plant-year for: (1) the time invested by the staff in upgraded training (increased course time, quality, etc.); and (2) instruction upgrade (time, quality, etc.). For 129 affected plants with an average remaining life of 28 years, this cost was estimated to be $360M.

Thus, the total industry implementation, operation, and maintenance cost for the possible solution was estimated to be $399M.

NRC Cost: The NRC cost to implement the accreditation was estimated to be $0.635M which was equivalent to 330 man-weeks to: (1) accredit, predicated on the possibility that INPO accreditation would not be forthcoming; and (2) develop accreditation standards and regulations for adoption by the industry. The annual operational cost to the NRC was estimated *' to be

$100,000 or one man year for additional OIE efforts to ensure industry maintenance of standards (at all plants). For an average remaining plant life of 28 years', this operation and maintenance cost was estimated to be $2.8M. Thus, the total NRC cost for a solution to this issue was $3.435M.

Total Cost: The totalindustry and NRC cost associated with the possible solution was $(399

+ 3,435)M or $402.4M.

Valuellmonet Assessment Based on an estimated public risk reduction of 26,180 man-rem and a cost of $402.4M for a possible solution, the value/ impact score was given by:

S = 26.180 man-rem

$402.4M 12/31/97- 1.1.A.2-20 NUREG-0933

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= 65 man-rem /4M Other Considerations

The industry accident' avoidance cost was estimated by PNL to be 414M. The occupational

risk reduction'was estimated to be 22,170 man-rem resulting from accident avoidance (170

? man-rem) and from operation and maintenance of the solution (22,000 man-rem).

CONCLUSION Although the value/ impact score was low, this issue was given a medium priority ranking (see Appendix C) because of the magnitude of the potential public risk reduction. However, in June 1985, the Commission recognized that the industry had made progress in developing programs to improve nuclear utility training and personnel qualification. As a result, the Commission adopted a Policy Statement on Training and Qualifications which made the training

, accreditaticn program managed by INPO the focus of training improvement in the industry.*

Thus, this item was RESOLVED and no new requirements were established.

REFERENCES l

.16. ' WASH-1400 (NUREG 75/014), " Reactor Safety Study, An Assessment of Accident Risks

. in U.S. Commercial Nuclear Power Plants," U.S. Atomic Energy Commission, October 1975.

' 48. NUREG-0660, "NRC Action Plan Developed as a Result of the TMl-2 Accident," U.S.

Nuclear Regulatory Commission, May 1980, (Rev.1) August 1980, i

54. NUREG/CR-1659, " Reactor Safety Study Methodology Applications Program," U.S. l

- Nuclear Regulatory Commission, (Vol.1) April 1981, (Vol. 2) May 1981, (Vol. 3) June  !

1982,- (Vol. 4) November 1981. l 1

64. NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety issue Prioritization Information Development,"' U.S. - Nuclear Regulatory Commission, February 1983, 1 (Supplement 1) May 1983, (Supplement 2) December 1983, (Supplement 3) September

- 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.

98. TNUREG-0737, " Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory y Commission, November 1980, (Supplement 1) January 1983. j 174.7 NUREG-0585, "TMl Lessons Learned Task Force Final Report," U.S. Nuclear Regulatory W

! Commission, October 1979.

226. Regulatory.Guids 1.8, " Qualification'and Training of Personnel for Nuclear Power .

Plants," U.S. Nuclear Regulatory Commission, March 1971, (Rev.1) September 1975,

' (Rev.1-R) May 1977, (Rev. 2) April 1987. l l

251. SECY-79-330E, " Qualifications of Reactor Operators," July 30,1979.

i

'rG .

l

" 252.1. NRR-80-117, " Study of Requirements for Operator Licensing," February 4,1980. j i

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.1.1.A.2-21 . NUREG-0933 -

Rsvision 6 1 253. ANSI /ANS 3.1, " Selection, Qualification, and Training of Personnel for Nuclear Power ]

j Plants," American National Standards Institute,1981.

379. Memorandum for H. Denton from R. DeYoung, " Draft Report on the Prioritization of Non-NRR TMI Action Plan items," January 24,1983.

437. Memorandum for H. Denton from R. Minogue " Draft Report on the Prioritization of Non-NRR TMI Action Plan items," March 29,1983.

438. Memorandum to G. Cunningham, et al., from W. Dircks, *NRC Actions Required by Enactment of the Nuclear Waste Policy Act of 1982," January 19,1983.

651. NUREG-0985, "U.S. Nuclear Ragulatory Commission Human Factors Program Plan," U.S.

Nuclear Regulatory Commission, August 1983, (Rev.1) September 1984, (Rev. 2) April 1986.

777. Memorandum for W. Dircks from H. Denton, "Closecut of TMI Action Plan items I.A.2.2 and I.A.2.7 Training and Qualifications of Operating Personnel," June 24,1985.

800. NUREG/CR-4258, "An Approach to Team Skills Training of Nuclear Power Plant Control Room Crews," U.S. Nuclear Regulatory Commission, July 1985.

801. Memorandum for W. Dircks from H. Denton, " Team Training for Nuclear Power Plant Control Room Crews," July 10,1985.

802. NUREG/CR 3739, "The Operator Feedback Workshop: A Technique for Obtaining Feedback from Operations Personnel," U.S. Nuclear Regulatory Commission, September 1984.

803. NUREG/CR-4139," The Mailed Survey: A Technique for Obtaining Feedback from Operations Personnel," U.S. Nuclear Regulatory Commission, May 1985.

804. Memorare' m for W. Dircks from H. Denton, "TMI Action Plan item I.A.2.6(4),"

September 25,1985.

956. Memorandum for V. Stello from H. Denton, "Close-out of the Division of Human Factors Technology TMl Action Plan items," January 6,1987.

966. Federal Register Notice 50 FR 11147, "10 CFR Ch.1, Commission Policy Statement on Training and Qualification of Nuclear Power Plant Personnel," March 20,1985.

1043. SECY-86 348, " Final Rulemaking for Revisions to Operator Licensing 10 CFR 55 and Conforming Amendments," November 21,1986.

1044. Federal Register Notice 52 FR 16007, " Regulatory Guides; issuance and Availability,"

May 1,1987.

1045. Memorandum for V. Stello from E. Beckjord, " Resolution of TMl Action Plan Items and Human Factors issues," May 18,1987.

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TASK l.A.3: LICENSING AND REQUALIFICATION OF OPERATING PERSONNEL The objectives of this task were to: (1) upgrade the requirements and procedures for nuclear power plants operator and supervisor licensing to assure that safe and competent operators and senior operators are in charge of the day-to-day operation of nuclear power plants; and (2) increase the requirements for initialissuance of licenses and for license renewals and provide closer NRC monitoring of licensed activities.

l ITEM l.A.3.1: REVISE SCOPE OF CRITERIA FOR LICENSING EXAMINATIONS I DESCRIPTION

This NUREG-0660'* item called for NRR to notify all operator license holders and applicants of a the new scope of examinations and criteria for issuance of reactor operator (RO) and senior reactor operator (SRO) licenses and renewal of licenses. Simulator examinations were to be included as part of the license examination. As a result of Public Law 97-425, it was determined that additional staff work on the issue was required and a proposed rule for operator licensing was presented to the Commission in SECY-84-76.'" Approval of this rule  !

was expected to effectively close out this item. '

'(w CONCLUSION This item was clarified in NUREG-0737 and new requirements were established.

ITEM LA.3.2: OPERATOR LICENSING PROGRAM CHANGES DESCRIPTION This NUREG-0660 item called for NRR to take the following actions:

(1) Develop and implement a plan to relocate Operator Licensing Branch (OLB) examiners at Nuclear Power Plant Simulator. Training Centers or in Inspection and Enforcement Regions.

- (2) Conduct a study of the staffing of the operator licensing program and the qualifications and training of examiners.

(3) Develop and implement a plan to report operator errors and to act on operator errors with respect to continuation of licensing.

- In response to the above actions, the following were accomplished:

f3 - (1) The administering of examinations ~and issuance / renewal of operator licensing were transferred to Region lllin FY 1982 and to Region 11in FY 1983. As a result of these

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R:: vision 6 changes, all regions had operator licensing authority in FY 1984. NRR provided oversight l

and guidance, including examination procedures and criteria.es (2) A study of the staffing of the operator licensing program and the qualifications and training of examiners was completed in November 1980 and documented in NUREG/CR-1750. Examiner standards were published in NUREG-1021.n2

)

(3) A plan for reporting operator errors and for acting on operator errors with respect to continuation of licensing was developed in NUREG/CR-1750." However, after review of this recommended plan, DHFS/NRR concluded that no further action was required."

CONCLUSION This item was RESOLVED and no new requirements were established."

ITEM l.A.3.3 REQUIREMENTS FOR OPERATOR FITNESS DESCRIPTION Historien! Backaround This NUREG-0660" item called for the staff to develop a regulatory approach to: (1) provide assurance that applicants for RO and SRO licenses were psychologically fit; and (2) prohibit licensing of persons with histories of drug and alcohol abuse or criminal backgrounds. The regulations were to be applied to all operating and future power plants.

A proposed rule addressing alcohol and drug use and the broader issue of fitness for duty of operating licensee personnel and contractors was forwarded to the EDO on April 16,1982, with the concurrence of several NRC offices. The proposed Fitness for Duty Rule was issued for public comment in the Federal Register on August 15,1982, with the public comment period extending to October 5,1982. A final rule package was completed on December 1, 1982, and a final rule was expected to be published by April 1,1983. The rule, when promulgated, would have required facilities licensed under 10 CFR 50.21(b) or 10 CFR 50.22 to establish and implement adequate written procedures to provide reasonable assurance that persons with unescorted access to protected areas of nuclear power plants, while in those areas, are not under the influence of alcohol, other drugs, or otherwise unfit for duty due to mental or physicalimpairments. Secondly, a proposed rule amending 10 CFR 73.56 regarding access authorization for nuclear power plants had not been completed, although a value/ impact analysis in support of the proposed rule had been prepared. Staff studies of the issue were 8

published in NUREG/CR-2075 " and NUREG/CR-2076.*

Safety Sionificance There could be significant damage if impaired personnel were performing critical safety operations. Legal and institutional problems could limit a thorough implementation of the proposed program. Given that there was an adequate program implemented at all power plants and integrated into overall plant operations, the new program would reduce operator error which, in turn, would lower the risk associated with operation of the plants.

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> - Possible Solutions This issue had two components: the first involved initial access to protected areas of nuclear

' power plants and the second involved continuing fitness for duty once initial access has been granted. The proposed Fitness for Duty Rule, issued for public comment 'on August 15,1982,

,was directed toward the second. component of this issue, mandating behavioral observation

= programs for power plants licensed by the NRC. Behavioral observation was also a part of the -

proposed Access Authorization Rule directed toward the first component of this issue.

The second component of this issue ~ dealt with limiting' access of psychologically unstable

~

Lindividuals to vital plant areas. This component was expected to have a major cost impact on the industry because this access authorization program was comprehensive in that it was aimed

~

at limiting the access to vital plant areas of disgruntled employees, psychologically unsuitable .;

employees, as well as personnel under the influence of drugs or alcohol.  ;

?The: access. authorization program had the following three parts: (1) background search;-

(2) psychological assessment; and (3) behavior observation. The first two parts would occur prior to granting an individual an unescorted access authorization to protected and vital areas, '

- and the:last part would be an ongoing activity for individuals who have oeen granted an unescorted access authorization. The background check would examine an individual's past for unstable activities, a criminal record, credit problems, and previous employment problems. It was established by NRC personnel that data on psychological screening showed that 2% to 3%

of white-collar workers were identified as unstable and, for blue-collar employees, the rate O

V ranged from 7% ~to 10%. These figures provided the background for the assumptions made in -

the evaluation below.

PRIORITY DETERMINATION

' This issue was assessed by PNL"in consultation with a number of engineers with expertise in reactor operator licensing, reactor operations, utility field work, and general reactor safety areas. )

Ammumotions -

The major, result of this issue was assumed to be a reduction in operator error. For some utilities, this new system would result in some reduction in operator error whereas, in others,

' it would have no discernible'effect. Based on engineering judgment, an average of 2% 'was-L arrived at by PNL.to apply to all operating and future plants. Thus, this issue assumed the implementation ~of the access authorization system at all 134 plants, either under construction L(63) Lor-in' operation (71), with average remaining lives of 28.8-y' ears for 90 PWRs and j

27.4 years for'44 BWRs.' Thus, the total remaining life of the affected plants was [(28.8)(90)

/+- (27.4)(44))RY or 3,798 RY. Neither the implementation, operation,' or maintenance of the solution' would involve any changes in occupational dose.

l For the analysis performed by PNL," Oconee-3 was taken as the representative PWR It was

? assumed that the fractional risk and core-melt frequency'reduct'ons for the representative BWR n '

A ) , (Grand _ Gulf-1) would be equivalent to those for the repre

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Revision 6 was conducted only for a PWR, but the fractional risk and core-melt frequency reductions were also applied to a BWR.

Frecuency Estimate All release categories were affected by this issue, but the principal Release Categories affected by the solution were 3,5, and 7; the numerical calculations were based on these categories.3 The calculated reductions in core-rnett frequency were 4 x 104 /RY for PWRs and 1.8 x 10 s /RY for BWRs.

Consecuence Estimate The dose calculations were based on a reactor site population density of 340 people per square-mile and a typical midwest meteorology, Based on the above core-melt frequency reduction and Release Categories, the total estimated public risk reduction was 16,000 man-rem. The occupational risk reduction for implementation, operation, and maintenance was zero.

Cost Estimate Industrv Cost: A value/ impact analysis in support of the anticipated Access Authorization Rule had been prepared by the staff and industry cost estimates had been developed. These cost estimates, which were reviewed and accepted by the Atomic industrial Forum (AIF), were as follows:

(1) For all operating plants, the implementation cost was $140,000/ plant and included the preparation of plant and associated procedures ($33,000), licensee management and clerical staff ($63,000), training to implement the behavioral observation program

($34,000), and storage for files ($10,000). The totalindustry implementation cost for operating plants was $(140,000)(71) or $9.94M.

(2) 'r all future plants (at which none of the employees were to be grandfathered), the i fementation costs were estimated to be $590,000/ plant. In addition to the costs no. ' above for operating plants, this implementation cost included background invewations ($375,000), review process and appeals procedures ($36,000), increased file sto. age requirements ($30,000), and miscellaneous crirninal checks with the FBI, etc. ($9,000). The totalindustry cost for future plants was ($590,000)(63) or $37.2M.

(3) The cost of operating of the access authorization system at each plant was estimated to be $300,000/ year. This cost included background investigations for new people as a result of employee turnover ($94,000), professional management and clerical staff

($63,000), a review and appeal process ($67,000), refresher training for old supervisors

($19,000), training of new supervisors ($9,000), plan maintenance and updates

($8,000), file storage ($39,000), and criminal history checks with the FBI for new people ($2,000). The totalindustry cost for operation and maintenance of the access authorization system was ($0.3M/RY)(3,798 RY) or $1,140M.

The totalindustry cost for the possible solution was $(9.94 + 37.2 + 1,140]M or $1,187M.

O 12/31/97 1.1.A.3-4 NUREG-0933

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Rsvision 6 i

NRC Conti Further development and issuance of the proposed plan was estimated to take.1.5

, man-years; at a rate of $100,000/ man-year, the cost for this effort was $150,000. The review

". and modification of the utilities' implementation plans was astimated to take L5 man-years.

4 For the 134 affected plants, this amounted to_0.6 man-week / plant. At a cost of $2,270/ man- -

week, the implementation cost was $182,500. Review of the operation and me:ntenance of l

the possible solution was estimated to require 1 man-week /RY for all plants. At a cost of-  :)

$2,270/ man-week, the total operation 'and maintenance of the solution was' 48.6M. Thus, the .

itotal NRC cost for the possible solution was $[0,15 + 0.1825 + 8.6]M or 48.9M.

Total Cost: The totalindustry and NRC cost associated with the possible solution was $(1,187

+ B.9)M or $1,196M.

Valua/im s,t Aarund

- Based on an estimated public risk mduction of.16,000 man-rem and a cost of $1,196M for a .

'possible solution, the value/ impact score was given by:

S =' 18.000 man-rem

$1,196M

= 13.4 man-rem /$M Other Considerations D 11 was estimated by cognizant NRC personnd that the Fitness for Duty Rule would have a-negative cost impact on operating licensees in the long run. The staff estimated that initial licensee burden to develop written procedures required by the rule would be approximately 1,200 man-hours over a six-month period at a total cost between $50,000 and $75,000, if no fitness for duty program existed at a licensee's facility. While utilities such as TVA claimed that alcohol abuse alone cost them approximately $18.5M annually, fitness for duty programs of the type envisioned by the Fitness for Duty Rute were expected to save costs through quicker identification of employees not fit for duty and through assisting these employees, in whom

{ considerable resources had been invested, in returning to high levels of productivity.

Nationwide, absenteeism due to alcohol and' drug abuse cost _U.S. industries an average of 4300 annually for. every worker._ Alcohol drug-abusers lose an additional 25% of their productive time when on the job, at an average annual cost to U.S. industries of approximately

$2,900/ abuser. At the. time this issue was evaluated, the total annual cost to U.S. industries -

was between $12 Billion and- $15 Billion. Wrich, in "The Employee Assistance Program; .

Updated for the 1980's," Hazelden,1980, reported that U.S. industries' received a retum of

$10 in decreased absenteeism, accidents, and increased productivity for every dollar spent on fitness for duty._

CONCLUSION Although the estimated risk reduction was 16,000 man-rem and the value/ impact score was

. only 13.4~ man-rem /$M, this issue was given~a__high priority ranking (see Appendix C) because

.g L of its advanced state of completion.

]

N -

[ [12/31/97/ 1.1.A.3-5. ' NUREG-0933 '

R vision 6 On October 24,1984, the Commission notified the staff that it would not promulgate a rule on fitness for duty for a minimum of two years but would issue a policy statement on the sub}ect. A proposed policy statement was submitted to the Commission in SECY-85-21."8 in a separate action, a notice withdrawing the final Fitness for Duty Rule was submitted to the Commission in SECY 85-21 A."' The proposed policy. statement (SECY-85-21" ) was reaffirmed by the staff in SECY-85-21B."' In recognition of the industry's efforts in establishing fitness for duty programs, the Commission approved a Policy Statement"7 in July 1986. Thus, this issue was RESOLVED and no new requirements were established."'

ITEM LA.3.4 LICENSING OF ADDITIONAL OPERATIONS PERSONNEL DESCRIPTION Historical Backaround This NUREG-0660 item sought to upgrade the operations performance in nuclear power plants by imposing licensing requirements upon other operations personnel in addition to ROs and SROs.

Safety Slanificance it was possible that, by undergoing licensing, personnel such as managers, engineers, and technicians would be better qualified and less likely to commit errors in performing their functions.

Possible Solution A study could be undertaken to determine which, if any, personnel should be licensed.

Licensing would then be required by the NRC for those additional personnel.

PRIORITY DETERMINATION Assumotions It was estimated that the effects of resolution of this issue would be minimal for many utilities since there were existing practices that went a long way toward ensuring that qualified and trained individuals were in responsible positions. It was assumed that additional licensing requirernents would produce some improvement by assisting in the screening of potentiaily poor performers from the operations staff. The net effect was estimated to be equivalent to a 2%

reduction in human error rates for reactor operators and maintenance personnel."

Frecuencv Estimate Based on the 2% reduction in human error rate, the Oconee 3 (representative PWR) risk equation parameters were adjusted. All Accident Sequences except V were assumed to be affected and all Release Categories were affected. The reduction in core-melt frequency for Oconee 3 was calculated to be 1.4 x 104/RY. The reduction in core-melt frequency for Grand Gulf 1 was then calculated by assuming that the fractional core-melt frequency reduction for 1.1.A.3-6 NUREG-0933 12/31/97

2 Revislan 6 m

[ the representative BWR would be skuivalent to the fractional reduction for a PWR. Therefob,

since the Oconee-3 fractional' reduction was 0.017, the core-melt frequency reduction for Grand Gulf-1 was calculated to be'6.3 x 104/RY.

s.

Consequence Eatimata The corresponding reduction in public risk for Oconee-3 was calculated to be 2.4 man-rem /RY and the public risk reduction for Grand Gulf-1 was calculated to be 2.7 man-rem /RY. The total risk. reduction for each type of plant was given as follows:

PWRs: (28.5 yrs)(95 reactors)(2.4 man-rem /RY) = 6.5 x 108 man-iem BWRs: (27 yrs)(49 reactors)(2.7 man-rem /RY) = 3.6 x 10 8man-rem Therefore, the total risk reduction for this issue was 1.01 x 10* man-rem.

Coat Estimata

- Industry cae: it was assumed that the required additional effort to license the majority of the ~ l operations personnel at a plant would be roughly equivalent to the existing licensing efforts for

. ROs and SROs; this was estimated to be $250,000/ plant. For operation, industry would have 1

to' provide new training staff,' staff time for training and exams, and administration; this was  !

estimated to be.450,000/RY. Therefore, the totalindustry cost was $250M.

NRC Cost: To implement the solution, qualification criteria, licensing exams, and procedures o O would have to be prepared and would be a major undertaking. The implementation cost was estimated to range from $20M to $50M; for analysis purposes, 435M was used. To operate with the new licensing requirements, 50 additional staff members would be required at a total

, cost of $5M/ year. To perform the annual operational needs of the program, funds would be

< needed for travel, publications, etc. This'was estimated to be an additional $2M/ year.

Therefore, the total NRC cost was approximately $240M.

Total Cost: The totalindustry and NRC cost associated with the possible solution was 4(250 l

+ 240)M or 4490M.

Value/Imnact Assessment Based on;an estimated public risk reduction of 10,100 man rem and a cost of $490M, the

- value/ impact score was given by:

I S = 10.100 man-rem '

4490M i

= 20 man-rem /4M

..Un,.rtaintv Because the estimate of.the value/ impact score relied heavily on the estimated value of the l p ;possible reduction in human error rate, the effective improvement could vary significantly.

a -.

% L ' ,

, s

!3 p L12/31/97L 1.1.A.3-7 ; l NUREG-0933 0 y L

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Rsvision 6  !

Other Considerations DHFS/NRR had been pursuing this issue and the Commission concluded

• that licensing of managers should not be required. The other portion of the issue (i.e., licensing of other personnel - engineers, maintenance personnel, etc.) was still under study and was to be concluded in FY 1983.

CpNCLUSION Although the value/ impact score was low, the potential risk reduction was considered and this issue was given a medium priority ranking (see Appendix C). However, in February 1985, the staff determined that there was insufficient evidence to support the licensing of additional plant personnel." Thus, this item was RESOLVED and no new requirements were established.

ITEM l.A.3.& eSTABI'SH STATEMENT OF UNDERSTANDING WITH INPO DESCRIPTION As part of the overall evaluation of the TMlincident,it was determined'8 that a statement of understanding was needed to address tb3 mutualintent of NRC and INPO concerning the extent to which NRC should review or rciy upon training, certification, and other INPO activities.

Consideration was alu to be given to providing alternative mechanisms for industry to inform NRC of its general gogress on needed safety reforms. It was intended that the statement of understanding wot id provide a basis for evaluation of any safety reforms or programs. Since there was no direct risk that could be attributed to this issue, it was considered to be a Licensing issue.

CONCLUSION A Memorandum of Agreementue between INPO and NRC was issued in April 1982; however, it did not specifically address training and certification. Following further revision, the EDO agreed

• with the Coordination Plan for NRC/INPO Training-Related Activities (Appendix Four to the Memorandum of Agreement)in November 1983. With the issuance of the Memorandum of Understanding, this Licensing issue was resolved.

REFERENCES

48. NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S.

Nuclear Regulatory Commission, May 1980, (Rev.1) August 1980.

64. . NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety issue Prioritization information Development," U.S.' Nuclear' Regulatory Commission, February 1983, (Supplement 1) May 1983 (Supplement 2) December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.

88. Memorandum for All Employees from H. Denton, "Regionalization of Selected NRR Functions," June 15,1982.

O 12/31/97- 1.1.A.3-8 NUREG-0933

h Rsvision 6

/9 89. NUREG/CR-1750, " Analysis, Conclusions, and Recommendations Concerning Operator

( ) Ucensing," U.S. Nuclear Regulatory Commission," January 1981.

98. - NUREG-0737, " Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory Commission, November 1980, (Supplement 1) January 1983.

148. '" Memorandum of Agreement Between the institute of Nuclear Power Operations and the U.S. Nuclear Regulatory Commission," (Rev.1) April 1,1982.

181. SECY-82-155, "Public Law 96-295, Section 307(B), Study of the Feasibility and Value of Licensing Nuclear Plant Managers and Senior Licensee Officers," April 12,'1982.

289. NUREG/CR 2075, " Standards for Psychological Assessment of Nuclear Facility Personnel," U.S. Nuclear Regulatory Commission, July 1981.

290. NUREG/CR-2076, " Behavioral Reliability Program for the Nuclear Industry," U.S. Nuclear Regulatory Commission, July 1981.

440. Memorandum' for W. Minners from D. Ziemann, " Schedules for Resolving and Completing Generic 1,ssues," April 5,1983.

593. SECY-84-76, " Proposed Rulemaking for Operator Licensing and for Training and Qualifications of Civilian Nuclear Power Plant Personnel," February 13,1984.

594. Letter to E. Wilkinson (INPO) from W. Dircks (NRC), November 23,198?.

651. NUREG-0985, "U.S. Nuclear Regulatory Commission Human Factors Program Plan," U.S.

Nuclear Regulatory Commission, August 1983, (Rev.1) September 1984, (Rev. 2) April 1986.

778. Memorandum for W. Dircks from H. Denton, "TMI Action item I.A.3.4," February 12, 1985.

956. Memorandum for V. Stello from H. Denton, "Close-out of the Division of Human Factors Technology TMI Action Plan items," January 6,1987.

962. NUREG-1021, " Operator Licensing Examiner Standards," U.S. Nuclear Regulatory Commission, October 1983.

963. SECY-85-21, " Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel,"

January 17,1985.

964. SECY-85-21 Ai" Withdrawal Notice: Fitness for Duty of Nuclear Power Plant Personnel,"

April 12,1985.

965. SECY-85-218, " Fitness for Duty of Nuclear Power Plant Personnel," August 26,1985.

N( 967. Federal Reaister Notice 51 FR 27921, " Commission Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel," August 4,1986.

.,12/31/97 1.1.A.3-9 NUREG-0933 '

' R vision 6 7

l

! n) ..

TASK f.A.4! SIMULATOR USE AND DEVELOPMENT The objectives of this task were to: (1) establish and sustain a high level of realism in the training and retraining of operators, including dearing with complex transients involving multiple

' permutations and combinations of_ failures and errors; and (2) improve operators' diagnostic -

j capability and general knowledge of nuclear power plant systems. '

ITEM 1.A.4.1f INITIAL SIMULATOR IMPROVEMENT ITEM l.A 4.1(1h SHORT-TERM STUDY OF TRAINING SIMULATORS

'l' DESCRIPTION This TMI Action Plan item called for a short-term study of training simulators to collect and

- develop corrections for identified weaknesses.

CONCLUSION.

l

- A study of training simulators was undertaken and NUREG/CR-14822se was published in June j 1980. Thus, this item was RESOLVED and no new requirements were established. j

/V)

ITEM l.A.4.1(2h INTERIM CHANGES IN TRAIN!NG SIMULATORS DESCRIPTION

- This TMI Action Plan item called for the development of requirements to correct specific training simulator weaknesses, based on the short-term study results from item I.A.4.1(1).

CONCLUSION

'This item was RESOLVED with the issuance.of Regulatory Guide 1.149 in April 1981 and new' requirements were established.

ITEM l.A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE The four parts of this item were combined and evaluated together.

DESCRIPTION i Historical Backoround

. Nuclear ' power plant simulators were recognized as an important part of reactor operator pf.

(

m

. training and this TMI Action Plan item called for a number of actions to improve simulators

. and their use.

p A 12/31/97' .

1.1.A.4-1 NUREG-0933 m >

R:; vision 6 There was significant interaction among the simulator-related action items and clear separation of this item was difficult. Item I.A.4.2 had a number of components dealing with long-term upgrades. The NUREG-0660 description called for research to: (1) improve the use of simulators in training operators; (2) develop guidance on the need for and nature of operator action during accidents; and (3) gather data on operator performance. Specific research items mentioned included simulator capabilities, safety-related operator action, and simulator experiments. The item also called for the upgrading of training simulator standards, specifically the updating of ANSI /ANS 3.5-1979. A regulatory guide endorsing this standard and giving the criteria for acceptability was also mentioned. The final portion of item I.A.4.2 called for a review of simulators to ensure their conformance to the criteria.

At the time the issue was initially evaluated, a significant portion of the activities to be conducted had been completed. For example, ANSl/ANS 3.5 was revised and issued in 1981 and Regulatory Guide 1.149, which endorsed this standard, had been published along with numerous research reports. It was clear that the regulations, the ANS standard, and the regulatory guide did not require a site-specific simulator.10 CFR 55 states that, if a simulator is used in training, it "shall accurately reproduce the operating characteristics of the facility involved and the arrangement of the instrumentation and controls of the simulator shall closely parallel that of the facility involved." ANSI /ANS 3.5-1981 called for a high degree of fidelity between the simulator and the " reference plant." However, there was no requirement that the reference plant be the same facility that the personnelin training would operate. Regulatory Guide 1.149'2' explicitly made the distinction stating "the similarity that must exist between a simulator and the facility that the operators are being trained to operate is not addressed in the guide and should not be confused with the guidance provided that specifies the similarity that should exist between a simulator and its reference plant."

The work that had been completed for item I.A.4.2(1) included the issuance of NUREG/CR-2353 8" (Volumes I and 11), NUREGICR-1908, NUREG/CR-2598,'" NUREG/CR-2534,

NUREG/CR-3092, and NUREG/CR-3123."* This item, however, had long-range requirements calling for: (1) the review of operating experience to provide data on operator responses; and (2) the design and conduct of experiments to determine operator error rates under controlled conditions. Items 1.A.4.2(2) and I.A.4.2(3) were completed with the issuance of Regulatory Guide 1.149.'** Item I.A.4.2(4) addressed the long-term training simulator improvement criteria ,

which were established in Regulatory Guide 1.149 and initiated in FY 1982. However, the l staff review of submittals from simulator owners for conformance with the criteria was an ongoing task in 1983. Therefore, the outstanding portions of this issue (the continuation of simulator research and the review for conformance to acceptability criteria) were evaluated.

1 Safetv Sionificance Use of simulators with high fidelity to the reference plant would significantly improve operator training in dealing with abnormal conditions thereby reducing operator error. Operators' performance under accident conditions was expected to be enhanced. Thus, a potential core-melt would be avoided and overall core-melt frequency reduced.

Possible Solution A possible solution was to establish a high level of realism in the training and retraining of plant operators by developing simulators with a high degree (,f fidelity to the ieference plant.

12/31/97 1.1.A.4-2 NUREG-0933

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)

" Revision 6 l l.

't PRIORITY DETERMINATION The assessment of this issue was conducted by PNL staff ** with ' experience'in reactor operator ,

' licensing, reactor operation, and general reactor safety, in consultation with General Physics Corporation. General Physics Corporation provided utility training services and had experience .

~

in reactor simulators; providing procurement and startup assistance, operation'and maintenance

. services, and simulator modifications. y i

' in the' assessment of this issue, it was necessary to acknowledge that many'of the TMl items j

associated with operator training were interrelated and that ranking problems surface when an .  ;

attempt .is 'made: to assess these ~ independently. For example, this item. was related to 1

ltems 1.A.2.6(1,2,3, and 5), which dealt with training improvements, including the enhanced use of existing simulators. l.A.4.1, dealt with initial simulator improvement, including short term and interim changes in training simulators. However, the final safety ranking of this issue was relatively insensitive to changes in the basic assumptions used to distinguish _ these interrelated issues by the very' nature of the ranking matrix. Therefore, it was possible to establish a priority.

ranking for this issue, despite the possible overlapping of potential benefits and costs with the

-other interrelated issues.

1 Assumotions it was assumed that the major effect of these issues, both in terms of safety benefit and cost incurred,1would be the enhancement of the level of realism imparted by simulators. The modeling capabilities'given under item I.A.4.1(2) and in ANSI /ANS 3.5-1981 reflected this

~

. feature.

lt was assumed that, in order to provide the intended level of realism, site-specific simulators

would be acquired. Such simulators would be significantly more realistic _when compared to the 1 specific faqilities, both in layout and operation, than existing generic simulators. In addition,

they were assumed to enhance transient and accident modeling capabilities.

. It was clear that provision of site-specific simulators, while not explicitly required, would meet

~ the requirements of item 1.A.4.1(2), the fidelity requirements of ANSI /ANS 3.5-1981, and the accurate reproduction requirements of 10 CFR 55. Less sweeping simulator enhancements might also. fulfill these requirements but would have to be decided.on a case-by case basis.

g Therefore, it was assumed that the enhancement would be effected by the introduction of site-

specific simulators.

,1 The public risk reduction (and occupational dose reduction due to accident avoidance) were _

associated with.the. reduction in operator error expected to result from the training and

^

z requalification of operators on improved simulators, inasmuch as any studies relating human error rates to the' realism of simulator training were not available, this assessment was based

.primarily'on PNI. engineering judgment. Therefore, it was estimated that a reduction in operator.

. error rate 6f 30% would result from the resolution of this issue. This estimate implied that, for specific instances, the' improvement could be much greater, but the 30% reduction was used as an estiniste'of the average improvement.

t-~ jThere were.90 PWRs and 44 BWRs' effected by this issue with average remaining lives of 28.8 .

Years end 27.4; years,= respectively? The representative' p~lants selected for. analysis were c+

<12/31/97f e ,

1. . 11.1.A.4-3 ~ NUREG-0933 s ,

M r i,, Q

Rsvision G Oconee-3 and Grand Gulf-1 for PWRs and BWRs, respectively. (It was assumed that the fractional risk and core-melt frequency reductions for Grand Gulf would be equivalent to those for a PWR which was calculated directly.)

Frecuency Estimate All release categories were affected by the resolution of the issue. The calculated core-melt frequencies were 8.2 x 104/RY for PWRs and 3.7 x 10'5/RY for BWRs. The reduction in these frequencies, based on the 30% reduction estimated for operator error, was 1.3 x 10~5/RY for PWRs and 5.9 x 104/RY for BWRs.

Conseauence Estimate The dose calculations were based on a reactor site population density of 340 people / square-mile and ;a typical midwest meteorology. The resulting total reduction in public risk was 150,000 man-rem.**

Cost Estimate Industrv Cost: The major effect of the resolution of these issues was assumed to be the acquisition and use of site specific simulators. The cost of such an undertaking would be substantial. If improved modeling changes were possible on existing simulators, the cost to industry would be substantially smaller. However, this was not clear at the time of the evaluation and it was assumed that new simulators would be required. (The impact of this assumption could be weighed subsequently in the final safety priority ranking. The assumption could be reevaluated at that time for any appropriate modifications.)

Assuming that new simulators would be required, the principalimplementation cost would be the purchase of the simulators and provision of the new training materials. The capital cost of a simulator was estimated to be $7M. The provision of training materials was estimated to be equivalent to a 7 man-year effort. j lt was assums .at a!! reactors, both operating and planned, would be affected. However, not ,

I every reactor would require a simulator. Many reactor sites have two or more reactors located together. If these reactors were sufficiently similar, a single simulator could serve them.

Examining the list of 134 operating and planned power reactors, it was estimated that 62 additional site-specific simulators would be adequate. This assumed that 20% of the potential i simulators were not required because either a site-specific simulator already existed or the plant in question was an older facility with limited remaining life.

The cost of the 62 new simulators spread over 134 reactors yielded $3.2M/ reactor in capital l cost and 3.2 man-year / reactor to provide new training materials. The operation and maintenance of the new simulators was estimated to require 3 man-years / simulator. Again, sharing the expense for 62 simulators over 134 reactors yielded 1.4 man-years / reactor. Industry may also experience costs stemming from participation in simulator experiments and research; however, these costs would be smallin comparison to the costs related to new simulators.

Based on these assumptions, the totalindustry cost was obtained as follows:

O

.12/31/97? 1.l.A.4-4 NUREG-0933

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f Ravision 6 t'

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~

.(1) implementation 1 Labor: I 7 man-vaar1Is2 simulatornil6100.0001 = $320,000/ plant C '

134 plants. ~ man-year" simulator?

. +1 Equipment: I62 mimulators1'l ' 47M 1 = . 43.2M/ plant 1

.134 plants - simulator Thus, the total industry implementation cost was (134 plants)$(0.32 + 32)M/ plant or

$470M.

(2)- Operation and Maintenance

[1.4 man-vaarlis100.0001((90 PWRs)(28.8 years) + (44 BWRs)(27.4 years))

reactor =  : man-year l -m .

- =) :4530M x

Theref. ore, the t'otalindustry cost was $(470 + 530)M or $1,000M. :s

.NRC Cost: There was no cost for development of a solution since all' work was essentially complete and a solution had been identified. The . principal costs were the continuation of-research and the conduct of the confirmatory reviews. No additional development costs were Lforeseen as ANSI /ANS 3.5 was being mv! sed and necessitated a revision to Regulatory-Guide 1.149.*

1 The continuing research was treated as an implementation cost. It was estimated to require one

- NRC man-year and $1M in contractor support. (This included the remaining costs associated 4 (with item 1.E.8.)~ The confirmatory reviews were also treated as an implementation cost and d" iwere. estimated to require 4 man-weeks / simulator, or a total of 248 man-weeks for the assumed 62 new simulators.

The operational' review cost to the NRC was minimal. It .was assumed that annually each simulator would be audited to ensure that reference plant updates had-been . adequately f represented' on the simulator. Such an annual review was estimated to require 2. man-

. weeks / simulator or 124 man weeks / year for all 62 new simulators assumed. NRC costs were

. estimated as follows: '

1 L(1)J Ji mniamentailon? .

' 6, , .. +

n

Continuing Research:01 man-vear .: = 0.33 man-week

- . 134 plants .  : plant i

Initial Simulator Reviews: 248 man-waaks = 1.9 man-weaka )

f -

.134 plants - ' plant k k^ . . .. .

Based on a total NRC . manpower of 2.23 man-weeks / plant, the implementation cost was

- -c ' <

y f~ ' $

fgiven by: q

3 ,

t i^

W+' 7 (12/31/9Q  : 1.1.A.4-5 L

- NUREG-0933.

, >g w y

( j

Revision 6 (2.23 man-weeks)( $2.2701(134 plants) = $678,300 plant man-week With contractor support estimated to be $1M, the total implementation cost was

$(0.6783 + 1)M or $1.7M.

f (2) Review of Ooeration and Maintenance i 2 man-weeks 1(67 simulators)( $2.2701 = $2,100/RY simulator-year 134 plants man-week The cost for review of operation and maintenance for all affected plants was ((90 PWRs)(28.8 years) + (44 BWRs)(27.4 years)]($2,100/RY) or $8M.

Thus, the totalimplementation, operation, and maintenance cost was $(1.7 + 8)M or $9.7M.

Total Cost: The totalindustry and NRC cost associated with the possible solution was $(1,000

+ 9.7)M or $1,010M.

Valuellmonet Assessment Based on an estimated public risk reduction of 150,000 man-rem and a cost of $1,010M for a possible solution, the value/ impact score was given by:

S= 150.000 m an-rem

$ 1,010M

= 148.7 man-rem /$M Other Considerations The estimated reduction in occupational L.,se was 820 man-rem, based on accident avoidance only, since there were no implementation or maintenance dose reductions associated with resolution.

CONCLUSION Based on the estimated risk reduction of 150,000 man-rem and the value/ impact score of approximately 150 man-rem /$M, this issue was given a high priority ranking (see Appendix C).

In view of the large estimated risk reduction, this ranking was essentially unaffected by any reasonable uncertainties in the cost estimates.

ITEM l. A.4.2fil RESEARCH ON TRAINING SIMULATORS This item was evaluated in item I.A.4.2 above and was given a high priority ranking (see Appendix C). In April 1987, the issue was RESOLVED with the publication of Revision 1 to '

Regulatory Guide 1.149* and new requirements were established.*5 12/31/97- 1.l.A.4-6 NUREG-0933

S- e Rsvision 6 ITEM l.A.4.2(2h UPGRADE TRAINING RIMULATOR STANDARDS !

This item was and RESOLVED with the issuance of Regulatory Guide 1.149" in April 1981 and new requirements were established.

~ ITEM l'[A:4.213h REGULATORY GulDE ON TRAINING SIMULATORS

- This item was RESOLVED with the issuance of Regulatory Guide 1.'149* in April 1981 and

new requirements were established.

' ITEM l.A.4.2(4h REVIEW SIMULATORS FOR CONFORMANCE TO CRITERIA

This item. was evaluated in item I.A.4.2 above and was given a high priority ranking (see

- Appendix C). Staff efforts in resolving the issue resulted in the publication of a rule and a simulation facility evaluation procedure.

When this item was originally identified in 1980, the staff's approach was to require a . submittal

- from each licensee in compliance with a regulatory guide (which later was issued as Regulatory

- Guide 1.149*) ad to conduct a review of each simulator; there was no simulator regulation in effect at that bme. However, in 1983, Section 306 of the Nuclear Waste Policy Act (Public

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Law 97-425) directed the NRC, in part, to establish " requirements for operating tests at civilian nuclear power plant simulators." This Congressional mandate had the effect of superseding the original intent of item I.A.4.2(4) and required the staff to develop regulations for simulators.

As a resultithe approach taken by the staff for the resolution of item I.A.4.2(4) was modified to comply with the Congr.essional mandate. The work scope was changed to reflect the fact that licensees, under the proposed regulation, would be required to certify their plant-referenced

' simulators to the NRC, and that NRC would perform an audit only when a need was identified, or upon request. Only in the case of those few licensees (estimated to be six), which were

- expected to seek NRC approval for a simulation facility that did not include a plant-referenced simulator, would the staff be obligated to review simulator documentation.

Tha final rule was published'* as.10 CFR 55.45 and states, in part: "The operating test will be; administered in .a plant walkthrough 'and in either (i) a simulation facility which the g- . Commissicn has approved for use after application has been made by the facility licensee, or (ii) a simtsation facility consisting solely of a plant-referenced simulator which has been certified to the Commission by the facility . licensee." In support of these regulations, the staff initiated

a program to develop a procedure for its evaluation of selected certified simulation facilities.

.This procedure was subjected to a pilot test prior to being issued in draft form for comment.

- As a result of comments; received, the procedure was revised and issued in final form as NUREG-12581** in December.1987. Thus, the item was RESOLVED and new requirements

were established.10981 m H

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r Revision 6 ITEM l.A.4.3: FEASIBILITY STUDY OF PROCUREMENT OF NRC TRAINING SIMULATOR DESCRIPTION The description of this issue in NUREG-0660 was as follows:

"In addition to the increased use of industry simulators for training of NRC staff (notably, the work by OIE with the TVA training center simulators), a feasibility study of the lease or procurement of one or more simulators to be located in the NRC headquarters area will be performed. These simulators would be used in familiarizing the NRC staff with reactor operations, in assessing the effectiveness of operating and emergency procedures and in gathering data on operator performance. The study will include development of specifications, development of procurement End commissioning schedules, estimation of costs, and comparison with other methods of providing such training for NRC personnel."

The intent of this issue was to improve the NRC staff's familiarization with reactor operations.

The study was an effort to establish the feasibility of procuring an NRC training simulator. The issue had no direct bearing on public risk reduction and, therefore, was considered to be a Licensing issue.

CONCLUSION Technical studies 2e2.2eue4 of the issue performed by BNL indicated that existing simulators had significant modeling limitations. It was established that the capability of existing simulators was not acceptable at any but near-normal operating conditions, and that the lack of technical capability during two-phase conditions was significant. These results had an adverse effect on the feasibility of a training simulator for the NRC staff. Thus, this Licensing issue was resolved.

ITEM LA.4.4! FEASIBILITY STUDY OF NRC ENGINEERING COMPUTER DESCRIPTION The purpose ** of this study was to fully evaluate the potential value of and, if warranted, propose development of an engineering computer that realistically modeled PWR and BWR plant behavior for small-break LOCA and other non-LOCA accidents and transients that may call for operator actions. Final development of the proposed engineering computer would depend on a number of research efforts.

Risk assessment tasks (interim reliability evaluation program, or IREP, for example) to define accident sequences covering severe core damage would also provide the guidelines for the experimental and analytical research programs needed to improve the diagnostics and general knowledge of nuclear power plant systems. The programs would assist the development and

' testing of fast running computer codes used to predict realistic system behavior for these multiple accident stur'jes. These codes would provide the basic models for use in the improved engineering comroter as waii es the capability for NRC audit of NSSS analyses. This issue had no direct effen on public risk reduction and, therefore, was considered a Licensing issue.

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Rsvision 6 CONCLUSION Reports **8 8" on the review of PWR and BWR simulators were completed by BNL while work on Plant Analyzers continued 'at BNL, INEL, and LASL RES believed that Plant Analyzers

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(Engineering Computer) would be helpful in uncovering potential operational safety problems in.LWRs, caused by operator errors or equipment malfunctions, which would lead to risk reductions through increased; operator awareness, improved procedures, and equipment redundancy.