ML20205R908

Addresses Commitments Made by NRC in SECY-98-138 & Requests Commission Approval of Staff Proposal to Implement Framework for Using Risk Assessment in Regulating Nuclear Matl Uses & Radwaste Disposal

ML20205R908
Person / Time
Issue date:	03/31/1999
From:	Travers W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-98-138-C, SECY-99-100, SECY-99-100-01, SECY-99-100-1, SECY-99-100-R, NUDOCS 9904230372
Download: ML20205R908 (28)

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March 31.1999 SECY-99-100 i FOR: The Commissioners FROM: William D. Travers <

Executive Director for Operations

SUBJECT:

FRAMEWORK FOR RISK-INFORMED REGULATION IN THE OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS J

PURPOSE:

To (1) address commitments made by the staff in SECY-98-138 and (2) request Commission approval of: (a) the staff's proposal to implement the framework for using risk assessment ir:

regulating nuclear material uses and radioactive waste disposal; (b) the staff's proposed approach for addressing risk management issues in those areas and, in particular, its development of risk metrics and goals; and (c) the formation of a joint Advisory Committee on Reactor Safeguards (ACRS)/ Advisory Committee on Nuclear Waste (ACNW) subcommittee to provide technical peer review of the staff's future efforts. (//[

SUMMARY

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This paper addresses commitments made by the staff in SECY-98-138. It describes the results i of an effort to scope the development of a framework for applying risk assessment methods to I the regulation of nuclear material uses and waste disposal and makes recommendations to the Commission on how to proceed. it first discusses the risk assessment considerations that were CONTACT: Seth M. Coplan, NMSS/DWM (301) 415-5873 l

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c The Commissioners 2 to comprise the scoping effort (i.e., an association of risk assessment methods with nuclear material uses and the regulatory use of such risk assessment methods). It next describes a proposed framework, steps for implementation of that framework, and reports the staff's conclusions about the value of current "Probabilistic Risk Assessment (PRA) Implementation Plan" activities in consideration of the proposed framework. The paper then discusses risk management issues related to nuclear material use and disposal, including the development and establishment of appropriate risk metrics and goals as part of implementing the framework.

Finally, it discusses stakeholder involvement, technical support, and peer review. It recommends that the Commission approve (1) the staff's proposal to implement the framework and (2) formation of a joint ACRS/ACNW subcommittee to provide technical peer review of the staff's future efforts.

BACKGROUND:

In SECY-95-280, the staff informed the Commissio of its framework for applying PRA in reactor regulation. This framework provides a gen .rrci structure to ensure consistent and appropriate application of PRA methods in regulat' ig nuclear reactors. Since the reactor framework was transmitted in November 1995, the offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES) have made substantial progress toward completing the six-step process that was envisioned to implement it.

In its staff requirements memorandum (SRM) of April 15,1997, about risk-informed (RI) and performance-based (PB) regulation, the Commission included direction to the staff to: (1) review its RI and PB approaches with regard to high-level radioactive waste (HLW) issues and nuclear material uses to assure that the needs of those licensees and areas receive adequate consideration; (2). review the bases for nuclear materials regulation to identify areas that can be made amenable to R1 or PB regulation with minimal additional resources; and (3) develop a i framework for applying PRA to nuclear material uses similar to the reactor framework. SECY-98-138 provided a preliminary response to the SRM. The staff concluded that the first two of these requests could not be addressed fully until a framework had been at least partially developed. The staff further concluded that: (1) the reactor framework was not directly applicable because of differences among the activities regulated by the Office of Nuclear Material Safety and Safeguards (NMSS) and collectively between those activities and reactors; and (2) development of an appropriate framework could be a substantial effort that would need to involve the Agreement States (AS') and other stakeholders. The staff informed the Commission that, given U.S. Nuclear Regulatory Commission's (NRC's) available resources, it would first use a task group (TG) to scope the development of a framework, estimate the requisite resources, and make a recommendation to the Commission ( n how to proceed. For the scoping effort, the staff proposed to: (1) make a preliminary atroc.ation of risk assessment methods with regulated uses of nuclear material; and (2) as appropriste for each regulated use and in coordination with the AS', make a preliminary identification of how the associated risk assessment method could be used in a risk-informed regulatory framework for nuclear materials regulation.

Part of the staff's response to the April 15,1997 SRM was the establishment of a NRC/AS' Working Group, the Nuclear Byproduct Material Risk Review (NBMRR) Group, to identify and document a technical basis for a risk-informed approach to regulation of certain material and to develop plans for a graded approach to regulation of that material using risk information. A

..- e The Commissioners 3 companion paper, SECY-99-062, " Nuclear Byproduct Material Risk Review," describes, in greater detail, one of the risk assessment methodologies that is discussed in this paper. Earlier papers provided detailed descriptions of other risk assessment methodologies that are discussed in this paper and apply principally to applications of nuclear materials. More specifically, SECY-94-228 described performance assessment (PA) and SECY-98-185 described integrated safety analysis (ISA).

DISCUSSION:

The staff has completed the proposed scoping effort using a TG drawn from each of the divisions of NMSS and from the offices of NRR and RES' (the TG members and their organizational affiliations are listed in Attachment 1). This paper discusses the results of the scoping effort and makes recommendations to the Commission on a framework for the use of risk assessment in nuclear materials regulation. These recommendations address issues related both to risk assessment [i.e., the variety of assessment methods that are now in use or could be used (and how these are or might be used) in RI regulation] and risk management (i.e., the establishment of metrics and goals for risk to appropriate individuals or groups). The framework proposed is consistent with tt e Commission's "PRA Policy Statement" and, at a high level, parallels the framework adopted for reactor regulation. The objeci!ves of the materials framework are the same as for the reactor framework: (1) enhance safety by focusing NRC and licensee resources in areas commensurate with their importance to health and safety; (2) provide a framework for using risk information in all regulatory matters; and (3) allow use of risk information to provide flexibility in licensing and operational areas. Although risk management issues were not identified as part of the scoping effort in SECY-98-138, they are discussed in this paper because the staff found that uses of risk assessment to meet these three objectives in regulating nuclear materials could be limited by an important policy gap relative to the reactor  ;

situation. Specifically, the Commission's " Safety Goals for the Operations of Nuclear Power i Plants Policy Statement" established important risk metrics and goals for the reactor program l and in that way provided a risk management foundation for subsequent use of PRA. No similar '

policy statement exists for material uses and disposal and, in consequence, the need to consider development of an analogous foundation is discussed in this paper -

Association of Risk Assessment Methods with Nudear Material Uses Broadly, the activities regulated by NMSS can be categorized in four groups: (1) activities that involve long-term commitment of a site or facility to the presence of nuclear material at a i planned, acceptable level (e.g., HLW disposal); (2) activities that involve use of engineered casks to isolate nuclear material under a variety of normal and off-normal conditions (e.g.,

transportation and storage); (3) activities that involve physical and chemical processing and possession of nuclear material at a large-scale facility (e.g., fuel fabrication); and (4) activities that involve the use of either sealed or unsealed byproduct material in a wide variety of industrial and medical applications. Not surprisingly, these groups correspond closely to the organization of NMSS. Their differences from one another include: the facilities, systems, or i devices employed; potential exposure pathways; potential accident initiators and frequencies; potential consequences; and populations at risk. Systematic analysis of these specific features

'The AS' were asked to participate in the TG effort; however, they decsded that their direct participation in the closely related NBMRR along with being kept informed of TG progress would meet their needs adequately.

The Commissioners 4 is the crux cf any risk assessment that might be applied to an NMSS-regulated activity.

Therefore, diiterent risk assessment methods are more efficient and effective for the activities of each group. Such methods have been developed or adapted from methods used for other similar technologies as the need has arisen. Accordingly, the degree of development of and experience in using these methods differs.

Geologic disposal of radioactive wastes, site cleanup, and mill tailings reclamation constitute group 1. Starting in the mid-1970s, the staff has been a developer of PA methodology for the assessment of risk associated with deep geologic disposal of HLW, land disposal of low-level radioactive wastes (LLW), and residual site contamination after decommissioning. From the beginning of the HLW and LLW programs, it was recognized that risk insights that can be derived using PA are particularly well-suited to address issues that arise from the long-term nature of HLW and LLW disposal. Thuc, NRC's existing regulations for deep geologic disposal of HLW (10 CFR Part 60) and land disposal of LLW (10 CFR Part 61) both anticipate the use of PA methodology to show compliance with long term performance objectives for those facilities.

A similar reliance on PA is an essential feature of site-specific regulations that are now being developed for HLW disposal at Yucca Mountain (10 CFR Part 63). More recently, NRC amended 10 CFR Part 20 to establish criteria for residual contamination at decommissioned sites. The staff is currently developing risk-informed guidance to implement these criteria.

Transportation and storage, particularly of spent fuel, comprise group 2. The staff made early efforts to apply risk assessment methodology for the analysis of transportation risk--most notably, the " Final Environmental Statement on the Transportation of Radioactive Matenal by Air and Other Modes"(NUREG-0170), and " Shipping Container Response to Severe Highway and Railway Accident Conditions"(NUREG/CR-4829, also known as the " Modal Study"). More recently, the staff has applied PRA methodology in deciding to approve the one-time shipment of the Trojan reactor pressure vessel, with intemais, for disposal at the U.S. Ecology site in the State of Washington. Also, the staff has nearly completed its re-validation of NUREG-0170 in light of proposed shipments to a repository (vs. reprocessing). The re-validation effort should be completed by the last quarter of fiscal year (FY) 1999 and includes a computer evaluation of cask response to severe accidents and probabilities, the use of current health effects models, and studies of population distributions about likely shipment routes. The staff intends to use the results of the NUREG-0170 re-validation in its efforts to update the Modal Study for the new generation of dual-purpose cask designs. The Modal Study update will focus on confirming severe accident probabilities and effects and will likely include partial or full-scale package tests. Moreover, the staff intends to encourage more RI decision-making with the U.S.

Department of Transportation and the International Atomic Energy Agency (IAEA). The staff believes that ISA and PRA are both appropriate risk assessment methodologies for transportation.

Dry cask storage is the other major group 2 activity. The staff thinks that ISA or PRA can be an appropriate risk assessment tool for this activity as well. At one point, the staff wanted to apply PRA to dry cask storage systems with staff and contractor resources, but suspended the project when the resources were needed for high-priority licensing and certification efforts, in lieu of this broad project, an initial ISA was developed by the staff for one particular dry cask storage system which, in essence, was a general scoping risk assessment of the vulnerabilities of this cask system. The report is currently under peer review and will be issued later this year.

The Commissioners 5 Although this appears to be an effective early risk assessment, further development of an ISA or PRA will be constrained, based on available staff resources.

Fuel fabrication, uranium enrichment, and mining and milling of source material exemplify group

3. There are several ongoing efforts to develop appropriate risk assessment methods for the processes included in this group. First, the staff and the major fuel cycle licensees have adapted risk assessment technology that was developed for the chemical process industry after the Bhopal accident. This adaptation, ISA, has proven to be an integrated hazard identification and assessment methodology for major fuel cycle facility operations. The staff and these licensees have been working to develop a regulatory approach using ISAs, and substantial progress has been made in that regard. In June 1999, the staff expects to transmit proposed revisions to 10 CFR Part 70 incorporating this approach for Commission approval. These proposed revisions will be the result of extensive staff and industry consideration of how the ISA concept should be applied to fuel cycle facilities, and the staff expects that it will be generally supported by the industry. Second, the staff and the U.S. Department of Energy (DOE) plan to use ISA technology to support design and operation of the pre-closure facilities related to a geologic repository for HLW. Such use is incorporated in the proposed Part 63. The staff is sponsoring a project at the Center for Nuclear Waste Regulatory Analyses (CNWRA) to develop guidance for the review of the pre-closure safety analysis for the repository, based on the ISA methodology. Finally, the CNWRA is starting a project to assess the risks associated with in situ leach extraction of Uranium. Risk insights gained from this project will be used to support risk-informed rulemaking for such facilities.

Industrial radiography, nuclear medicine, and well-logging exemplify group 4. With respect to risk assessment, the situation regarding the wide variety of activities in this group is complex.

In the early 1990s, the staff tested the use of PRA methodology to study the risk associated with a new medical procedure (gamma stereotactic surgery). The results were positive, but the  ;

approach was expensive and had some significant limitations. Although the PRA correctly j predicted human error to be the principal accident initiator, the fault tree / event tree l methodology was an inadequate tool for analyzing such accidents. Recently, the staff started I the NBMRR in partial response to the SRM of April 15,1997. The principal objective of the NBMRR was to develop the basis for a risk-graded approach to regulating the activities in group

4. This involved the development of appropriate risk assessment methods to address these activities. The staff believes that the project has resulted in significant progress in that regard and has provided the Commission with a more detailed description in SECY-99-062," Nuclear Byproduct Material Risk Review," dated March 1,1999.

Definitions of these risk assessment methods and a table that displays their specific association with the activities comprising these four groups are provided as Attachment 2 to this paper.

Use of Risk Assessment Methods in a Risk-informed Reaulatorv Framework The "PRA Policy Statement" provides general guidanca on what regulatory use should be made of risk assessment. Implementation of this general guidance can be accomplished by a variety of approaches involving staff and licensee use of risk insights and risk assessment in regulatory decision-making. In each case, there are two principal considerations: (1) What specific use is the staff expected to make of risk insights and risk assessment in development of regulations and guidance, licensing, inspection, assessment, and enforcement? and (2) What specific use

1 The Commissioners 6 is the licensee expected to make of risk insights and risk assessment in planning and conducting its operations? A number of factors are important to these two considerations.

They relate primarily to what can be gained in terms of safety and reduction of regulatory burden, on the one hand, versus the cost of transition and ultimate implementation, on the other. These factors were discussed in SECY-98-138 and include: hazard and complexity of the activity, degree of human involvement in the activity, technical sophistication of the licensee community, NRC staffing and training issues, AS issues, and others. Consideration of these factors in the context of the full variety of NMSS-regulated activities must involve stakeholders and can be expected to result in a number of specific approaches, each of which would be appropriate for the specific activity. Some of these approaches make or will make qualitative use of risk assessment to supplement traditional approaches (e.g., approach for regulating low-activity, sealed sources); others make or will make quantitative and more sophisticated use of such methods (e.g., the HLW approach). A tabulation of current staff thinking regarding such approaches is provided in Attachment 3. This tabulation was developed only as part of the scoping effort. It is preliminary and is likely to change substantially as the framework proposed below is implemented.

A Proposed Framework j l

During its deliberations about the appropriate scope of a nuclear materials framework, the TG developed a framework that is applicable to the materials area. This framework, described in Attachment 4, is similar to the reactor framework, but adopts a lower level of specificity. Like the reactor framework, it is a high-level structure that leaves the particulars of establishing and implementing specific risk-informed approaches to a series of implementation steps. These steps are also described in Attachment 4. Progress toward completing them would be reported and tracked in the "PRA Implementation Plan." With Commission approval, the staff will begin implementation of the framework described in Attachment 4.

Current PRA Implementation Plan Activities in Consideration of the Framework The SRM of April 15,1997, requested that the staff re-examine its Rl/PB approaches with regard to nuclear material licensees and to HLW issues, to ensure that the needs of those licensees and those areas receive adequate consideration. In Attachment B to Attachment 1 of SECY-98-138, the staff provided its preliminary response by re examining the approaches that are supported by Tasks 4 and 5 of the "PRA implementation Plan." The staff considered this response to be preliminary because it believed that some conclusions and priorities could have changed as work on a framework progressed. The staff now has re-examined these same approaches, given its proposed framework, and sees no reason to change its preliminary analysis and conclusions. Moreover, as is discussed above, the staff now expects to add activities to the "PRA Implementation Plan"if the Commission approves its framework.

Risk Manaaement Risk management for NMSS must achieve the overall regulatory goal--safety in the use and disposal of radioactive material. A fundamental element of risk management is to determine which risks to estimate (what are the risk metrics (i.e. what activity produces the risk, what individual, group, or property receives the risk, what conditions produce the risk)) and to determine what limits are acceptable (i.e., risk goals) for these various risks. In addition risk

The Commissioners 7 management would involve using risk insights to evaluate and manage aspects of the regulatory program in various programmatic areas, such as licensing, inspection, and rule changes.

Risk Manaoement Metrics and Goals for Nuclear Material Uses and Discos!

In developing risk metrics a fundamental aspect is whether the risk arises from normal operations with the attendant low-level exposure of workers and the public or whether the risk arises from upset or accident conditions [this is designated normal exposure and potential exposure by the intemational Commission on Radiation Protection (ICRP)]. For both power reactor operation and material uses, the risk metrics and goals have been established for normal operations by intemational and national standards-setting _organizatiohs (such as ICRP and the National Committee on Radiation Protection and Measurements (NCRP)) and further incorporated into law and regulation. For upset and accident conditions at power reactors, the Commission's " Safety Goals for the Operations of Nuclear Power Plants Policy Statement" establishes two qualitative safety goals that are supported by two quantitative health objectives (OHOs). The OHOs are supported, in tum, by two subsidiary risk goals for core damage I frequency and large early release frequency. Although it is attractive to consider the OHOs and analogues of the subsidiary risk goals for material risk management, this is not feasible, because of differences in the population at risk, the number of uses regulated, the nature and behavior of the systems regulated, and hazards posed by reactors versus material uses.

Therefore, the materials program must develop its own set of safety goals. Furthermore, because of the substantial differences among the various material uses, separate safety goals

• for each activity regulated under each program area must be contemplated. It should be noted that this approach could result in different risk goals (or levels of protection) being applied to different regulated activities; however, any such goals would provide reasonable assurance of adequate public protection.

At a minimum, the risk metrics and goals must address the safety of workers and the general i public for normal operations. It should be noted that, in the materials area, the risk associated with normal operation (especially for workers) tends to be large compared with the risk from accidents (e.g. see Table S-4,10 CFR 51.52) Metrics and goals for normal operations have l been established in terms of radiation dose. They apply for all activities and sources and include:

Population at Risk Risk Metric Risk Limit Regulation Workers Annual dose 0.05 Sv (5 rem) and Part 20 ALARA8 Public - Annual dose 1 mSv (0.1 rem) and Part 20 ALARA

' Under the overarching public risk limit, more restrictive limits have been established or are being considered for specific activities or sources. For example, a 0.25 mSv/ year (25 mrem /yr) 1 2

ALARA is the acronym for "as low as is reasonably achievable."

a

p, The Commissioners 8 and ALARA limit has been established in Part 20 for public doses from residual radioactivity at formerly licensed sites and a 0.25 mSv/ year (25 mrem / year) and ALARA limit has been established in Part 61 for public doses resulting from land disposal of LLW. Similarly,0.01 mSv/ year (1 mrem /yr) is being considered as a limit for public doses as a result of recycling of previously contaminated material.

The challenging part of establishing risk goals for materials uses, as for reactors, will be the upset or accident goals. In developing the reactor safety goals, NRC considered such factors as the population at risk surrounding each reactor, the number of reactors, and probability of severe accidents at those reactors. These same factors are important for establishing materials safety goals. For materials, the population at risk depends on the specific use, is quite variable, and can be quite large. For example, a large portion of the entire U.S.

population is at risk from transportation accidents; a smaller, but still large population group is at risk from medical procedures (primarily from diagnostic procedures). In contrast, the population at risk from the proposed repository at Yucca Mountain is comparatively small. The number of material licensees (-20,000) and number of sources for potential accidents are large compared with the number of power reactors (~75). Because there are so many more regulated sources, even if the accident rates were comparable to reactor accident rates, the numbers of incidents would be much larger. Since some of these accidents could be, and

- indeed have been, fatal, the safety goals for material uses must consider the large numbers involved and the likely adverse public reaction that a number of radiation-induced fatalities would engender, even if the risk were low. On the other hand, few material uses involve as much radioactivity as a power reactor and none has the high temperature and pressure that contribute to the greater hazard of the reactor source term. In sum, the risk associated with power reactors derives primarily from low-probability, high-consequence, events, whereas the risk from material uses and disposal derives primarily from higher probability, low-consequence, events. Because of these differences in the nature of the risks, it is appropriate to use different safety goals and different risk management strategies in the two arenas. Some risk metrics that might be useful for materials uses include: (1) overall risk of individual fatality from a particular material use for the appropriate population at risk (both workers8 and members of the public); (2) frequency of large exposures (e.g., exposures in excess of the dose limit for Abnormal Occurrence reporting-0.25 Sv/yr (25 rem / year)] for a particular material use; (3) the maximum dose possible from a particular material use given reasonably conservative assumptions (i.e., a dose cap); and (4) the probability of a criticality event at a facility using fissile material. It is premature to suggest risk goalm.o correspond to these metrics.

In addition to these substantial technical issues, a number of other factors must be considered in developing specific risk management strategies and the risk-informed regulatory approaches that would incorporate such strategies for materials regulation. First, developing, setting, and implementing radiation protection standards is shared by NRC with other stakeholders, including: (1) the U.S. Environmental Protection Agency (EPA) and other Federal agencies; (2)

State govemments (which, in some specific cases, have statutory authority to set and implement more restrictive standards than those established by EPA); and (3) Independent 8

8t should be noted that the safety goals for reactors do not address worker safety. For many, if not most. material uses, the risk to workers is the principal aspect of the risk and. therefore, much of NMSS' regulatory effort is directed toward worker protection from both accidents and normal exposure. Accordingly, risk metrics and goals for nuclear material uses would address worker protection from accidents.

. . 1 The Commissioners 9 standards-setting entities such as ICRP, NCRP, and IAEA. Therefore, NRC must accommodate these shared functions in developing a risk management approach and assure an appropriate level of communication with, and acceptance by, these stakeholders. Second,-

material licensees have a quite variable level of capability in risk assessment, different levels of resources available for all regulatory matters, and different levels of interest in pursuing a risk-informed approach. Except for a few such licensees (e.g., the DOE Yucca Mountain Project),

material licensees do not have a significant capability in probabilisitic safety assessment

. methods and do not have the attendant sunk costs for their development; this is diMerent from the situation for power reactors. These variations in capabilities, resources, and interest must be factored into any RI regulatory approach selected for a particular area of material use.

Thus, as part of the evaluation of altemative RI regulatory approaches, careful consideration must be given to: (1) the costs, both to the staff and licensees, of implementing a new approach; and (2) the benefits, in terms of risk reduction and/or elimination of unnecessary I regulatory burdens. Consistent with these technical and programmatic considerations, the series of implementation steps that are described in Attachment 4 for the proposed nuclear materials framework include consideration of risk management issues.

The variability in target populations, standard-setting authorities, and existing dose limits (and  !

thus in implied risk metrics and goals) is illustrated by the table in attachment 5.

Sjakeholder involvement. Technical Supoort. and Peer Review -

Implementation of the proposed framework and risk management approach will affect the public, other govemment agencies (at all levels), and licensees. Accordingly, the staff considers that a broad range of input will be needed to effectively expand the use of RI regulatory approaches in the materials area. To assure appropriate consistency across NRC, important aspects of implementation will be coordinated with the PRA Steering Committee.

When changes to an existing regulatory approach are contemplated, the staff plans to minimize the impact on NRC and stakeholder resources by: (1) seeking stakeholder involvement through public workshops, intemet postings, and pilot projects; (2) using technical consultants to supplement its own expertise; and (3) establishing a mechanism for technical peer review. For peer review, the staff believes that a joint ACRS/ACNW subcommittee (with appropriate input from the Advisory Committee on the Medical Use of Isotopes) could best integrate a knowledge of RI approaches in the reactor area and an understanding of materials issues into its reviews.

The staff has discussed this matter with the Executive Director of the ACRS and the ACNW.

He agrees with the staff's recommended approach and with Commission approval would form such a subcommittee.

RESOURCES:

To fully implement the approach described above, the staff in each NMSS program area would conduct most of the resource-intensive activities. These would include: evaluating the risk aspects of a progra..,m!M area; interacting with stakeholders; making the appropriate changes

. to the regulations, staff review phins, and Regulatory Guides; training; and developing or adapting needed tools. This would entail an effort of 5 full-time equivalents (FTEs) per year for 5 years starting in FY 2000. In addition, a small cadre of material risk experts would be needed to facilitate the activities in various areas and assure an appropriate degree of consistency across NMSS and within NRC. This would require an additional 3 FTEs per year starting in FY 4

The Commissioners 10 l

2000. In FY 2000, although no FTFs have been budgeted, $400,000 has been budgeted for i contractor technical assistance which would provide approximately 2 FTEs of contractor support for the effort. If the Commission approves the staff's recommendations, the remaining unbudgeted 6 FTE would be reprogrammed from other, as yet, unidentified NMSS efforts in FY 2000.

Some support would also be needed from OGC and ACRS/ACNW. Estimated resources for these offices are 0.2 FTE and 0.5 FTE per year, respectively. Although the OGC and ,

ACRS/ACNW budgets for FY 2000 did not include resources for this effort, these offices will l

reprogram resources from within their currently available budgets if the Commission approves I I

the staff's recommendations. Resources to fully implement this effort in FY 2001 and beyond will be addressed during the FY 2001 budget formulation process.

l ELECOMMENDATION:

That the Commission approve: (1) the staff's proposal to implement the framework set forth in Attachment 4 for using risk assessment in regulating nuclear material uses and disposal; (2) the staff's proposed approach for addressing risk management issues in those areas and, in particular, its development of risk metrics and goals; and (3) formation of a joint ACRS/ACNW l subcommittee to provide technical peer review of the staff's future efforts. i COORDINATION-i The Office of the General Counsel has reviewed this paper and has no legal objections. The  !

Office of the Chief Financial Officer has reviewed this paper for resource implications and has I no objections. The interoffice senior level PRA Steering Committee was briefed on this paper I and its comments have been appropriately incorporated. I illiam Dhrav#er j Executive Director for Operations Attachments: 1. Task Group Members

2. Definition of Terms

3. Potential Regulatory Use of Risk Assessment Methods

4. A Framework for Applying Risk Assessment to Regulating Nuclear Material Uses and Disposal

5. Gummary of Dose Limits and Target Populations

11

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Commissioners' completed vote sheets / comments should be provided directly to the Office of the Secretary by COB Friday, April 16, 1999.

Commission Staff Office comments, if any,-should be submitted to the Commissioners NLT April 9, 1999, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.

DISTRIBUTION:

Commissioners OGC OCAA OIG OPA OCA ACRS ACNW CIO CFO EDO SECY l

a 4

ATTACHMENT 1 l

i l

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'j QifJY;Q1 V9 F9.F"lV'M E 2 TASK GROUP MEMBERS Office of Nuclear Material Safety and Safeauards:

Stephen Koenick, Division of Fuel Cycle Safety and Safeguards Dennis Serig/ John Telford, Division of Industrial and Medical Nuclear Safety Norman Eisenberg, Division of Waste Management Lawrence Kokajko, Spent Fuel Project Office Seth Coplan (Chair), Division of Waste Management Office of Nuclear Reaulatorv Research:

Nathan Siu, Division of Systems Technology Office of Nuclear Reactor Reaulation:

Gareth Pany, Division of Systems Safety and Analysis

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• e e

ATTACHMENT 2 i

l l

2 DEFINITION OF TERMS Virtually all the safety assessment and compliance analysis tools used in the Office of Nuclear Material Safety and Safeguards (NMSS) are a variety of systematic safety assessment methods. A subset of these systematic safety assessment methods are Probabilistic Safety Assessment (PSA) methods. Because the terminology is not standardized and because each group of users of such methods tends to use terms to stress a particular aspect of the methodology or its application, a variety of terms have been developed and employed in various applications. The following definitions are provided for clarification:

System Analysis - System analysis is a directed process for the orderly and timely acquisition and investigation of specific system information pertinent to a given decision. (Fault Tree Handbook,1981)

Probabilistic Safety Assessment (PSA)- A wide class of probabilistic methods used to assess safety; this includes probablistic risk assessment (PRA), risk assessment, failure mode and effects analysis, and performance assessment (PA).

Risk - The risk triplet is the set, <s,, f,, x,>, in which s, represents the ith scenario (sequence or progression); f,is the associated frequency; and x, is the resulting consequence. (S. Kaplan and B. J. Garrick, "On the Quantitative Definition of Risk')

Risk Assessment (RA) " Risk Assessment refers to the technical assessment of the nature and magnitude of risk." (from:" Risk Analysis: A guide to principles and methods for analyzing health and environmental risks." J.J. Cohrssen and V.T. Covello, CEO,1989)

Probabilistic Risk Assessment "Probabilistic Risk Assessment is an analytical technique for integrating diverse aspects of design and operation in order to assess the risk of a particular nuclear power plant [ facility) and to develop an information base for analyzing plant-specific

[ facility-specific] and generic issues. In achieving these objectives, probabilistic risk assessments serve many purposes." (from PRA P ocedures Guide.1982.) Note, this is a definition of PRA focused on U.S. Nuclear Regulatory Commission (NRC) reactor activities and is used as a term of art, within NRC, to denote analyses of reactor safety, usually with considerable detail regarding the component and system failures that lead to an accident. In '

some cases the plant systems analysis (Level i PRA) is expanded to include an analysis of accident progression and source term (Level ll PRA) and further expanded to include consequence analysis and risk integration (Level lli PRA) A broader community uses PRA to mean a broader variety of analyses devoted to other systems and with a wider range of complexity and detail.

Performance Assessment (PA) - PA, a type of systematic safety analysis, is a method: (1) to estimate the potential health, safety, and environmental effects of creating and using a nuclear waste facility; (2) to characterize these effects in terms of their magnitude and likelihood; (3) to compare the characterization of these effects with acceptability standards; and (4) to present the results of these analyses in a format useful to reguiators, scientists, and the public.

(Adapted from N. A. Eisenberg, et al.. "A proposed validation strategy for the U.S. DOE Office of Civilian Radioactive Waste Management geologic repository program," GEOVAL 1987) PA, as used programmatically in NMSS, includes any quantitative assessment or modeling

3 performed to evaluate a waste facility or part thereof, regardless of the degree to which the analysis is probabilistic.

Total System Performance Assessment (TSPA) " Performance assessment is a method of forecasting how a system or parts of a system designed to contain radioactive waste will behave over time. Its goalis to aid in determining whether the system can meet established performance requirements. A TSPA is the subset of performance assessment analyses in which all of the components of a system are linked into a single analysis." (U.S. Department of Energy (DOE), " Viability Assessment of a Repository at Yucca M,untain, Total System Performance Assessment." 1998) This is clearly a term of art used by DOE to emphasize the complete nature of the analysis. It should be noted that although a TSPA must calculate some measure of total system performance, it may also calculate (most analyses siq calculate) the performance of subsystems or provide intermediate results.

Integrated Safety Analysis (ISA) - An ISA is a systematic analysis to identify plant and extemal hazards and their potential for initiating accident sequences; the potential accident sequences and their likelihood and consecucaces; and the items (i.e., site, structures, systems, equipment, components, and activities of personnel) that are relied on for safety. This methodology, adapted from the chemical process industry, provides for flexibility in the scope and detail of the analysis, depending on the magnitude of the hazards and the nature of the systein. This method has been used in NMSS to address the safety in fuel fabrication facilities and in spent fuel storage facilities.

l 1

6

4 ASSOCIATION OF RISK ASSESSMENT METHODS WITH REGULATED USES OF NUCLEAR MATERIALS Group Description Regulated Activities Risk Assessment Method 1 Activities that involve long-term commitment of .-High level waste (HLW) Performance Assessment a site or facility to the presence of nuclear disposal matenal at a planned, acceptable level -Low level waste (LLW) esposal

-Decommissioning (residual contamination)

Mill tailings reclamation 2 Activibes that involve the use of engmeered -Transportation Probabilistic Risk casks to isolate nuclear material under various -Dry cask storage Assessment or Integrated normal and off-normal condbons Safety Analysis (ISA) 3 Activities that involve chemical and physical -Mining and milling of source ISA processmg of nuclear material at a large-scale material facility -Uranium hexaflouride conversion

-Enrichment

-Fuel fabricabon

-Pre-closure actndt6es related to HLW and LLW dsposal '

-Waste treatment facility (vitnfication) 4 Achvities that involve the use of orther sealed or -trradiators Hazard / Barrier Analysis unsealed byproduct material in industrial and -Radography (Nuclear Byproduct medical applications -Medcal Uses Material Risk Review

-Well Logging assessment methodo!ogy)

-Laboratory Use

-Manufacturing ano Distribution

-Gauges

-measuring Systems

-Waste Disposal (incinerabon.

packaging processing) i l

a 4

0 o

J 1

1 I

ATTACHOENT 3 l

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2 POTENTIAL REGULATORY USE OF RISK ASSESSMENT METHODS Group Activity Regulatory Marufestation of Risk Licensee Use of Risk Staff Useof Risk insights Assessment Assessment 1 HigMovel weele (HLW) Probatnishc does standard coeted by Performance _; PA to develop nok insghts esposal rule (PA) to show comphance in sipport of reemalung and tuth standard developrnere of gudence PA to s.eport independent renew of bonnese's analyes 1 1ow level weste (LLW) Dose standard, for reasonable scenenos, PA to show comphance with PA to develop nok insights

• =Paad codited by nde standard inesportof development of guidance. PA to soport independere renew of licensee's analyes 1 E_ _ ,(remdual Does atendard, for reasonable scenenos. PA to show comphance enth PA to develop nok insegles

-- _' cadded by nde standard Gedance nWI inogportof development pommt empteed analysis in of guNience. PA to soport moetcases. mdspondent rowaw of liconess's analyes t hell taihngs reclamaban Urarsum 6481 Taihngs Raeaban Coreal None None Ace and hW Enwronmental Protecton Agency standards a=aah% gn immutable bass 2 Transoortaban Portormance beesd cntena and guMsarm Apphcars4scenase may Prah=he=he nu and nok iniormed regidatory decessons perform nok amassemort to _ (PRA)or (e.g.. Trolen vessel) domed from nok soport regulatory acsons integrated Safety Analyes anaghts (e.g.,Trelan reactor veneel). (ISA) to develop nok mmghts that wiserpm readah-w and giederce 2 Dry cask storege Performance based crhens and giadence .'..' -e"~ may PEIA or !$A to domlop nok and nak rnlormed regidatory amamans peric+m nok ._ ; to insgres flat underpm derhed from risk inaghts soport regdatory accons. regdemore and gudenes 3 henmg of source metenal Presonp6ve entens and pedance denved None ISA to deve6op nak insigtes from nek inaghis that underpm regdenons and padance 3 nadimg of sowce malenal Preacnptve criteria and psdence denved None ISA to develop nok meights ham nok inaghts that undepm regdebens and gudence 3 UF 6 converson Porkwmance requwements compneed of ISA swnmary to ISA summary forms the reenlogmai conesquenoms, pven the demonstrate compharm bass for regidatory -

khshhood of occwrence wnh performance acewties. l

). regiarements l

3 Ennchmert Portormance requiremores compneed of ISA swnmary to ISA summary forms the resologmal consequences, pven the demonstrate comphance bass for regidesory bhahhood of occurrence weiperformenos acewhos.

reewromeras 3 Fumilabncemor Portormance seguiremeras congneed of ISA summary tc ISA summary forms the I rasolopost consequeras even the demonstrate comphance bass for reguistory  !

hhohhood of ocawrence enth porkwmence acevotes. j requiremeres. j 3 Pre <$oewe acevees for HLW Does atendard for normal opershons and a ISA to show comphance with ISA to ogport mdependere disposal spectnsn of likely scenanos does standards rowowof anyIscensos ]

analyons that may beer i agmtcarely cri posth l raf"'~y performance 3 Pre censure acervity for LLW Dose standard for normal operaton erd None None l, esposal presenptive requremeres kw off nomul l conenons 1

e e

e 3

3 Weste treatment decihty Performance req 6sromeras compneed of ISA sunmary to ISA summary forms the (vitnacehon) reensopcar consequences, even the demonstrou comphence bene for regusatory hkahhood of occurrence. wth performerge activmes.

requrements 4 Sealed Sources Regdetory reawremente ranging from None Ongomg reAnomert to enomsson to specroc heenemo enwne develop nok insipes that underpm regdetsons.

Ikonang and inopschon practices, and guidance 4 Unsemied syproduct Metonal Regulatory requremeras renyng from None Ongong retnement to esemphon to spec *c heenang encene deveiop nok insges that wulerpm tedahq_

Econung and inspectan prachces. and gudence A

f

ATTACHMENT 4 1

2 A FRAMEWORK FOR APPLYING RISK ASSESSMENT TO REGULATING NUCLEAR MATERIAL USES AND DISPOSAL

1. THE REACTOR FRAMEWORK OF SECY-95-280 As described in SECY-95-280, the reactor framework is a general structure to ensure consistent and appropriate application of probabilistic risk assessment (PRA) methods. It has four parts. The first defines regulatory application areas (e.g., graded quality assurance) in which PRA can play a role in the U.S. Nuclear Regulatory Commission's (NRC's) decision-making process. The areas are grouped by the expected sophistication of the PRA required (ranging from PRAs based on generic data to state-of-the-art PRAs using plant-specific data).

The second part entails an evaluation of the deterministic engineering considerations underlying the application area to ensure that the existing deterministic engineering approach is altered only after careful consideration. Factors to be considered include: defense-in-depth, the single-failure criterion, and appropriate codes and standards. The third part of the framework is an evaluation of risk issues in support of the proposed regulatory action.

Elements of this evaluation include: scope and level of detail of the PRA, human and equipment reliability, sensitivity and uncertainty analyses, and assurance of technical quality.

The final part integrates the deterministic and risk considerations to ensure a consistent and scrutable decision-making process and to ensure that the underlying bases for rules, regulations, regulatory guides, and staff review guidance are maintained or modified to the extent supported by the risk and engineering conclusions of parts two and three.

This framework is implemented through a six-step process. The first step is to identify the specific regulatory applications that are amenable to expanded use of PRA information and to identify responsible staff organizations and pilot plants. The second is to conduct pilot programs for selected regulatcry application areas. These projects provide insight into the treatment of issues, the selection of risk metrics, and the development of standards and guidance. The third step of the implementation process is to develop and document the acceptance process and criteria. The fourth step is to make near-term regulatory decisions in .

response to industry requests and initiatives. The fifth is to develop formal PRA standards, working with appropriate professional societies and industry groups. Finally, the sixth step is to make long-term modifications to the regulations, if necessary.

2. RISK ASSESSMENT IN MATERIALS REGULATION-COMPARISON WITH REACTORS '

i SECY-98-138 discussed the following differences between the nuclear materials and reactor programs in terms of developing a framework for using risk-assessment in nuclear materials regulation:

l

1. PRA may be applicable only for a few nuclear material uses; other risk assessment methods may be needed for most such uses;

2. Integrating probabilistic and deterministic considerations is not as important in regulating nuclear material uses as it is in reactor regulation;

3

3. Relating analytical methods to specific applications is much more important for materials applications;

4. A broad range of licensee and regulator circumstances will need to be considered.

3. A FRAMEWORK FOR NUCLEAR MATERIAL USES AND DISPOSAL These differences are addressed by a framework that is quite similar to the reactor framework cf SECY-95-280. It too has four parts. Like the reactor framework, the first part defines regulatory application areas in which risk assessment methods can play a role in NRC's d> cision-making process. The areas are grouped by regulated use (e.g., fuel fabrication) and within each use by regulatory application (e.g., graded quality assurance). The second part entails an evaluation of the current considerations underlying the application area to ensure that the existing approach is altered only after careful consideration. Factors to be considered include: deterministic considerations [ hazard, relative importance of human vs. equipment error, -

defense-in-depth (where applicable), codes and standards);, current risk considerations (e.g.,

use of performance assessment in geologic repository licensing); and institutional considerations (existing statutory requirements, Agreement State issues, and licensee circumstances). The third part of the framework is an evaluation of new risk considerations in support of the proposed regulatory action. Elements of this evaluation include: scope and level of detail of the risk assessment, sensitivity and uncertainty analyses, and assurance of technical quality. The final part integrates the current considerations and new risk considerations to ensure a consistent and scrutable decision-making process and to ensure s that the underlying bases for rules, regulations, regulatory guides, and staff review guidance are maintained or modified to the extent supported by the conclusions of parts two and three.

This framework will be implemented through a five-step process. The first step is to identify the specific regulatory applications that are amenable to expanded use of risk assessment information and to identify responsible staff organizations. This step would be accomplished by identifying a full set of regulatory application areas as defined above and then screening them to establish a set of applications that would be amenable to risk-informed (RI) regulatory approaches. The staff would intend to systematically evaluate all of its regulatory applications in this manner, but extemal considerations would be used to prioritize which would be treated first. For example, the staff is currently working with an RI approach for total system performance of a geologic repository for high-level radioactive waste (HLW) because of extemal considerations regarding the national HLW program. Because of limited resources, the staff is proposing this step-by-step approach, rather than a comprehensive reevaluation in i all areas simultaneously. On this prioritized basis, the technical and programmatic factors I affecting the choice of risk metrics and goals in each regulatory application area would be systematically evaluated. Consideration would be given to: (1) the costs, both to the staff and licensees, of implementing a new approach; and (2) the benefits, in terms of risk reduction and/or elimination of unnecessary regulatory burdens. This evaluation would use predictive or actuarial risk studies, as appropriate. Given these considerations, the staff would decide whether it seems appropriate to change the existing regulatory framework and, if so, propose risk metrics and goals as a basis for interaction with stakeholders. Such interaction would I include stakeholder workshops, Intemet postings, and possibly pilot projects.

4 The second step is to decide how to modify the current approach of the regulatory application areas that are determined to be amenable to RI approaches. Stakeholder workshops, Internet postings, and pilot projects will be used as an important source of information to address the following considerations: (1) what specific use ir the staff expected to make of risk insights and risk assessment in development of regulations and guidance, licensing, inspection, assessment, and enforcement? and (2) what specific use is the licensee expected to make of risk insights and risk assessment in planning and conducting its operations? The third step is to make the appropriate changes to the rules and regulations, staff review plans, and Regulatory Guides. Where feasible, the staff would encourage industry development of voluntary standards. The fourth step is staff training to assure consistent and knowledgeable j implementation of the new RI approaches, and the fifth step is to develop or adapt needed tools (e.g., risk assessment methods or computer codes). This five-step implementation process is shown in Figure 1.

1

{

i i

l Step 1 identify: (1) candidate regulatory applications for risk-informed approaches; and (2) responsible organizations.

Y Step 2 Decide how to modify current regulatory approaches.

Y Step 3 Change regulatory approaches.

l l

V Step 4 Implement risk-informed approaches.

Y Step 5 Develop or adapt risk-informed tools.

Figure 1. Five-step implementation process.

9 4

4 ATTACHMENT 5 i

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SUMMARY

OF DOSE LIMITS & TARGET POPULATIONS DOSE OR REOULATORY TARGET PopuurloN REFERENCE DOSE LtMrr BASIS 500 mEM ICRP & NCRP MAX EQUlVALENT DOSE TO THE SKIN OF AN OCCUPATIONAL NCRP#116 RECOMMENDATION WORKERS FOR EMERGENCY UFE*SAVING EFFORTS (P. 44) 400 REM NCRP CAREER MALE ASTRONAUT WHOLE BODY NCRP #98 (P. 7)

RECOMMENDATION DOSE EQUlVALENT UMIT' 300 REM NCRP CAREER FEMALE ASTRONAUT WHOLE SODY NCRP #98 (p. 7)

RECOMMENDATION DOSE EQUfVALENT UMIT' 300MEM 10 CFR I00 MAX TOTAL MADLATION DOSE FOR A 2 HOUR PERICO TO THE I O CFR 100. I I (A)( 1 )

THYRotD FROM A POSTULATED FISSION PRODUCT RELEASE IF AN INDIVIDUAL WERE PRESENT AT ANY POINT OF A NUCLEAR REACTOR *S ExCLuStoN AREA SOUNDARY 250 MEM NCRP THEORETICAL OCCUPATIONAL LIFETIME DOSE NCRP #98 (P. 7)

- RECOMMENDATION i 00 REM NCRP WHoLE soDY DOSE FOR UFE-SAvlNO ACTIONS NCRP #39' (P. I00)

RECOMMENDATION (VAUD UNTIL l986) 50 REM ICRP & NCRP MAX EFFECTIVE DOSE To AN OCCUPATIONAL WORKER NCRP # 1 I 6 RECOMMENDATION FOR EMEROENCY UFE-SAVINO EFFORTS ICRP 60 50 REM 1 O CFR 20 ANNUAL ORGAN OR TISSUE DOSE OTHER THAN LENS OF THE EYE; I O CFR 20. I 201 SMALLOw DOSE EcuivALENT TO THE SKIN OR ANY EXTREMITY

> 25 nEM EPA PROTECTIVE VOLUNTARY WHOLE soDY DOSE Fon uPE-SAVINO ACTIONS & EPA-400-R-9 2-OO I ACTION GUIDES PROTECTION OF LARGE POPULATIONS (MAY f 992) 25mEM IO CFR 100 max TOTAL RADIATION DOSE FOR A 2 HOUR PERIOD TO THE WHOLE i 0 CFR I 00. I I(A)( l) oODY rRow A POSTULATED FiS$ ion PRODUCT RELEASE 1r AN DNDIVIDUAL WERE PRESENT AT ANY PotNT OF A NUCLEAR REACTOR'S EXCLUSION AREA SOUNDARY 25 REM EPA PROTECTIVE WHoLE aODY DOSE FOR LIFE-SAVING ACTIONS & EPA-400-R-92-OO f ACTION GUiOES & PROTECTION OF LARGE POPULATIONS (MAY l992);

USNRC RG 8.29 RG 8.29 (P. I 3) 25 MEM l O CFR 20 & LIFETIME DOSE UMIT FoR INDIVIDUALS PARTICIPATINO IN IO CFR 20. I 206(E)(2) 1 O CFR 835 PLJ.NNED SPECIAL EXPOSURES & I O CFR 835.204

' CAREER wwoLE sODY DOSE toUlVALENT UMIT AT AGE 55 sASED oN A UFETIME EXCESS RISK OF CANCER MORT 5M OF 3x 1 O* PER RAD.

NCRP REPORT No. 39(1978) HAS eEEN SUPERSEDED BY NCRP REPORT No. I 16(1993)

IPARiSK_HLT.wpD CsoNES, NRC 05/15/97 REv 2 ATTACHMENT

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'DosEoR REOULATORY . TARGET POPULATION REFERENCE Dost LIMIT BASIS l 8.75 REM 29 CFR I 9 i O max OUARTERLY DOSE FOR HANDS AND FOREARMS: FEET 29 CFR 19 ) 0.96 (e)

AND ANKLES (OSHA-REOULATED ACTIVITIES)*

I 5 REM 10 CFR 20 ANNUAL EYE DOSE EQUIVALENT (LENS OF THE EYE)  ! O CFR 20 I 20 l iO REM USNRC RG 8.29 ACUTE EMEROENCY EXPOSURE FOR PROTECTING RG 8.29 (1996)

VALUASLE PROPERTY (P. 8. 29-l 3) iO REM NCRP ACUTE EMEROENCY EXPOSURE FOR UFE-SAVING ACTIONS NRCP # 9 i RECOMMENDATION (P. 36) 7.5 REM 29 CFR I 9 I O max OUARTERLY Dose To SKIN OF WHOLE SODY OF 29 CFR 1910.96 (e) 3 OCCUPATIONAL WORKERS (OSHA-REOULATED ACTIVITIES) l 5 REM 10 CFR 20 & ANNUAL EXPOSURE LsMiT FoR OCCUPATIONAL WORKERS I O CFR 20. I 201 &

iO CFR 835 (NRC, DOE & STATES) 10 CFR 835.202 MAX WHOLE SODY DOSE TO ANY INDIVIDUAL LOCATED ON OR j 5 REM i O CFR 72 sEYOND THE NEAREST BOUNDARY OF THE CONTROLLED I O CFR 72.106 AREA OF AN ISFSI OR MRS' 5 REM l O CFR 35 NOTIFICATION UMITS FOR MEDICAL MISADMINISTRATIONS 60 FR 48623 INVOLVINO MEMSERS OF THE PUBUC (OCT 1995) 3 REM 29 CFR I 91 O max OuARTERLY Dost To THE WHoLE soDY 29 CFR 19 I O.96 (OSHA-REOULATED ACTIVITIES)

REMEDIAL ANNUAL ACTION LEVEL FOR NATURALLY NCRP # 1 I 6 2 REM EPA oCCURRlNO RADIATION (RADON) FOR MEMBERS OF THE . (p, 49)

PUsuC (CORRESPONDS TO 2 WLM*)

I .875 PEM OSHA max OLARTERLY HAND OR FOREARM DOSE TO 29 CFR 19 I O.96(s)(3)

A MINOR (UNDER AOE l 8) lAEA THRESwoto roR CoNOUCTINO ENVIRONMENTAL MONITORINO AND lAEA SAFETY SERicS #6 i.5 REM RECOMMENDATION ASSESSMENTS OF RADIATION EXPOSURE LEVELS IN WORK AREAS (l985)

DUE TO THE TRANSPORT OF RAceOACMvE MATERIAL i .25 REM 49 CFR I 72 max OUARTERLY EDE FOR OCCUPATIONAL RAD 6ATION 49 CFR I 72.803 EXPOSURE RESULTINO FROM TRANSPORTATION ACTIVITIES (B)( l )

l REM - Av0 ASTRONAUT Exposure PER FUGHT MISSION NCRP #94 i

OSHA-REoVLATEo ACTivlTIES INCLUDE OCCUPAMONAL EXPOSURE FROM FACILITIES OTHER THAN THOSE REOULATED SY NRC OR AN AOREEMENT STATE - THESE MAY INCLUDE RADIAMON EXPOSURES FROM X-RAYS OR LINEAR ACCELERATORS OPERATED BY NON-AORCEMENT STATES.

• ISFSI = $NDEPENDENT SPENT FUEL STORAGE INSTALMMON. MRS = MoNrTORED RETRiEvAsLE STORAOC INSTALLAMON ONE WORKING LEVEL MONTH (WLM) IS APPROXIMATELY EQUAL TO AN ANNUAL EXPOSURE TO AN AVERAOE OF 4 PCI PER UTER Cr RAooN IF THE RApoN PRooVCTS ARE IN 50% EoutLismiUM wrTH THE RADON. ONE WLM EXPOSURE Wouto RESULT FROM eEsNO EXPO 6ED TO i WORKINO LEVEL (WL) FOR A PERIOD OF l WORKING MONTH (1.E 170 HRS)

EP MISK,HLT.WPO CJONES. NRC OS/l S/97 REv 2 ATTACHMENT

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DOSE oR REGULATORY TARGET POPULATION REFERENCE DOSE L Mrr BASIS I REM EPA EPA PUsuC PROTECTION ACTION OUIDE LIMIT FOR EPA 400-R-92-OO i EVACUATION & SHELTER (PP. 2-6) 750 MREM OSHA MAX OUARTERLY SKIN OF WHOLE BODY DOSE 29 CFR 19 i O.96(s)(3)

TO A MINOR (UNDER AGE l 8)

I 650 MREM -

Avo EDE' PER DIAONOSTIC NUCLEAR URAIN SCAN ' NCRP #93 (P 46) i 540 MREM l O CFR 20' Avo ANNUAL MEASURABLE DOSE PER RADIOORAPHER NUREG-0713 i (l993)* Vot 15 (p. 4-6) f,00 MREM 1 O CFR 35 PROPOSED PATIENT RELEASE CRITERIA SECY-96-100 &

NUREG- 1492 l 500 MREM - i 0 CFR 20, MAX DOSE EQUIVALENT UMIT TO THE EMBRYO / FETUS I O CFR 20. I 208, 1 O CFR 835 & (ENTIRE OESTATION PERIOD) i O CFR 835.206 & l 49 CFR I 72 49 CFR I 72.803 (s)(3) i i

l 500 MREM ANSI, NoN- DESIGN CRITERIA FOR SHIELDINO FoM ANSI N433.1 &

AGREEMENT STATE RADIATION-PRODUCINO MACHINES NCRP #49 REOS (f.E., TELETHERAPY, X-RAY MACHINES, IRRADIATORS)

SCO kREM NCRP MAX ANNUAL EFFECTlYf DOSE UMIT FOR INFREQUENT NCRP # l 16 l RECOMMENDATION ANNUAL EXPOSURES TO MEMBERS OF THE PUSUC (P. 46) i 500 MnEM NCRP REMEDIAL ANNUAL ACTION UMIT RECOMMENDED FOR NCRP #116 RECOMMENDATION CONTINUOUS EXPOSURES FROM NATURAL SOURCES (P. 50)

(EXCLLDING RADON)

{

500 MREM 4 9 CFR I 72 & max ANNUAL RADIATION EXPOSURE TO MEMBERS OF THE 49 CFR I 72.803 EPA FRC OENERAL PUBUC FROM TRANSPORTING RADIOACTIVE (s)(2) l GUIDANCE

• MATERIAL lAEA SAFETY SEmits #6 360 MREM --

ANNUAL TEDE FOR PUSUC NCRP#1OI (INCLUDING ANNUAL MEDICAL EXPOSURE) (P. 73) 300 MREM -

ANNUAL TEDE FOR PUsWC NCRP #94 (EXCLUDING ANNUAL MEDICAL EXPOSURE) 6 EDE = EFFECTivs DOSE coulVALENT RESULTANT AVERAOC DOSE FROM THE APPLICATION OF REOULATORY REQUIREMENTS IN 10 CFR PART 20 0.E., ALARA)

NUMBER OF RAccoORAPMERS MONITORED FOR RADIATION EXPOSURE IN l 993 WAS 4720.

' EPA'S FEDERAL RAciAttoN COUNCIL (FRC) OUIDANCE WAS ISSUED IN 1960. EPA 18 CURRENTLY DEVELOPING OUIDANCE FOR REOUL.ATORY AGENCl!S FOR UMf71NO RADIATION EXPOSURES TO MEMBERS OF THE OENERAL PUBUC, AND THE ANTICIPATED ANNUAL UMf7 IS EXPECTED To SE I OO MREM /yR. HoWEVER, AS OF l 996. THIS NEW EPA OUiDANCE DOCUMENT HAS NOT SEEN

-ISSUED.

IP:WSM,MLT.WPD CJONES, NRC 05/15/97 REV 2 ATTACHMENT

4-4 5

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t DOSE OR REOULATORY TARGET POPULATION REFERENCE Dose LIMIT BASIS 270 MREM - I O CFR 20' AVO ANNUAL MEASURABLE OCCUPATIONAL DOSE PER I O CFR 20. I 201 l

WORKER AT LWRS (1993)

200 MREM -

AVO ANNUAL DOSE TO MEMBERS OF THE PUSUC NCRP #93, # 116 FROM RADON (P. 59; 45) 160 MREM OSHA AVO ANNUAL DOSE EQUlVALENT To NCRP #94 AIRPLANE CREW MEMBERS (P. 22) i i 25 MREM OSHA MAX QUARTERLY WHOLE BODY DOSE TO A MINOR 29 CFR l 9 l O.96(S)(3) i (UNDER AGE l 8) 100 MREM l O CFR 20 & max ANNUAL DOSE UMITS FOR MEMSERS OF THE PUBUC 1 O CFR 20. I 301 &

l 10 CFR 835 I O CFR 835.208 l' LOO MREM lAEA B.S.S. max ANNUAL DOSE EQUIVALENT FOR NON-RADIATION lAEA SAFETY SERIES WORKERS (& SHIELDINO DESION SPECIFICATIONS) 1 I 5-1 100 MR/wM 49 CFR I 72 MAX WEEKLY RADIATION EXPOSURE TO MEMBERS OF THE 49 CFR I 72.803 j PUBUC FROM TRANSPORTATION OF RADIOACTIVE MATERIAL (S)(2)

, 85 MREM PROPOSED MAX DOSE

• CAP" TO AN INDIVIDUAL FOR RESTRICTED USE SECY-96-082 &

~40 CFR 196 (EPA'S PROPOSED DECOMMISSIONING STO) PROPOSED 40 CFR I 96.1 I (D)(2) 1 MAX ANNUAL DOSE EQUIVALENT TO THE THYROfD OF ANY 75 MREM I O CFR 72 REAL INDIVIDUAL LOCATED BEYOND THE CONTROLLED AREA I O CFR 72.104 RESULTINO FROM RADIOACTIVE MATERIALS IN EFFLUENTS AND DIRECT RADIATION FROM AN ISFSI oR MRS I

50 MREM I O CFR 20 - ANNUAL TEDE TO MEMSERS OF THE PUSUC RESULTING PART 20 l APP B, TSL 2 FROM THE INHALATION OR IMOESTION OF RADIONUCUDES CONTINUOUSLY FOR A YEAR j 50 MREM 29 CFR 191 O MAX TEDE FROM INHALATION OR INOESTION 29 CFR I 9 i O.96(C)(2)

To A MINOR (UNDER AGE 18) (REFS To 10 CFR 20)

! LICENSEES (1.E., FUEL CYCLE FACIUTIES) SUBJECT TO 1 O CFR 20. I 301(D) &

25 MREM iO CFR 20 EPA'S OENERALLY-APPUCASLE ENVIRONMENTAL RADIATION 40 CFR 190. I O l STANDARDS IN 40 CFR I 90 l

25 MREM i o CFR 40. APP A MAX ANNUAL PUBUC DOSE EQUIVALENT CANNOT EXCEED 25 I O CFR PART 40, MREM WHOLE SoDY, 75 MREM THYRotD, AND 25 MREM TO CRITEmiON 8 ANY OTHER ORGAN AS A RESULT OF EXPOSURE TO PLANNED DISCHAROES OF RADIOACTIVE MATERIALS, RH-220 AND ITS DAUOHTERS EXCEPTED TO ENVIRONMENT.

RESULTANT AvERAoE DOSE sH I 993 FRoM THE APPucATioN or REout.AToRY REouiREMENTS IN 1 O CFR PART 20 0.E.,

ALARA)

" TOTAL NUwSER or COMMERCIAL LWR WORKERS MONITORED FoR RADIATION . l 993 WAs EXPOSURE 169,862. NURE& IN 0713, VoL 15, P.445, f IAEA S.S.S er INTERNATIONAL BASIC SArETY STANDARDS PoR PROTECTION AoAjNST ton 12lNo RADIATION AND FoR THE SAFETY or RADIATION SOURCES, SArETY SERIES No, i 65-1(1994).

(P:\RISM,HLT.wpD CJoNES, NRC 05/15/97 REv 2 ATTACHMENT

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d s

6 DoSEoR REGULATORY TARGET POPULATION REFERENCE Dose LsMIT basts 25 MREM l O CFR 6 i MAX OFrSITE RELEASES To ANY MEMSER OF THE PUSuC 1 O CFR S i .4 i FOR SOTH OPERATIONS AND POST-CLOSURE ARE UMITED TO 25 MREM WHoLE BODY, 75 MREM THYROID, & 25 MREM OTHER ORGAN 25 MREM IO CFR 72 max ANNUAL DOSE EQUIVALENT TO THE WHOLE BODY oR I O CFR 72.104 OTHER ORGAN OF ANY REAL INDIVIDUAL LOCATED SEYOND TH,E CONTROLLED AREA RESULTINO FROM RADIOACTIVE MATERIALS IN EFFLUENTS AND DIRCCT RADIATION FROM AN ISFSI OR MRS 25 MREM 40 CFR 4 90 ANNUAL DOSE EOUlVALENT SHALL NOT EXCEED 25 MREM 40 CFR 190.10 WHOLE SODY 75 MREM THYROID, & 25 MREM OTHER ORGAN AS THE RESULT OF PLANNED DISCHAROES FROM URAN!UM FUEL CYCLE OPERATIONS TO THE EMYlRONMENT.

25 MREM NCRP MAX ANNUAL EXPOSURE TO MEMSERS OF THE PUSUC FROM NCRP #116 RECOMMENDATION A SINOLE SOURCE OR SET OF SOURCES UNDER ONE (P. 47)

CONTROL 20 MREM -

MAX INDIVIDUAL PUBUC EXPOSURE DUE To NCRP #92 TRANSPORTATION OF RADIOACTIVE MATERIAL (P. I65) 20 MMAD I O CFR PART 50 MAX ANNUAL BETA AIR DOSE FROM GASEOUS EFFLUENTS AT I O CFR 50, APP I APPENDIX l ANY LOCATION NEAR OROUND LEVEL FROM EACH LWR FoR SEcmON ll (B. l .)

ANY SNDIVIDUAL OCCUPYING AN UNRESTRICTED AREA 15 MREM PROPOSED ANNUAL EDE FROM ALL EXPOSURE PATHWAYS FROM A 40 CFR I 96. I I 40 CFR I96 DECOMMISSIONING SITE I 5 MREM max ANNUAL ORGAN DOSE OR DOSE COMMITMENT FROM I O CFR 50, APP 1, 10 CFR PART 50 RADIOACTIVE LODINE OR RAM IN PARTICULATE FORM FROM SECTION ll (C.)

APPENDIX l EFFLUENTS RELEASE FROM EACH LWR FOR ANY SMDIVIDUAL OCCUPY 1NO AN UNRESTRICTED AREA IO MREM -

Av0 ANNUAL EFFECTIVE DOSE EQUlVALENT TO INDIVIDUALS NCRP #93 IN THE U.S. rRow CONSUMER PRODUCTS (P, 59) iO MRAD I O CFR PART 50 MAX ANNUAL OAMMA AIR DOSE FROM OASEOUS EFFLUENTS i O CFR 50, APP l

- APPENDIX I AT ANY LOCATION NEAR ORO!JND LEVEL FROM EACH LWR SECTION 11 (A)

FOR ANY INDIVIDUAL OCCUPYING AN UNRESTR#CTED AREA 1O MREM EPA *S CLEAN AsR MAX DOSE UMIT To MEMSERS oF m4E PUSuC rRoM 40 CFR PAR- 6 i ,

ACT RADICACTIVE AIR EFFLUENTS RESULTINO FROM FACIUTIES SUSPARTi REOULATED UNDER THIS SUSPART

- 1 O MREM 10 CFR PART 50 max ANNUAL ORGAN DOSE OR DOSE COMMITMENT FROM I O CFR 50, APP 1 APPENDIX l UQUlO EFFLUENTS FROM EACH LWR FoR ANY BNDIVIDUAL IN SECTION ll (A)

AN UNRESTRICTED AREA (P:\RasK,HLT.wPD CJONEs. NRC 05/4 5/97 REV 2 - ATTACHMENT

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7 DOSE OR REOULATORY "fARGET POPULATeoN REFERENCE DOSE LIMIT BA$i$

Max ANNUAL Dose To ANY INTERNAL ORGAN OR THE TOTAL sODY" 40 CFR I 96.23 (I) 4 MREM PROPOSED CORRESPONDING 70 INDIVIDUAL MCLs sPECiFIED 'N (SEE AL.SO 40 CFR I 96 i O CFR 14 i FOR PROTECTION OF OROUNDWATER AT A REMEDiATED site 40 CFR I 4 I . I 6)

• 3 MREM i O CFR PART 50 MAX ANNUAL TOTAL BODY DOSE OR DOSE COMMTTMENT I O CFR 50, APP I APPENDIX l FROM UQUID EFFLUENTS FROM EACH LWR FOR ANY SECTION ll (A)

INDIVIDUAL IN AN UNRESTRICTED AREA 2 MREM IN I O CFR 20 max DOSE LIMIT TO MEMsERS OF THE PUBUC IN AN I O CFR 20. I 301 ANY ONE HR UNRESTRICTED AREA FROM EXTERNAL SOURCES" (A)(2)

MAX EXTERNAL RADIATION LEVEL FOR PACKAGES IN ANY 2 MR/HR I O CFR 7 i NORMALLY OCCUPtED SPACE (i.E., LOCATION OF DRIVER i O CFR 71.47 (s)(4)

TRANSPORTINO RADIOACTIVE MATERIAL) 2 MR/HR 49 CFR I 72 MAX RADIATION EXPOSURE TO MEMBERS OF THE OENERAL 49 CFR I 72.803 PUsuC FRoM TRANSPORTATION OF RADIOACTIVE MATERIAL (s)(2)

MAX ANNUAL INDivlDUAL DOSE EQUIVALENT PER SOURCE OR IAEA SArETY SERIES 89; I MREM IAEA SAFETY PRACTICE WITHIN THE RANGE OF RISKS TO BE CONSIDERED IAEA-TECDOC-855 &

SERIES " TRIVIAL. " NCRP # I 16 (P. 5)

ALSO CALLED "NEOUOISLE INDIVIDUAL DOSE (NID)"

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l 13 THE 4 MREM /r4 ORoVNDWATER STANDARD la DERIVED FROM THE AVERAOL ANNUAL CONCENTRA90N OF sETA PARTICLE AND PHOTON RADeOACTIVITY FROM HAN-MADE RADIONUCUDEs IN DRINKINO WATER WHICh WOULD PRODUCE AN ANNUAL DQsE EQUlVALENT OF 4 MREM TO THE TOTAL sODY OR ANY INTERNAL ORGAN (sEg 40 CFR 6 4 4,16). NBS HANDsOOK 69 (AUo 1963) is UsED As THE sAsis FoR DEMIVING THEsE QUANTTTIEs, AND EACH VART FROM THE 4 MREM STANDARD (FOR EXAMPLE.. THE MCL FOR SR*

90 =- 0.07 MREM /YR; THE MCL FOR URAN 8UM au Q,7 pqgM/yR) 14 IN THE STATEMENTS OF CONalDERATION FOR THE REVISED I O CFR PART 20 (3gg 56 FR 23374). THE REASON STATED FOR THE INCLus4CN OF THE DOSE RATE UMIT OF 2 MREM IN ANY ONE HOUR WAs THAT THE UMrT "PROVIDEs A MORE READILY MEAsVRAsLE QUANTtTY THAN THE I CO MREM /YR VALUE AND CAN sE MORE EAstLY VERIFIED sY sHORT11[RM MEASUREMENTS."

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8 REFERENCES ANSI N433. I , " SAFE DEsloN AND UsE or SEtr CONTAINED, DRY SOURCE StoRAos GAMMA lRRADMTORs (C4TEooRY D, t 977.

EPA-400-R-92Co l , EPA PMoTECnvE ACTION Guides," 1992.

IAEA SAFErr SEmits No. 6, "REouLATioNs FOR THE sAFs 'RANsPORT or RADIOACTIVE MATERIALS (l 985 EDmON)

Ill 204, 4705).

lAEA SAFETY SEMits NO. 37, " Advisory MATERIAL FoR THE lAEA REouLATioNs FoR THE SAFE TRANsPoMT oF RADIOACTIVE MATERIAL ( l 98 5 EDmoN) [1 A-470. I 3.

IAEA SAFETY SEMiEs i I 5-l, "INTERNAVoNAL Basic SAFETY STANDARDS FoR PROTECTION AoAINsT lONiZlNo RADIATION AND FoR THE SAFETv or RADIATION sources," l 994.

lAEA TECDOC-855, " CLEARANCE LEVELS roR RADioNUCUDEs IN SOUD MATERIALS" (lNTERIM REPORT FoM COMMENT), i996.

IAEA SAFETY SERIES 89, " PRINCIPLES FoR THE ExEMPnoN OF RADIAnow sources AND PRACnCEs FRoM REouLAToRY CONTROL." l988.

ICRP PusUCATioN 60, " ! 990 RECOMMENDATIONS OF THE ICRP." { J l

NATIONAL BUREAU oF STANDARDS (NBS) HANDsOOK No. 69, ' MAxlMUM PERMissisLE BODY BURDENS AND maximum J PERMissisLE CONCENTRAtoNs or RADioNuCuDEs IN AIR oR WATEn FOR OCCUPATIONAL Exposure," 1969.

NCRP REPORT No. 39, " basic RADIATION PRoTECVoM CRITERIA " l 97 I (SUPERSEDED sY NCRP REPORT 9 l).

NCRP REPORT No. 49,

• STRUCTURAL SHIELDINo DEsloN AND EVALUATION roR MEDICAL UsE or X RAYS AND GAMMA Rare oF ENERoiEs up to 1 O MEV." 1976.

NCRP REPORT No. 9 i , " RECOMMENDATIONS ON UMITs FOR Exposure To loNIZINo RAD 4ATioN,* i 9 8 7.*

(SUPERSEDED sY NCRP REPORT No. I i 6).

NCRP REPORT #92, "PusuC RAD:ATiON ExposuME FMoM NUCLEAR power GENERATION IN THE U.S.," l 987.

NCRP REPORT No. 93, "lONIZINo RADIATION Exposure oF THE Popui ATioN or THE U.S.," 1987, NCRP RtLPORT No. 94, " Exposure oF THE POPULATION or THE U S AND CANADA FRoM NATURAL BACKoROUND '

RADIATION," l 98 7. -

NCRP REPORT No 98, *GulDANCE ON RADIATION RECElvtO tN SPACE ACTivmts " l 989. l NCRP REPORT No.101, " Exposure oF THE U.S POPULATION FROM OCCuPADONAL RADtATiON," 1989.

NCRP REPORT No. I 16. "I.JMITATION OF Exposure To loNIZINo RADIADON " 1993.

NUREG-07 l 3 VoL. I 5. " OCCUPATIONAL RAD 4Amow Exposure AT COMMERCIAL NUCLEAR PowtR REACTORS AND OTHER FACILmEs, I 993."

NUREG-1492, *REouLAToRf ANALYSIS oN CRITERIA FoR THE RELEASE or PATIENTS ADMINISTERED RADioACnvE MATEmiAL," 1994.

U.S. NRC REouLAToRY GutDE 8.29, "lNsTRUCTION CONCERMINo Risks rMoM OCCuPATloNAL Exposure," 1996.

I O CFR PART 20 (NRC REoutAtioNs)

Io CFR PART 35 (NRC REoutATioNs) l O CFR PART 50 (NRC REouLATiows) 10 CFR PART 835 (DOE REouLATioNs) 20 CFR PART I 9 i O (OSHA REouLATioNs) 40 CFR PART S I (EPA REouLATioNs) 40 CFR PARY 190 (EPA RzoutAvioNs) 40 CFR PART I 96 (EPA REouLAT1oNs) 43 CFR PART I 72 (DOT REouLATioNs)

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