Resolution of Generic Safety Issues: Appendix G: Generic Issues Program Current and Historical Procedures ( NUREG-0933, Main Report with Supplements 1–35 )

I. BACKGROUND

History

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

Long Quote

UNRESOLVED SAFETY ISSUES PLAN

Sec. 210. The Commission shall develop a plan providing for specification and analysis of unresolved safety issues (USIs) 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 public dissemination of information related to the issues as they progressed. This program was published in NUREG-0410³⁸⁷ in January 1978 and, shortly thereafter, the Commission issued a Policy Statement¹¹⁹⁰ on the Nuclear Regulatory Commission (NRC) "Program for Resolution of Generic Issues Related to Nuclear Power Plants."

(1) In 1977, all 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).¹⁴⁰ With increased confidence in this risk assessment 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 Plan⁴⁸ 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 (GSI). With this approach, 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-81-513.¹ In April 1983, this approach was refined and resubmitted to the Commission for approval in SECY-83-221.¹¹⁸⁸ After Commission review, approval to use the methodology was given in December 1983.¹¹⁸⁹

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, while still ensuring attention to issues that require it. The raising of the numerical safety thresholds was accompanied by strengthened attention to uncertainties and special considerations, to help recognize instances when a priority rank higher than the indication from the numerical formula was 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 simplification of the way in which costs were considered reflected the confirmation from experience that risk significance was indeed the primary factor in priority ranking, with a more bounded role for safety-cost trade-offs.

Operating 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 Guidance (NUREG-0885, Issue 2),²¹⁰ in addressing the area of Coordinating Regulatory Requirements (Planning Guidance, Item 5, p.6) called for "...a priority list of generic safety issues including TMI-related issues based on 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-83-221.¹¹⁸⁸

After another NRR reorganization in November 1985, the task of preparing and maintaining the list of GSIs and their priority was assigned to the Safety Program Evaluation Branch, Division of Safety Review and Oversight (SPEB/DSRO/NRR). Following an NRC reorganization in April 1987, this responsibility was assigned to the Advanced Reactors and Generic Issues Branch, Division of Regulatory Applications, Office 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 GSIs was an ongoing staff function that was reflected annually in the NRC Policy and Planning Guidance.²¹⁰ This document was superseded in 1987 by the NRC Five-Year Plan.

II. GENERIC ISSUES PROGRAM (1983-1999)

After issuance of the Policy Statement¹¹⁹⁰ in 1978, the NRC program to resolve generic issues underwent many reviews and changes. As a result, the Commission concluded in April 1989 that the 1978 Policy Statement no longer reflected the NRC's generic issues program and withdrew it from the public record.¹¹⁹¹

From 1983 to 1999, the generic issues program consisted 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. During this period, approximately 836 generic issues were processed in accordance with the steps outlined below.

After an issue was prioritized and approved for resolution, the first task was 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 varied in scope and depth in accordance with issue priority and the depth of information on a given issue. The second task involved development of a technical solution. Typically, the information used to resolve an issue came from experience data, experiments, tests, analyses, and probabilistic risk assessments (PRAs). The results of such work or the technical findings may have been published in contractor and staff NUREG reports which were made available through the NRC Public Document Room (PDR), Washington, D.C., or the National Technical Information Service, Department of Commerce, Springfield, Virginia.

In the final stage of resolution, the technical findings were used as a basis to develop a proposed resolution for the issue involving a change to NRC requirements or guidance. Several alternatives were 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 were used in selecting a proposed resolution. If a backfit was proposed, first, a determination was made as to whether the backfit was required to provide adequate protection to the health and safety of the public, or simply provided for enhancement of public health and safety. If it was determined that the backfit was necessary to provide an adequate level of protection, the backfit was imposed, regardless of the costs to achieve it. If it was determined that the backfit provided for enhancement of public health and safety, a generic analysis was required that treated the nine factors specified in 10 CFR 50.109(c).

Once the cognizant NRC Office Directors agreed to a proposed resolution, it was 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 were published in the Federal Register for public comment. Comments received were then incorporated, as appropriate, with the final product published in the Federal Register. Resolution of a generic issue took from several months to a few years, depending on the length of time required by the deliberations involved at each of the above steps.

Imposition

Imposition was the step in the generic issues program where each affected licensee and/or applicant was required or guided to prepare a schedule for implementing the generic issue resolution consistent with a Rule, Policy, Regulatory Guide, generic letter, bulletin, and/or licensing guidance developed during the resolution stage. Normally, NRC requirements, policies, and/or guidance did not provide for NRC consideration of a licensee's modifications prior to their implementation at an affected plant. This facilitated 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 were expected to submit (normally for NRC approval) their plans (including schedules) for plant modifications prior to their implementation. In all cases, licensees were expected to certify in writing to the NRC that plant modifications had been completed.

For the exceptional cases, the staff reviewed each applicant'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 modifications. For backfits, imposition was complete when each affected licensee had committed to compliance actions and schedules for implementing these actions. For forward-fits, the imposition of a generic issue resolution was complete when the new requirement or guidance became effective as an integral part of NRC regulations, policies, and/or guidance.

During this stage, a resolved GSI was identified as a Multiplant Action (MPA) for licensee action. The imposition status of all MPAs was tracked in the Safety Issue Management System (SIMS). This step was replaced by Regulation and Guidance Issuance (Stage 5) in MD 6.4.¹⁸⁵⁸

Implementation

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 required/committed actions, the implementation stage is complete, unless it is determined by the staff from subsequent verification inspection that additional licensee actions are needed for compliance. This step was retained as Implementation (Stage 6) in MD 6.4.¹⁸⁵⁸

Verification

III. PRIORITIZATION (1983-1999)

Purpose and Scope

The primary purpose of prioritization was to assist in the timely and efficient allocation of resources to those safety issues that had a high potential for reducing risk, and in decisions to remove from further consideration issues that had little safety significance and held little promise of worthwhile safety enhancement. However, issues of such gravity that consideration of immediate action was called for were excluded from prioritization because of the compressed time scale in which decisions for such issues had to be made. Generally, immediate action took the form of a Bulletin or Order. Both operating and future plants were considered in the priority ranking process.

Prioritization focused 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 was used to identify changes in existing requirements that may have significantly reduced the impact (usually cost) on licensees without any substantial change in public risk. Issues of this type were classified as Regulatory Impact issues (RI) to clearly differentiate them as not improving the safety of nuclear power plants but, nevertheless, possibly worthwhile.

In order to identify GSIs, all issues originated in accordance with OL #1¹¹⁹² were reviewed to determine their safety significance. Issues that primarily concerned environmental protection or the licensing process and did not involve significant safety improvement elements were classified accordingly and noted for separate consideration outside the GSI priority ranking process. These issues were classified as either environmental issues or licensing issues. Environmental issues (EI) involved impacts on the human environment and the values sought to be protected by the National Environmental Policy Act (NEPA). Licensing issues (LI) were not directly related to protecting public health and safety or the environment, but related 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 contained the following groups:

(1) TMI Action Plan items identified for development in NUREG-0660⁴⁸; these issues are covered in Section 1. The priority recommendations in this report excluded those issues that were designated for implementation in NUREG-0737.⁹⁸

(2) Task Action Plan items identified in NUREG-0371² and NUREG-0471,³ 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, issues designated as USIs were excluded from prioritization because of the high-priority attention they were given based on priority decisions previously made.

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

(4) Human Factors Program Plan (HFPP) items identified for development in NUREG-0985⁶⁰³; these items are covered in Section 4.

(5) Chernobyl Issues identified in NUREG-1251¹¹⁹⁵; these issues are covered in Section 5.

A comprehensive listing of all issues in the above five groups is given in Table II which includes the following information for each issue: (1) the NRC person responsible for the prioritization evaluation; (2) the lead NRC office, division, and branch responsible for 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 issues in each category is shown in Table III. 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 the Work Was Done

The work was done, 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 was 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, reviewed and applied 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 contributed to the assurance that the analysis was complete and accurate, and that the judgments were soundly based. This review was done in two stages. First, each analysis was reviewed by the NRC organizational unit or units whose area of responsibility or specialized knowledge was substantially involved. Second, any comments made were then resolved, where practical, and factored into the analysis, as appropriate. Upon completion of peer review, the analysis was then finalized and prepared for approval by the responsible Office Director. Once approved, it was placed in the PDR and published in a supplement to this report, after which, additional comments from the ACRS, the industry, and the public were considered in any further reassessment of the issue's priority.

Priority Categories: Their Meaning and Proposed Use

Four priority rankings were used: HIGH, MEDIUM, LOW, and DROP. They were 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 HIGH priority ranking meant that strong efforts to achieve the earliest practical resolution were appropriate. This was because: (a) an important safety concern may have been involved (though generally the concern was not severe enough to require prompt plant shutdown); or (b) the uncertainty of the safety assessment was unusually large and an upper-bound risk assessment would have indicated an important safety concern.

 The priority rank is always HIGH when any of the following risk (or risk-related) thresholds are estimated to be exceeded (or when extraordinary uncertainty suggests that they may well be exceeded):

1. 11000 person-rem estimated public dose per remaining reactor lifetime
2. 50,000 person-rem total estimated for all affected reactors for their remaining lifetime (e.g., 500 person-rem/reactor for 100 reactors)
3. 10^-5/reactor-year large-scale core-melt
4. 5 x 10^-4/year large-scale core-melt (total for all affected reactors)

All unresolved HIGH priority issues were periodically reviewed in accordance with the criteria stated in NUREG-0705 "Identification of New Unresolved Safety Issues Relating to Nuclear Power Plant Stations," issued March 1981⁴⁴ for possible designation as USIs. A USI is defined as a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants affected.¹⁸⁶

A MEDIUM priority ranking meant that no safety concern demanding high-priority attention was involved, but there was believed to be potential for safety improvements or reductions in uncertainty of analysis that may have been substantial and worthwhile.

The priority rank is least MEDIUM priority:

1. 10 or more percent of the always-HIGH criteria

Efforts at resolution were planned over the ensuing years, but on a basis that did not interfere with pursuit of HIGH-priority GSIs or other high-priority work.

A LOW priority ranking meant that no safety concerns demanding at least MEDIUM-priority attention were involved, and there was little or no prospect of safety improvements that were both substantial and worthwhile.

The priority rank is at least LOW priority:

1. 1 or more percent of the always-HIGH criteria
2. less than 10 percent of the always-HIGH criteria

When the prioritization process resulted in a LOW priority ranking for an issue, approval of this ranking by the responsible Office Director signified that the issue had been eliminated from further pursuit. However, in accordance with Staff Requirements Memorandum (SRM) 871021A,¹⁴⁹³ the staff conducted a periodic review of existing LOW-priority GSIs to determine whether there was any new information that would necessitate reassessment of the original prioritization evaluations.

The DROP category covered proposed issues that were without merit, or whose significance was clearly negligible. Issues were also DROPPED from further consideration if it was determined that their safety concerns had been addressed in previously prioritized or resolved issues.

The priority rank is a DROP category:

1. less than 1 percent of the always-HIGH criteria

When the prioritization process resulted in a DROP priority ranking for an issue, approval of this ranking by the responsible Office Director signified that the issue had been eliminated from further pursuit.

 The figure below depicts priority ranking for determining High (H), Medium (M), Low (L), and Drop (D) categories.

An issue was considered RESOLVED, indicated by NOTE 3 in Table II, when its resolution resulted in either: (a) the establishment of regulatory requirements or guidance (by Rule, SRP¹¹ change, or equivalent); or (b) a documented authoritative decision that no change in requirements was warranted. Priority rankings were not assigned to issues that had been resolved. However, in those cases where issues were resolved after having been identified for further pursuit by the prioritization process, the related calculations were retained in the text of this document for future use.

Priority rankings were not assigned to issues that were nearly-resolved (denoted by NOTES 1 and 2 in Table II) because approval of changes to requirements, based on the resolution of an issue, required that a detailed value/impact evaluation of the safety benefit, implementation costs, and other relevant factors be made. Prioritization would have duplicated this value/impact analysis, but in a less comprehensive manner. Therefore, the effort that would have been needed to prioritize an issue was 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 was considered to be identified, indicated by NOTE 1 in Table II, when a possible technical resolution was under evaluation and the evaluation was nearing completion. Further work may have been required as part of the review and approval process before a change in requirements or guidance was issued. Resolution of an issue was considered available, indicated by NOTE 2 in Table II, when proposed or recommended changes to requirements or guidance were documented in a NUREG report, NRC memorandum, Safety Evaluation Report (SER), or equivalent.

Priority rankings were also not assigned to those issues whose safety concerns were determined to be covered (at the time of prioritization) in other issues of broader scope that were being prioritized, or were being resolved. Issues in this category were integrated into the issues of broader scope. A detailed listing of all such issues is given in Table V.

Criteria for Assigning Priorities

1. Basic Approach

The method of assigning priority rank involved two primary elements: (i) the estimated safety importance of the issue; and (ii) the estimated cost of developing and implementing a resolution. Special considerations may have influenced the proper use of the estimates. These elements were applied as follows:

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

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

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

(d) A numerical impact/value ratio was calculated by dividing the estimated cost entailed by the estimated potential risk reduction. The ratio measured the safety value received in return for the cost impact incurred.

(e) A priority rank (HIGH, MEDIUM, LOW, or DROP) was obtained by application of criteria in which both the safety significance of the issue and the impact/value ratio were taken into account. The ratio was not always directly applied to determine the priority rankings. In some cases, the safety significance of the issue was so great that it demanded a HIGH priority, or so minor that only a LOW priority (or a decision to DROP) was warranted irrespective of the impact/value assessment.

(f) The priority ranking was reviewed and modified, if appropriate, in light of any special factors (discussed below) that: (i) might bring into question the applicability of the necessarily simplified calculation technique; and (ii) call for special consideration of NRC management decisions or large uncertainties in the quantitative estimates.

In summary, while the method had a quantitative emphasis, the calculated numerical values were used as an aid to judgment and not as determinative of the ranking results. The nature of the specific issue, the quality of the data base, and the scope of the necessarily limited analysis determined in each case the dependability of the numerical indications as a judgment aid.

2. Safety Significance

The safety significance of an issue was represented by the reduction in risk that resolution could affect. Risk was 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 was important within the necessarily rough risk estimates, the risks were summed.

The potential risk reduction calculated in this way was used in calculating the impact/value ratio as part of the simplified impact/value analysis, discussed in Paragraph III.3 below. It was also used directly as a measure of safety significance, as discussed in Paragraph III.4 below, in arriving at a priority rank that was influenced by the safety significance of an issue, as well as by the estimated value/impact relation of a projected solution, or was determined on the basis of safety significance alone.

The person-rem-based risk reduction estimate may not have been the only appropriate measure of an issue's safety significance in all cases. For example, when a possible core damage was involved but release outside containment would be minor or highly improbable, contribution to the core-damage probability may well have been more indicative of safety significance. Provision was made, as described in Paragraph III.4 below, for use of alternative measures of safety significance in determining a priority ranking when such alternative measures were useful.

3. Impact/Value Relation

(a) The Impact/Value Ratio Formula

To the extent reasonably possible, quantitative estimates were made of the possible solutions to a GSI by calculating an Impact/Value Ratio that reflected the relation between the risk reduction value expected to be achieved and the associated cost impact. The formula for the impact/value ratio (R) was:

R = Safety Benefit

where the safety benefit was the estimated risk reduction (event frequency x public dose averted) that may have been achieved, and the cost was that thought necessary to develop and implement a resolution in the number of plants involved. The scoring computation for any issue was then:

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 frequency reduction (event/reactor-year); D = public dose from the radioactive material released from containment (person-rem); and C = total cost of developing and implementing the resolution of the issue for all plants affected (dollars).

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

Simplified calculations usually sufficed, since only an approximate impact/value ratio was required. Reference was made to the current version of the Value-Impact Handbook,⁹⁷⁰ where necessary, to supplement 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, made any of various possible impact/value score formulas necessarily imperfect. Accordingly, provisions were made to compensate for those imperfections to the extent practical (as discussed in Paragraph III.5 below).

The formula selected measured a total-cost/total-safety-benefit relation. As discussed herein, it was applied within limits set by other possible considerations where a safety issue was either too important to depend on safety-cost tradeoffs, or too trivial to merit attention at all. Two principal arguments favored a formula of this type:

(1) The denominator was designed as a direct measure of the safety values that it is NRC's primary mission to protect. The numerator was designed to measure the overall cost impact, including industry as well as NRC costs, and should thus reflected the entire public interest in economy. The resulting impact/value ratio, subject to the stated caveats, should have reasonably approximated measuring the overall public interest in safety value received for total resources expended.

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

(c) Risk Estimates

The risk estimates developed for GSIs were useful as rough approximations for comparative purposes, but were 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 were not intended to be applied as impact/value determinations for any regulatory proposal that may ultimately result from efforts to resolve an issue. In addition, the assumed resolutions were not intended to prejudge the final resolutions, but are only assumptions that are necessary to perform quantitative analyses.

The basis of frequency estimates generally involved the following:

(1) Identification of the specific events which were the basis for the concern, for which the consequences were to be established, and which were 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 the numbers used

(4) Consideration of the probability of common mode as well as random independent failures.

Where possible, numerical estimates were based on operating experience, usually Licensee Event Reports (LERs). Other sources included prior PRAs and other risk and reliability studies. Some numbers were based on engineering judgment; in such cases, the basis for that judgment was stated.

For the identified end event(s), the expected radiological consequences were expressed in person-rem generally based on the radioactive release categories described in WASH-1400¹⁶ (Appendix VI, pp. 2-1 to 2-5), reproduced as Appendix A to this report. Exhibit A below gives estimated Curies released and approximate population doses for each release category.

Exhibit A - Estimated Release Public Dose** Category (Curies) (Person-rem)

PWR-1   1.2 x 10⁹ 5,400,000

PWR-2 9.3 x 10⁸ 4,800,000

PWR-3 5.2 x 10⁸ 5,400,000

PWR-4 2.8 x 10⁸ 2,700,000

PWR-5 1.3 x 10⁸ 1,000,000

PWR-6 1.0 x 10⁸ 150,000

PWR-7: 2.1 x 10⁶ 2,300

PWR-8:* 7.7 x 10⁵ 75,000

PWR-9:* 1.1 x 10³ 120

BWR-1 1.1 x 10⁹  5,400,000

BWR-2 1.1 x 10⁹  7,100,000

BWR-3 5.0 x 10⁸  5,100,000

BWR-4 2.1 x 10⁸  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) may be updated in the future to be consistent with the ongoing evaluation to revise the Source Term following a postulated severe accident.

 The following assumptions and parameters were used for the calculations at this stage (Step (b) described under "Basic Approach"):

• Consequences were represented by the whole body population dose (person-rem) received within 80.5 kilometers (kms), (50 miles) of the site.
• An exclusion area of 0.5 mile was assumed, with a uniform population density of 340 persons per square mile beyond 0.5 mile. This was the mean 50-mile radius population density projected for 2000 (NUREG-0348, "Demographic Statistics Pertaining to Nuclear Power Reactor Sites," issued November 1979, p.T52).⁷⁰

• The evacuation of people was not considered because of the possible large variations in evacuation capability for each plant site.
• The basis for the tabulated numbers included all exposure pathways 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.)
• Meteorological data were taken from the U.S. National Weather Service station at Moline, IL.

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

An estimated occupational dose of 20,000 person-rem from post-accident cleanup, repair, and refurbishment was also considered.

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

Where more direct issue-specific ORE information was lacking, dose estimates were obtained by assuming an average dose rate of 2.5 millirem/hour (based on the PNL analysis⁶⁴ cited above) and multiplying by the estimated number of man-hours involved.

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

The sum of the estimated risks of all the separate issues were likely to exceed the existing estimate of the total risk of nuclear power plants because of two factors. First, individual accident sequences could have been affected by more than one issue. The resolution of one issue would have reduced the probability or consequences of a certain set of accident sequences. Some or even all of these sequences could have been the same as some or even all of the sequences affected by another issue. However, issues were assessed independently, and this interaction of their risk significance was not ordinarily considered. This interaction was strongest for issues related to human factors, since human error affected almost all sequences. The sum of the reductions in core-melt frequency estimated for all of the human factors-related issues may have been 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 were much less strongly interrelated.

(d) Cost Estimates

Because cost estimates were used here only in relation to risk estimates which were generally subject to more or less wide uncertainties, only approximate costs were needed.

No separate estimates were generally made for offsite property damage; reasonably conservative use of the public dose estimates was an adequate surrogate in this application. Furthermore, there was no readily-available data on offsite damage that was realistic and detailed enough to make estimates meaningful, reasonably accurate, and generically applicable. If unusual or special offsite effects were not adequately represented by the public dose in some issues, this fact was considered separately and explicitly in evaluating such issues.

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

NRC costs included the following: (1) issue identification, analysis, resolution, and report issuance; (2) research to establish proposed 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 cost of NRC professional time was based on $100,000 per person-year.

The costs to industry generally consisted of some combination of the following: (1) licensing; (2) design; (3) equipment procurement; (4) installation; (5) testing, inspection, monitoring, and periodic maintenance; and (6) plant downtime to effect a change, taken as the cost of replacement power at $300,000/day. Industry manpower costs were ordinarily taken as $100,000 per person-year.

Averted plant damage costs may have affected the priority of a GSI. Estimates for such averted costs were multiplied by the accident frequency and used as negative costs, i.e., subtracted from the (positive) costs of implementing the resolution of the issue.¹⁴⁷³ The averted costs may have included those of averted equipment failures, limited-time plant outage, or limited plant-contamination cleanup. In the extreme, they also included averted permanent loss of the plant, estimated at approximately $2 billion present worth. This estimate for a "generic" plant included 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 was multiplied in each case by the reduction in frequency of such events that would be brought about by resolution of the GSI. The plant loss estimate included allowance for typical plant age at the time of the accident, as well as replacement power costs together with apportioned cost of a replacement plant. The plant-wide cleanup estimate reflected cleanup to the point at which the plant was ready for decommissioning or refurbishing for restart.³⁹³ Refurbishing costs, when restart was more economical than decommissioning, depended on the nature of the accident and ranged from a fraction of the total plant loss figure to a cost approaching that figure.

Some fixed costs were one-time, initial costs; others may have occurred at future times. Future costs were discounted to present worth at a 7% rate. Where costs were continuous or periodically recurring throughout a plant's remaining life, the periodic cost was 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) Uncertainty Bounds

Major sources of uncertainty in the priority score were identified and judgments as to their quantitative significance were indicated as information warrants. Where data warranted, the method described in NUREG/CR-2800,⁶⁴ Section 5, for the general case of combining uncertainties for random variables with unknown distributions (as well as some special cases) were used. [See also Paragraph III.5(a)]. Most often, however, a rigorous uncertainty analysis was not warranted. In most cases, the uncertainty in the point estimates of risks and costs was known to be large. However, sufficient information was not usually available to make a meaningful quantitative analysis of the uncertainty bounds of these point estimates. Decisions were 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) was partially based on the recognition that the uncertainties are large. In cases where uncertainty had a special character or importance, this was discussed and considered in the conclusion of the analysis of the GSI.

4. Priority Ranking

(a) Priority Ranking Chart

A chart showing how the tentative priority rankings were 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 III.4(b) and III.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.¹⁶⁸⁹

(b) Preliminary Screening for Safety Significance

The determination of a priority rank started with a triage based on safety significance, i.e., the incremental risk associated with the issue. For a reduction in (delta) core damage frequency CDF greater than 10^-4 per reactor-year (RY), a HIGH priority was assigned on the basis of safety importance alone, regardless of other considerations, such as an initially estimated high cost, which might result in a low priority score.

At the other extreme, an issue's safety significance could have been too minor to warrant diversion of attention from more important safety issues, even if it had a low impact/value ratio because an inexpensive solution was believed to be available. Below a minimal safety significance threshold, the priority was always DROP; where the potential risk reduction was trivial, there was no basis for regulatory action on safety grounds.

In between, there may have been issues of less extreme importance or unimportance, for which a HIGH, MEDIUM, LOW, or DROP priority may have been appropriate, based on consideration of the impact/value relation as well as safety significance. As indicated in Figure 1, a HIGH priority was assigned to an issue exclusively on the basis of a high safety significance; the threshold shown on the chart is (delta) CDF=10^-4 /RY. For an issue with a safety significance lower than the threshold for an always-HIGH priority but at least 10% of that threshold ((delta) CDF=10^-5 /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 was always DROP for (delta) CDF<10^-7 /RY. Cost trade-offs entered in the 10^-7 to 10^-4 /RY (delta) CDF 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 (delta) in CDF attributable to resolution of the issue. This was often the most useful safety significance measure in GSI prioritization, though for some issues other measures may have been required or appropriate. For example, a measure based on radiological consequences (probability-averaged over the remaining reactor life) was used when the issue under consideration involved containment bypass, or related to containment performance or other features or actions to mitigate the radiological consequences of a core damage. Also, the thresholds may have needed to accommodate the possible influence of the number of reactors affected on the appropriate priority ranking. Therefore, Figure 1 was repeated in Figure 2, with auxiliary abscissae providing additional measures of safety significance. These were used when the principal abscissa was inapplicable, or when an auxiliary abscissa led to a higher priority indication. Thus, the abscissae for total effect on all plants were considered when more than 30 plants were affected.

(c) Impact/Value Ratio Thresholds

When the safety significance was in the intermediate range discussed above, (i.e., (delta) CDF between 10^-7 and 10^-4 /RY, or between 0.1% and 100% of the threshold for an always-HIGH priority), the impact/value ratio (R) was taken into account in the ranking indicated by the chart (Figure 1). This was done as follows:

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

(2) In the range of 1% to 10% of the threshold for an always-HIGH priority, the indicated priority was MEDIUM if R was below $2,000/person-rem; otherwise, the indicated priority was LOW.

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

5. Other Considerations

The formula-based rankings represented the primary concern of the NRC: public safety. The secondary concern was the impact on licensees, evaluated in terms of cost. However, the tentative priority rankings were 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 may have caused added difficulty in priority assignment. While the formula-based tentative rankings generally indicated that the safety significance was sufficient to justify NRC action, other considerations not adequately reflected, or not reflected at all, in the numerical formula were often needed to corroborate or adjust the results. Decision-making was helped by explicit identification of such other considerations and explanation of their bearing on the resulting final priority ranking, whether the effect was one of corroborating or of changing the estimates.

Listed below are some factors that may have been important in arriving at a sound priority ranking, and may have led to adjustment of a tentative, formula-derived ranking. Possible effects of occupational doses and uncertainty bounds [1(a)(1), (a)(2), and (b)(1) below] required particularly careful consideration for all issues. The factors listed were not considered all-inclusive. Others thought significant were discussed and, when practical, quantified appropriately in the overall risk significance measure and impact/value ratio along with their associated uncertainties. Sometimes, there were special considerations that were 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 related to safety were given the most consideration. The following is a partial list of other factors 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-5382¹⁵⁶³ and an RES report.¹⁵⁶⁴ 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:

(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.

(2) Situations where uncertainty is extraordinarily large (in accident probability, consequences, or cost, or any or all of these). If there 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 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.

(6) Potential for human intervention, using available equipment.

(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 was assigned. Thus, where no solution was evident, assignment of a priority consistent with the safety significance of the issue may have led to a search for resolution or mitigation at an acceptable cost. Generally, when uncertainties narrowed or perceptions changed in the course of time, the priority rankings were reexamined in the light of new developments and retained or changed. When different classes of plants were expected to be very differently affected by a potential resolution, the priority assignment was governed by the class of plants for which resolution was most worthwhile and urgent. (Resolution in such cases could have involved a new requirement for some class of plants and no action for others.) Where resolution differed for different classes of plants, differing priorities were assigned.

6. Concluding Remarks

The criteria and estimating process on which the priority rankings were based were neither rigorous nor precise. Considerable application of professional judgment, sometimes guided by good information but often tenuously based, occurred at a number of stages in the process when numerical values were selected for use in the formula calculations, and when other considerations were taken into account in corroborating or changing a priority ranking. What was important in the process was that it was systematic, that it was guided by analyses that were as quantitative as the situation reasonably permitted, and that the bases and rationale were explicitly stated, providing a "visible" information base for decision. The impact of imprecision was blunted by the fact that only approximate rankings (in only four broad priority categories) were necessary and sought. Beginning in June 1999, the above method of prioritizing generic issues was replaced with the screening process of MD 6.4.¹⁸⁵⁸

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 III. 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); (2) USI, HIGH-, and MEDIUM-priority issues that are being resolved; (3) nearly-resolved issues (NOTES 1 and 2); (4) issues 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. Tables II and III, and Appendix B also incorporate the results of those generic issues processed in accordance with MD 6.4¹⁸⁵⁸ since 1999.

IV. GENERIC ISSUES PROGRAM (1999-2007)

Introduction

In accordance with Section 210 of the Energy Reorganization Act of 1974, the NRC documented progress on the resolution of USIs in each NRC annual report to Congress. However, with the passage of the Federal Reports Elimination and Sunset Act of 1995, the statutory requirement to send Congress an NRC annual report ended on December 21, 1999. In accordance with SECY-00-0038, "The 1999 NRC Annual Report," dated February 15, 2000¹⁸⁵⁹ the last annual report to the Congress that included unresolved GSIs was the 1999 edition.

Beginning in 1999, all new GIs identified were subjected to the process delineated in NRC Management Directive (MD) 6.4, "Generic Issues Program."¹⁸⁵⁸ The Generic Issues Program (GIP) was designed to provide internal guidance for determining whether a candidate GI represents an adequate protection issue, a substantial safety enhancement issue, or a reduction in unnecessary regulatory burden issue. In addition, the GIP identifies cost effective solutions to GIs, implements, and then verifies the adequacy of solutions for GIs, as appropriate. Thus, the GIP provides an opportunity for the NRC and Agreement State staff and other parties to recommend safety or security-related (or reduction in unnecessary regulatory burden) improvements to the agency's regulations and/or implementation of these regulations. Candidate generic issues may arise from many sources, including safety evaluations, operational events, or even suggestions on the part of individual staff members, outside organizations, or members of the general public. Additionally, new issues identified as USIs or any staff concerns that are raised as part of the NRC's Differing Professional Opinion (DPO) program may also be addressed under the GIP. The staff periodically conducts reviews of the open GIs to identify USIs. Detailed staff guidance is provided in Appendix B, "Unresolved Safety Issue Assessment Criteria," of MD 6.4.

Because of the varying technical disciplines and levels of difficulty encompassed by GIs, the processing of GIs does not lend itself to a strict, proceduralized process. The guidance in MD 6.4 is intended to provide a useful, consistent framework for handling, tracking, and defining the minimum documentation associated with the processing of GIs.

- Only potential adequate protection, substantial safety enhancement, and reduction in unnecessary regulatory burden issues are subject to the GIP process.

- Resolution of a GI may involve developing and imposing new or revised rules, developing new or revised guidance, revising the interpretation of rules or guidance, or providing information for voluntary actions.

- Resolution of a GI may affect licensees, certificate holders, or other entities regulated by or subject to NRC's regulatory jurisdiction.

Overview of Generic Issues Program Stages

The GIP consists of the following eight stages:

1. Identification

Candidate GIs may be identified by organizations or individuals internal or external to NRC, including the NRC staff, the

Agreement State staff, the ACRS, the Advisory Committee on Nuclear Waste (ACNW), the Advisory Committee on the Medical Uses of Isotopes (ACMUI), licensees, certificate holders, industry groups, or the general public.

If any identified candidate GI has the potential for involving an adequate protection issue, prompt actions is taken to evaluate the issue and to initiate any necessary compensatory measures.

Candidate GIs may be identified by NRC or Agreement State staff during routine activities. Sources of candidate GIs include, but are not limited to, NRC staff concerns; public concerns; licensee event reports; morning reports; inspection reports; investigation reports; accident sequence precursor program reports; major studies; allegation reports; component failure reports; 10 CFR Part 21, "Reporting of Defects and Noncompliance," reports; industry reports; and reports of operational events at foreign facilities.

Guidance for identifying GIs from operational safety data reviews is contained in Management Directive (MD) 8.5, "Operational Safety Data Review." ¹⁹²⁷

Candidate GIs are submitted to the GIP Manager in RES, who forwards them to either the Reactor Generic Issue Review Panel or the Materials Generic Issue Review Panel, as appropriate. For candidate GIs that involve both program areas, the GIP Manager consults with the program offices to establish a combined review panel including representatives of the NRR, the Office of Nuclear Material Safety and Safeguards (NMSS), and RES. For security-related candidate GIs, NSIR participation is required.

The issues identified as Unresolved Safety Issues (USIs) or any staff concerns identified as part of the Differing Professional Opinion program may also be addressed under the GIP.

2. Initial Screening

During initial screening, the appropriate Generic Issue Review Panel assesses whether the candidate GI should be processed in the GIP, should be excluded from further analyses, or should be sent to another NRC program for review. Also, the scope of the candidate GI (and thus the GI) is defined at this stage.

Initial screening is complete after the appropriate Generic Issue Review Panel reviews the information submitted in accordance with Appendix A of MD 6.4, including any other supporting documentation, as well as any staff-generated screening analysis of the candidate issue, and submits its findings and recommendations to the Director of RES for reactor issues or to the Director of NMSS for materials issues.

For reactor candidate GIs, risk and cost benefit thresholds are provided in Appendix C of MD 6.4, "Criteria and Guidance for Technical Assessment of Candidate Reactor Generic Issues." During initial screening (Stage 2), the Reactor Generic Issue Review Panel uses, to the extent practicable, Appendix C of MD 6.4 in a comparative manner to determine whether the issue should be excluded from further analyses, or continue on to Stage 3, technical assessment, in which a quantitative analysis would be performed. For materials candidate GIs, the initial screening stage may be folded into the technical assessment stage. Appendix F of MD 6.4, "Criteria and Guidance for Technical Assessment of Candidate Materials Generic Issues," provides guidance on the conduct of panel meetings.

Figure A.G.1, "Candidate Generic Issue Screening Process," and Appendix G of MD 6.4 provide the questions that must be addressed during the GI classification process, primarily in Stages 2 and 3 of the GIP.

On the basis of established risk thresholds or engineering judgment, the Reactor or the Materials Generic Issue Review Panel assesses whether the candidate GI has the potential to be classified as either an adequate protection, a substantial safety enhancement, or a reduction in unnecessary regulatory burden issue. (The actual classification into one of these categories will be made at the technical assessment stage.) The Reactor or the Materials Generic Issue Review Panel makes its assessment on the basis of information readily available or easily obtained with reasonable resources.

For a candidate GI, either the Reactor or the Materials Generic Issue Review Panel, as appropriate, issues an initial screening memorandum consisting of a forwarding note with attached findings and recommended actions. In some instances, the appropriate Generic Issue Review Panel may recommend that the screening and assessment stages for reduction in unnecessary regulatory burden issues be modified, or performed at a lower level of effort. The panel documents its recommendation in its initial screening memorandum. As a minimum, the initial screening memorandum is to include a clear, concise description of the GI, its safety significance, and Appendix A of MD 6.4, information prepared by the submitter.

3. Technical Assessment

In the technical assessment stage, staff (a) perform additional review of those GIs that may represent an adequate protection issue, a substantial safety enhancement issue, or a reduction in unnecessary regulatory burden issue; (b) determine if these should be designated as unresolved safety issues (USIs); and (c) identify a cost-effective solution to the GI.

Technical assessment also provides technical justification for excluding from further analyses a GI that has little safety significance, would not result in a substantial safety enhancement, is not cost justifiable, or is a necessary regulatory burden.

Guidance for performing a technical assessment of a reactor GI is provided in Appendix B, "Unresolved Safety Issue Assessment Criteria," and Appendix C of MD 6.4. Guidance for performing a technical assessment of a materials GI would use more qualitative methods, expert elicitation, and judgment as outlined in Appendix F MD 6.4.

Technical assessment is an "indepth" study of a GI and may involve contractor support. To form a technical basis for taking or not taking regulatory action, the technical assessment stage may include the following:

- expert elicitation

- a review of operational data and events

- a review of related generic communications and GIs

- model development, experiments, and tests

- system and computational analyses

- field studies and inspections

- probabilistic risk assessments

- integrated safety assessments

- a detailed regulatory analysis

- corrective action development, including recommendations

The extent of these activities varies in accordance with the scope, complexity, or significance of the GI and the depth of information available on a given GI.

Either RES (for reactor issues) or NMSS (for materials issues), as appropriate, conducts or oversees the technical evaluation of the GI, verifies the legitimacy of the concern expressed, verifies that the benefits sought will be obtained, establishes the technical basis for new or revised regulations or guidance, and identifies solutions that are likely to result in substantial net facility safety improvements or reduction in regulatory burden without significant decrease in safety.

Technical assessment is complete when the RPM informs either the Director of NRR (for reactor issues), or the Director of NMSS (for materials issues) whether (1) the issue should be excluded from further consideration, (2) new or revised rules or guidance are needed, and/or (3) new or revised NRC programs are needed.

4. Regulation and Guidance Development

 Regulation and guidance development involves an in-depth review of potential facility or program changes to address the GI and selection of needed regulatory actions. Technical findings obtained during the technical assessment stage are used, as necessary, as a basis for developing or revising rules, guidance, and programs. Products to be produced during the regulation and guidance development stage could include draft rules, regulatory guides, bulletins, generic letters, information notices, new or revised inspection procedures, and the CRGR briefing packages.

 Typically, NRC rules and guidance are contained in Title 10 of the Code of Federal Regulations, standard review plans, regulatory guides, and, to some extent, bulletins, generic letters, information notices, and regulatory issue summaries, as well as pertinent office staff guidance.

The development of rules, guidance, or programs can take from several months to a few years, depending on the length of time required by the deliberations involved. If rulemaking is a potential option to address the GI, coordination between this directive and MD 6.3, "The Rulemaking Process,"¹⁹²⁸ is required. The GI TAP, in accordance with this directive, the rulemaking plan, and  MD 6.3, is coordinated to reduce duplication of effort.

 Regulation and guidance development is complete when the RPM informs either the Director of NRR (for reactor issues) or the Director of NMSS (for materials issues), whether (1) the GI should be excluded from further consideration or (2) new or revised regulations, guidance, or programs have been developed to address the GI.

 5. Regulation and Guidance Issuance

 The staff issue documents clearly describing the facility or program changes developed during the regulation and guidance development stage to address the GI in a timely and effective manner. New or revised regulations may require the review and approval of the Commission except in limited circumstances when the EDO is authorized to conduct rulemaking in accordance with MD 6.3.

 Regulation and guidance issuance is complete when the RPM informs either the Director of NRR (for reactor issues), or the Director of NMSS (for materials issues), whether (1) the issue should be excluded from further consideration or (2) new or revised regulations, guidance, or programs have been issued to address the GI.

 Regulation and guidance issuance is complete when the RPM informs either the Director of NRR (for reactor issues) or the Director of NMSS (for materials issues) whether (1) the issue should be excluded from further consideration or (2) new or revised regulations, guidance, or programs have been issued to address the GI.

 6. Implementation

 The objective of the implementation stage is to determine whether the licensee, the certificate holder, or other entity regulated by or subject to the regulatory jurisdiction of NRC has established and is implementing a program to ensure that facility or program changes made to address a GI are effective and in accordance with commitments.

 The implementation stage occurs when the affected licensee, certificate holder, or other entity performs the actions necessary to implement the regulatory action to close the GI. These may include modifications or additions to the following:

• ·       the systems, structures, components, or design of a facility
• ·       the design approval or manufacturing license for a facility
• ·       the technical specifications, procedures, programs, or organization required to design, construct, or operate a facility

 The implementation stage is complete for an affected licensee, certificate holder, or other entity once it has formally informed the appropriate NRC program office that facility or program changes have been implemented. The GIP Manager in RES monitors the implementation of GI facility or program changes by the licensee, the certificate holder, or other entity as reported by the RPM and includes this information in updates to the GIMCS.

 7. Verification

 In the verification stage, the appointed staff determines whether licensees, certificate holders, or other entities have adequately demonstrated the efficacy of facility or program changes in addressing the GI.

 The verification stage involves auditing and inspection of individual licensees and certificate holders to verify that effective actions have been implemented. Depending on the number of affected licensees, certificate holders, or other entities, the risk significance of the GI, and the complexity of the corrective actions, it may not be necessary to perform a 100-percent inspection of facility or program changes made in order to declare a GI closed.

 The verification stage is complete for an affected licensee or certificate holder once the final inspection report has been issued, and the appropriate NRC program office determines that facility or program changes are adequate. The RPM provides documentation giving the basis for declaring the verification stage complete for a specific licensee, certificate holder, or other entity to the GIP Manager in RES for review.

 8. Closure

 Closure can begin when the verification stage is complete for all affected licensees, certificate holders, or other entities once:

 - All final verification inspection reports have been issued. 

 - The appropriate NRC program office has determined that actions have been implemented and are adequate to classify the GI as closed. 

 - The RPM has prepared a memorandum to the Executive Director for Operations, through the GIP Manager in RES and the Director of RES, indicating the basis for declaring the GI closed.

V. GENERIC ISSUES PROGRAM (2007-2012)

SECY-07-0022¹⁸⁸⁸ described improvements to the GIP, which the staff will implement to ensure comprehensive and timely resolution of future GIs. The staff will implement these conceptual GIP improvements through a revision to MD 6.4.

 To improve the GIP, the following elements were identified to be incorporated in MD 6.4:

 1. With the appropriate regulatory office involvement, RES will have overall responsibility for GIP management, including routine tracking and documentation of GIP status as well as periodic reporting to Congress and the Commission.

 2. The appropriate regulatory office will have well-defined roles, responsibilities, and accountability in all stages of GI assessment and resolution.

  3. All offices will be involved with applying the screening criteria to identify issues that are suitable for the GIP. Issues for which the risk or safety significance cannot be adequately determined due to phenomena or other uncertainties, and would require long-term studies and/or experimental research to establish the risk or safety significance will be excluded from the GIP, consistent with current processes specified in MD 6.4.

 4. Issues, particularly high-risk issues, that should be addressed by other NRC programs and processes or industry initiatives, will be appropriately directed to those programs and processes. The role of the GIP will be clarified with the roles of other programs that address the concerns of employees and stakeholders such as the Differing Professional Opinion (DPO) Program and the Allegation Program to ensure that GIP does not serve as an alternative to these programs.

 5. To gain efficiency and effectiveness and improve timely assessment of GIs, the staff will employ enhanced risk-informed techniques, which have already been developed as part of other established programs (e.g., the Accident Sequence Precursor [ASP] Program).

 6. RES will ensure necessary inter-office coordination throughout the process. After the issue is screened in as a formal GI, the GIP will consider participation by nuclear industry stakeholders, when feasible, to identify possible solutions (e.g., a regulatory product or industry initiative).

 7. The GI process will be concluded when the regulatory product is identified. The regulatory office will proceed, under other established programs and processes, to develop and implement the identified regulatory solution, and perform appropriate verification.

VI. GENERIC ISSUES PROGRAM (2012-Current)

 In 2012, the Director of NRR requested the formation of a Tiger Team to identify refinements to the agency's GI Program in order to more efficiently resolve GIs and more effectively communicate progress with key stakeholders. The completion of efforts by the Tiger Team was documented in a letter from the team lead, Sher Bahadur, to the Director of NRR, Eric J. Leeds, dated October 24, 2013.²⁰³⁶ The Tiger Team presented its results to the Deputy Executives for Operations and informed the Commission in SECY-13-0081, "Summary of Activities Related to Generic Issues Program," dated July 30, 2013.²²¹⁹

 The Tiger Team performed an extensive review of the existing processes, using industry standard business process improvement techniques. The team found that the GI Program, following the major program overhaul in 2007, is a fundamentally solid program and performs reasonably well. Before establishing the Tiger Team, ongoing GI Program improvement activities had already self-identified potential enhancements. Nevertheless, the business process improvement approach led to productive engagement with many internal stakeholders, resulting in a comprehensive set of proposed program enhancements, supporting recommendations, and a suggested implementation approach.

 One of the Tiger Team's significant recommendations was to reduce the number of stages to three by combining the Screening and Acceptance stages into just a Screening stage (Stage 1). The Assessment stage will be Stage 2, and the Regulatory Assessment Stage will be renamed as the Regulatory Office Implementation stage (Stage 3). The team also recommended changing the time that an issue is declared as a GI from the end of the Screening stage to the end of the Assessment stage. In January 2015, the staff published a new revision of MD 6.4 incorporating the Tiger Team's recommendations.

 Among the other program enhancements recommended by the Tiger Team, the NRC staff worked with information technology contractors to create an online dashboard display on the NRC public Web site that shows the status of active GIs that are in the Regulatory Office Implementation stage. The dashboard shows the current status of active GI and each GI's activities, milestones, and accomplishments. The dashboard also shows information related to recently closed GIs and proposed GIs that were screened out of the process during the Screening stage.

Parent topic: Appendices