ML20233A589
| ML20233A589 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 08/10/2020 |
| From: | Lane H Nuclear Energy Institute |
| To: | |
| SECY/RAS | |
| References | |
| 85FR31709, NRC-2020-0055, PRM-50-121 | |
| Download: ML20233A589 (8) | |
Text
Page 1 of 1 PRM-50-121 3 85FR31709 As of: 8/20/20 10:23 AM Received: August 10, 2020 PUBLIC SUBMISSION Status: Pending_Post Tracking No. 1k4-9ib8-d5w6 Comments Due: August 10, 2020 Submission Type: Web Docket: NRC-2020-0055 Voluntary Adoption of Revised Design Basis Accident Dose Criteria Comment On: NRC-2020-0055-0002 Voluntary Adoption of Revised Design Basis Accident Dose Criteria Document: NRC-2020-0055-DRAFT-0005 Comment on FR Doc # 2020-10599 Submitter Information Name: Allison Borst General Comment See attached file(s)
Attachments 08-10-20_NRC_NEI Comments on PRM 50-121 https://www.fdms.gov/fdms/getcontent?objectId=09000064847ee0fa&format=xml&showorig=false 08/20/2020
HILARY LANE Director, Fuel and Radiation Safety 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202-341-7951 hml@nei.org nei.org August 10, 2020 Ms. Annette Vietti-Cook Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Rulemakings and Adjudications Staff Submitted via Regulations.gov
Subject:
Industry Comments on PRM-50-121; Voluntary Adoption of Revised Design Basis Accident Dose Criteria; Docket No. NRC-2020-0055
Dear Ms. Vietti-Cook,
The Nuclear Energy Institute 1 (NEI), on behalf of its members, submits the following comments pertaining to PRM-50-121, Voluntary Adoption of Revised Design Basis Accident Dose Criteria, 2 published in the Federal Register on May 27, 2020 (85 FR 31709). This petition for rulemaking (PRM) requests that the NRC revise its regulations to allow power reactor licensees to voluntarily revise the accident dose acceptance criteria under 10 CFR Part 100, Reactor Site Criteria, as an alternative to the accident dose criteria specified in § 50.67, Accident Source Term. According to the petition, the revised accident dose criteria would be described in a separate voluntary rule, § 50.67(a), specifying a uniform value of 100 milli-Sieverts (10 rem) for offsite locations and the control room design criteria.
While we recognize that the petition is proposed as a change that licensees would adopt voluntarily, we are unaware of any licensees that would pursue this option, even as a voluntary initiative. Thus, an NRC decision to pursue this rulemaking would constitute a significant expenditure of agency resources with no corresponding net gain.
Following our review of the PRM and the issues raised by the petitioner, NEIs recommendation is that this PRM be denied and no changes be made to the existing regulations.
Discussion of the Petition The petition states that the NRC design basis accident dose criteria and the resulting design of accident mitigation systems appear incongruent. According to the petition, Criterion 19, Control Room, to Appendix 1
The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.
2 The petition is found in ADAMS at Accession No. ML20050M894.
Ms. Annette Vietti Cook August 10, 2020 Page 2 A to Part 50, General Design Criteria for Nuclear Power Plants (GDC-19), states that design criterion for the main control room restricts the calculated 30-day accident dose to the annual occupational limit of 5 rem, while 10 CFR Part 100, Reactor Site Criteria, allows for a calculated dose of 25 rem in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The petition states that the offsite dose criterion was derived from the siting practices of the earliest reactors (greater than 50 years ago) and does not reflect current health physics knowledge or modern plant construction. As a result, the petitioner states that the design of accident mitigation systems could be further optimized, and that the control room accident dose criterion has proven challenging to demonstrate, with most plants having very little margin to meet the current regulations.
NEI does not agree with the petitioners conclusions. Specifically, § 50.67, 10 CFR Part 100, Reactor Site Criteria, GDC-19, 10 CFR Part 20, Standards for Protection Against Radiation, and EPAs Protective Action Guidelines (PAGs) all were established for different, yet complementary, purposes, as discussed in the attachment. Given these differences, there is no need for uniform standards. Additionally, it is important to understand that these different requirements work together to establish a defense-in-depth strategy to protect the workers and the public. For example, GDC-19 was established to implement and regulate a uniform standard for the design of the main control room, as opposed to case-by-case design criteria for each nuclear plant. The GDC holistically embeds design features into plant construction to ensure radiation levels post design-basis accident facilitate operator actions in the event of an emergency. Likewise, 10 CFR Part 100 determines only whether the site is appropriate for a proposed reactor design (i.e., if a licensee cannot comply with 10 CFR Part 100 criteria, the licensee would not be authorized to build at that location and would be required to find a new site).
Finally, a large part of the petitioners rationale revolves around the perception that NRC emphasizes protection of the control room over protection of the public. No evidence, however, is provided that the public believes this to be the case. Indeed, the criteria NRC has established for control room operators likely enhances the perception that protection of the public is of primary concern. Also, the 10 CFR Part 100 regulations appear to explicitly address this concern by stating that the criteria are not intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions.
Industry stakeholders recognize that plant siting and design criteria are not the sole criteria applicable when a plant responds to off-normal operating conditions in an event. Furthermore, it is recognized that 10 CFR Part 20 dose limits are not directly applicable in an emergency, and that industry uses 10 CFR Part 20 in conjunction with the EPAs Protective Action Guides, in responding to a significant plant event.
Additionally, we feel that licensees who contemplate the proposed voluntary initiative will ultimately be deterred in adopting it. Changing nuclear power plant licensing basis regulations would place additional burdens on licensees (revising licensing basis documents, procedures, and training programs, etc.) with no commensurate improvement in safety.
Ms. Annette Vietti Cook August 10, 2020 Page 3 We can assure you that safe and secure operations of commercial nuclear facilities and protection of the workers, public, and the environment are the industrys highest priorities. We would be open to a public meeting at your convenience to discuss these matters. If there are any further questions or comments, please do not hesitate to contact Marty Phalen of my staff, mjp@nei.org.
Sincerely, Hilary Lane cc: Kevin Hsueh, NRR/NRC
Ms. Annette Vietti Cook August 10, 2020 Page 4 ATTACHMENT 1 Title 10 CFR Part 100, Reactor Site Criteria, Criterion 19, Control Room, to Appendix A to Part 50, General Design Criteria for Nuclear Power Plants, (GDC-19), 10 CFR Part 20, Standards for Protection Against Radiation, and the EPAs Protective Action Guides (PAGs) all were established for different purposes. Given these different purposes, there is not a need for uniform dose standards. Specifically, each regulation serves a complementary purpose that, in combination, protects the plant operating staff and the general public.
10 CFR Part 20, Standards for Protection Against Radiation, and § 50.47, Emergency Plans The PRM states, in part, that Although the scope of Part 20 does not specifically address radiation protection standards during emergency conditions, it doesnt specifically exclude emergency conditions either.
Contrary to the above, § 20.1001(b) states, in part, that nothing in this part shall be construed as limiting actions that may be necessary to protect health and safety.
Additionally, NUREG-1736, Consolidated Guidance: 10 CFR Part 20- Standards for Protection Against Radiation, consolidates various guidance documents into a single comprehensive source; and explicitly states, in part, in section 3.20.1001 (Purpose):
This section also makes clear that Part 20 regulations do not apply to licensee activities performed in order to mitigate potential health and safety consequences from accidents or from other incidents involving radioactive material. They also do not apply to emergency actions taken by personnel engaged in NRC-licensed activity Nothing in Part 20 should be interpreted as limiting any activity or action taken to protect public health and safety, such as lifesaving or maintaining confinement of radioactive materials. However, efforts should be made to adhere to these requirements during responses to emergencies, because the requirements were designed to protect the health and safety of workers and the general public.
In responding to plant emergencies, Part 50 licensees implement their NRC approved, station-specific emergency plans in accordance with emergency planning standards specified in §50.47(b)(11): Means for controlling radiological exposures, in an emergency, are established for emergency workers. Specifically, emergency radiation exposures are controlled in accordance with the licensee emergency plans under the direction of the stations Emergency Director and Emergency Radiation Safety Officer. For emergency radiation exposures greater than those of 10 CFR Part 20, licensees implement the guidance of EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. The purpose of this manual is to assist officials in establishing emergency response plans and in making decisions during a nuclear incident. It provides radiological protection guidance that may be used for
Ms. Annette Vietti Cook August 10, 2020 Page 5 responding to any type of commercial nuclear incident or radiological emergency. EPA 400-R-92-001 states, in part, that, Doses to all workers during emergencies should, to the extent practicable, be limited to 5 rem. There are some emergency situations, however, for which higher exposure limits may be justified.
Justification of any such exposure must include the presence of conditions that prevent the rotation of workers or other commonly-used dose reduction methodsthe dose resulting from such emergency exposure should be limited to 10 rem for protecting valuable property, and to 25 rem for life saving activities and the protection of large populations.
In the context of this guidance, worker exposure that is incurred for the protection of large populations may be considered justified for situations in which the collective dose avoided by the emergency operation is significantly larger than that incurred by the workers involved. EPA 400-R-92-001 also states, in part, that doses in excess of 25 rem are permitted for lifesaving or protection of large populations only on a voluntary basis to persons fully aware of the risks involved.
Additionally, emergency radiation exposure controls are distinct from the requirements of 10 CFR 20.1206, Planned Special Exposures. Planned special exposures (PSEs) are not restricted to Part 50 licensees.
10 CFR Part 20 permits any licensee to authorize a planned special exposure (PSE) only in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable or impractical and are controlled separate and distinct from emergency radiation exposures.
However, §20.1201, Occupational Dose Limits For Adults, states that Doses received in excess of the annual limits, including doses received during accidents, emergencies, and planned special exposures, must be subtracted from the limits for planned special exposures that the individual may receive during the current year [annual §20.1201 limits].
Additionally, the petitioner states that Information Notice (IN) No. 84-40, Emergency Worker Doses, informs licensees of their obligation to include doses received during emergency conditions in determining compliance with the occupational dose limits. IN 84-40 specifically states that the Guidance from the NRC staff is provided to clarify this issue and to inform potential emergency preparedness volunteer workers of possible post emergency work restraints subsequent to emergency response activities. This IN does not provide guidance to implement dose controls for emergency workers at Part 50 licensees. Rather, the IN states that all occupational doses, including emergency doses, are required to be included as part of a worker's exposure history.
Appendix A to Part 50, General Design Criteria for Nuclear Power Plants Under the provisions of §50.34, Contents of Applications; Technical Information, an application for a construction permit must include the principal design criteria for a proposed facility. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that
Ms. Annette Vietti Cook August 10, 2020 Page 6 provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
The General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.
GDC-19 specifies that, Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body [total effective dose equivalent (TEDE)], or its equivalent to any part of the body, for the duration of the accident. The General Design Criteria in Appendix A were developed and issued to establish minimum necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Early versions of the control room criteria specified that adequate radiation protection shall be provided to permit access, even under design-basis accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10 CFR Part 20 limits. Although, the explicit basis for the selection of 5 rem whole body dose limit for GDC-19 is not described in the statements of consideration (SOC) for the 1971 rule (which published the GDCs), the SOCs addressed the criteria in the aggregate.
It is generally understood that the objective of the GDC was, within the bounds of the design-basis accidents, to allow for workers to have routine access and occupancy of the main control room, the remote shutdown panels, and the emergency plans technical support center (TSC) within the dose limits of 10 CFR Part 20. These GDCs are facility design requirements and are separate and discrete from maintaining and controlling workers emergency radiation exposures. Specifically, GDC-19 was written to ensure that the design of the main control room maintained habitability for plant staff within the limits of 10 CFR Part 20, for the duration of the design-basis accidents.
10 CFR Part 100, Reactor Site Criteria The issues identified in the PRM concerns the dose limits for 10 CFR Part 100, Reactor Site Criteria.
Specifically, these dose limits for an exclusion area are based on radiological exposures, two hours immediately following onset of the postulated fission product release (i.e., not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure). A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
Ms. Annette Vietti Cook August 10, 2020 Page 7 As stated, these are the reactor siting criteria and not the requirements to necessarily manage public or worker radiation exposures. Licensee response to fission product releases are defined in and are in accordance with emergency planning standards §50.47, Emergency Plans, which states that no initial operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The reactor siting process, by design, limits the magnitude of the dose consequences of a design-based accident fission product release. However, the reactor siting criteria is but a single (not sole) barrier to provide reasonable assurance of public health and safety under accident conditions.