ML21270A005
| ML21270A005 | |
| Person / Time | |
|---|---|
| Issue date: | 10/15/2021 |
| From: | Robert Beall NRC/NMSS/DREFS/RRPB |
| To: | |
| Beall, Robert | |
| Shared Package | |
| ML20289A534 | List:
|
| References | |
| 10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML21270A005 (10) | |
Text
THIS PRELIMINARY PROPOSED RULE LANGUAGE AND ACCOMPANYING DISCUSSION IS BEING RELEASED TO SUPPORT INTERACTIONS WITH STAKEHOLDERS AND THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS). THIS LANGUAGE HAS NOT BEEN SUBJECT TO COMPLETE NRC MANAGEMENT OR LEGAL REVIEW, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS.
THE NRC STAFF PLANS TO CONTINUE WORKING ON THE CONCEPTS AND DETAILS PROVIDED IN THIS DOCUMENT AND WILL CONTINUE TO PROVIDE OPPORTUNITIES FOR PUBLIC PARTICIPATION AS PART OF THE RULEMAKING ACTIVITIES.
This first iteration of key elements of an alternative design/licensing approach supporting more traditional methodologies (e.g.,
Deterministic selection of postulated initiating events, inclusion of single failure criterion) has been prepared below in the form of sections to be added to 10 CFR Part 50 for convenience. The final location of this alternative design/licensing approach will be determined at a later date. This new section would also apply as alternative technical requirements for applicants using the licensing processes in 10 CFR Part 52. The staff has found that this is a more efficient placement for the time being because Part 50 reflects the traditional methodologies such as those used in some international standards. This preliminary proposed rule language would provide a technology-inclusive alternative to the technical requirements specifically developed for light-water reactors (LWRs). However, the staff continues to evaluate where to place this alternative in relation to the more PRA--centered methodology in the preliminary proposed Part 53 subparts and the LWR--centered technical requirements in Part 50.
THE STAFF IS PRIMARILY SEEKING INSIGHTS REGARDING THE CONCEPTS IN THIS PRELIMINARY LANGUAGE AND SECONDARILY SEEKING INSIGHTS RELATED TO DETAILS SUCH AS NUMERICAL VALUES FOR VARIOUS CRITERIA OR FORMATTING ISSUES SUCH AS THE LOCATION OF THE REQUIREMENTS WITHIN TITLE 10 OF THE CODE OF FEDERAL REGULATIONS.
STAFF DISCUSSION OF TECHNOLOGY-INCLUSIVE ALTERNATIVE REQUIREMENTS FOR COMMERCIAL NUCLEAR PLANTS - PRELIMINARY PROPOSED RULE LANGUAGE (October 2021)
Preliminary Language Discussion
§ 50.200 Technology-inclusive alternative requirements for The following preliminary proposed rule language has commercial nuclear plants been prepared for 10 CFR Part 50 to support the public release of this document. The staff would like to receive feedback on whether Part 50 is the appropriate location or if this proposed rule language should be incorporated into Part 53, and if so, how.
§ 50.210 Applicability Consistent with the currently issued preliminary Applicants submitting an application after [ENTER EFFECTIVE proposed Part 53 language, this approach could be DATE OF FINAL RULE] for a commercial nuclear plant under Part 50 used by any reactor applicant. Uses the Part 50/52 or Part 52 of this chapter may elect to adopt the following technology- regulatory framework as a baseline. The applicability 1
inclusive requirements as an alternative to technical requirements in for some of the alternatives may be limited to non-specified sections of Parts 50 and 52. LWRs to provide a technology-inclusive complement to existing LWR requirements.
§ 50.220 Definitions For the purpose of §§ 50.210 to 50.290:
Anticipated operational occurrences mean those conditions of Currently only defined in 10 CFR Part 50 Appendix A.
normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.
Commercial nuclear plant means a utilization facility consisting of one or more nuclear reactors and associated co-located support facilities, which may include one or more reactor modules, [using nuclear fission, nuclear fusion, or accelerator-driven reactor technologies] that are used for producing power for commercial electric or other commercial purposes. The commercial nuclear plant includes the collection of sites, buildings, radionuclide sources, and structures, systems, and components (SSCs) for which a license is being sought after [ENTER THE EFFECTIVE DATE OF THE FINAL RULE].
Non-light-water reactor means a reactor that does not use water that does not contain deuterium as its coolant and neutron moderator.
Reactor coolant pressure boundary has the definition specified in New definition replaces the reactor coolant pressure
§ 50.2. Where reactor coolant pressure boundary is used in the boundary portion of safety-related (below).
definition of basic component in § 50.2, any plant structure, system, component, or part thereof that is relied on to perform other safety related functions identified in §§ 50.250 and 50.280, such as cooling to maintain the integrity of required systems and barriers, should also be classified as a basic component in the context of § 50.55(e), as applicable.
Safety related SSCs has the definition specified in § 50.2 for light The definition of Safety related is still under water reactors. For non-light water reactors, an acceptable alternative development, both in terms of the substance of the is: those safety-related SSCs that are relied on in design basis events, definition and the use of this term for SSC safety as defined in § 50.49, to assure: categorization in these sections of preliminary
- 1) the capability to perform other safety functions as identified in proposed language.
§§ 50.250 and 50.280, such as cooling to maintain the integrity of required systems and barriers; 2
- 2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or
- 3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 52.79 of this chapter, as applicable.
All other terms in §§ 50.200 et seq. have the meaning set out in 10 CFR 50.2 and 10 CFR 52.1 or Section 11 of the Atomic Energy Act, as applicable.
§ 50.230 Requirements These requirements are overarching elements that Applicants must meet the following requirements: should be addressed by all applicants. These do not (a) Single failure criterion. constitute new requirements; rather, they are identified (1) In using the provisions in §§ 50.210 through 50.290, the separately here due to conflicts with existing language applicant must evaluate events assuming the worst single failure of (single failure as part of the GDC) or for emphasis active safety-related SSCs as part of the analysis and evaluations (use of PRA in a supporting, instead of leading, role).
required to demonstrate the design adequately mitigates the consequences of AOOs and DBAs.
(2) If an SSC is designated as passive, the designer must perform a comprehensive analysis and evaluation of test and/or operational data for the SSC to demonstrate that the reliability of the SSC is such that its failure probability is sufficiently low to justify not It is expected that any plant under this section will applying the single failure criterion [in (a)(1)]. reflect through its design, construction, and operation (b) Probabilistic Risk Assessment. Applicants using the an extremely low probability for accidents that could provisions of §§ 50.210 through 50.290 are required to develop a result in the release of significant quantities of probabilistic risk assessment (PRA) and provide a description of the radioactive fission products.
PRA and its results in their applications.
(c) Defense-in-depth. Applicants need to demonstrate adequate Defense-in-depth is called out explicitly here as it is defense -in- depth is provided in the design to prevent and mitigate referenced in the analytical requirements below. The AOOs and DBAs and to address beyond design basis events, including approach taken is consistent with Commission policy, those potentially resulting in severe plant conditions. and more information can be found in NUREG/KM-0009. Note that defense in depth may be addressed, at least in part, in existing requirements such as the general design criteria for LWRs (and available guidance for principal design criteria for non-LWRs).
§ 50.240 Principal design criteria This section more directly addresses PDC and their role. Use of a deterministic approach is likely to rely 3
(a) In lieu of § 50.34(a)(3)(i) for construction permits and more on top level design goals in the form of design operating licenses, § 52.47(a)(3)(i) for design certifications, criteria as opposed to a more integrated assessment.
§ 52.79(a)(4)(i) for combined licenses, § 52.137(a)(3)(i) for standard This language would allow for the use of the criteria in design approvals, and § 52.157(a) for manufacturing licenses, non-light IAEA SSR 2/1 - the applicable standards include, but water reactor applicants may provide alternatively developed principal are not limited to: the existing GDC, Regulatory Guide design criteria (PDC) for the facility that do not follow the general design (RG) 1.232, and IAEA SSR 2/1.
criteria (GDC) in 10 CFR Part 50, Appendix A. PDC establish the necessary design, fabrication, construction, testing, and performance Because of the existing rule language, the proposed requirements for SSCs important to safety.
text is applicable to non-LWRs only. NRC staff is (b) Non-light water reactor applicants are required to provide considering how to allow LWRs to use already principal design criteria using the GDC or other generally accepted developed alternatives such as IAEA SSR 2/1 as part consensus codes and standards to inform the development of the of this approach.
provided PDC. Sufficient information must be provided to provide reasonable assurance that the final design will conform to the design bases with adequate margin for safety. Paragraph (b) is based on similar wording in § 50.34(a)(3)(iii).
§ 50.250 Anticipated operational occurrences and design basis accidents (a) Applicants are required under § 50.34 for construction permits and operating licenses, § 52.47 for design certifications, § 52.79 for combined licenses, § 52.137 for standard design approvals, and
§ 52.157 for manufacturing licenses to provide an analysis and evaluation of the design and performance of SSCs of the facility.
(b)(1) In lieu of §§ 50.46, 50.34(a)(4), 52.47(a)(4), 52.79(a)(5),
52.137(a)(4), and 52.157(f)(1), applicants using the provisions of §§ These requirements are consistent in concept with 50.210 through 50.290 are required to identify postulated initiating existing regulations and international standards for events for anticipated operational occurrences and design basis these classes of events. Applicants should provide accidents using a generally accepted, risk-informed approach for analysis for AOOs and DBAs, and features used to systematically evaluating engineered systems. mitigate and prevent these events should be safety (2) Those applicants are also required to define acceptance related.
criteria for safety-related SSCs to provide reasonable assurance that their performance during anticipated operational occurrences and design basis accidents adequately mitigates the consequences of such events.
(3) The analyses must demonstrate that there is reasonable assurance that fission products are retained within specified barriers for 4
each analyzed accident or otherwise that the dose to an individual located at the exclusion area boundary or low population zone outer boundary remains below the reference values specified elsewhere in this part. SSCs required to mitigate against anticipated operational occurrences and design basis accidents must be classified as safety-related.
(4) Safety-related SSCs must be designed and located with due consideration to the environments and conditions associated with the internal and external hazards associated with design basis events.
(5) Applicants must provide an analysis and evaluation of the The requirement in (5) is based on 10 CFR design and performance of SSCs with the objective of assessing the risk 50.34(a)(4).
to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.
(6) Applicants may elect to perform a single or multiple bounding The requirements in (6) provide an avenue for an analyses and evaluations to demonstrate the design appropriately applicant to provide bounding analyses for some or all mitigates the consequences of accidents; in taking this approach, of the analytical requirements for this part. To some applicants must demonstrate that the bounding evaluation(s) adequately extent, this is consistent with exiting practice - a single envelope conditions for the full range of anticipated operational analysis to cover a category of event (e.g.,
occurrences and design basis accidents with sufficient margin. Such an overcooling) is often provided as part of a safety evaluation may not be realistic in order to provide reasonable assurance analysis. This would go a step further and allow for that operation of the facility could not exceed the conditions imposed for bounding analyses (potentially involving non-realistic the bounding evaluation(s). assumptions) to be provided to cover larger portions of (c)(1) Applicants must identify limiting parameters that serve as the AOO and DBA analytical space, provided the safety acceptance criteria for the analyses of events and provide these analysis envelopes the full range of conditions it is values as part of the application. stated to bound.
(2) For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects these Further, this section incorporates requirements safety acceptance criteria, the applicant or holder of a construction adapted from § 50.46(a)(3) - applicants are required to permit, operating license, combined license, or manufacturing license identify surrogate safety acceptance criteria, akin to must report the nature of the change or error and its estimated effect on peak cladding temperature for LWRs, and track and the safety analysis to the Commission at least annually as specified in report errors in the analysis for these acceptance
§§ 50.4 or 52.3 of this chapter, as applicable. criteria. For LWRs, staff expects § 50.46 criteria will (3) If the change or error is significant, the applicant or licensee be the ones chosen.
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must provide this report within 30 days of identification and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with other requirements in
§§ 50.210 through 50.290.
§ 50.260 Beyond design basis events This section replaces SBO and ATWS regulations with (a) In lieu of §§ 50.62, 50.63, 52.47(a)(15), 52.47(a)(16), a broader category of events, and draws on the 52.79(a)(9), 52.79(a)(42), 52.137(a)(15), 52.137(a)(16), 52.157(f)(5), international concept of defense-in-depth level 3b or and 52.157(f)(7) of this part, applicants using the provisions of §§ 4a.
50.210 through 50.290 must perform additional assessments and analyses to identify design features or programmatic controls for enhancing the plants capabilities to withstand, without undue risk, events that are either more severe than design basis accidents or that involve additional failures. Events include unlikely but credible events that could lead to situations beyond those considered for DBAs, multiple It requires applicants to evaluate and provide credible failures (e.g., common cause failures in redundant SSCs) that prevention/mitigation features (non-safety related) prevent safety systems from performing their intended function, or against events more severe than DBAs based on credible failure sequences that are not assessed within the scope of operating experience, engineering judgement, and DBAs but are mitigated by other plant SSCs outside the scope of the sequence-based assessment. These SSCs that are credited safety function of those SSCs. credited should have quality treatments in accordance (b) Design features or programmatic controls should be with their function.
developed to establish supplementary protections to mitigate against recognized BDBE initiators (e.g., reduction of risk from anticipated transients without scram, loss of all alternating current power) or complex accident sequences that may have substantial uncertainty associated with them, as well as other conditions specific to the design derived on the basis of engineering judgement, deterministic assessments, and probabilistic assessments. These features provide additional assurance of safety and defense-in-depth.
(c) SSCs required to mitigate beyond design basis events need The bounding analyses that may be used for AOO or not be classified as safety related, but should have appropriate DBA requirements may be expanded for use by treatments identified to ensure these SSCs function as specified in the applicants here.
analyses required in (a) to mitigate these events. If an applicant elects to provide a bounding evaluation as described in § 50.250, that Special treatments include, but are not limited to evaluation may be used to address any or all of the event(s) required as availability controls (e.g., TS) or augmented quality.
part of §§ 50.210 through 50.290 provided the bounding evaluation is demonstrated to envelope these beyond design basis events.
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§ 50.270 Severe accidents (a)(1)(i) In lieu of §§ 50.34(a)(1)(ii)(D), 52.47(a)(2)(iv), These requirements replace existing severe accident 52.47(a)(23), 52.79(a)(1)(vi), 52.79(a)(38), 52.137(a)(2)(iv), requirements. This section borrows from the 52.137(a)(23), and 52.157(d) in this part, applicants using the provisions international concept of defense-in-depth level 4 or 4b.
of §§ 50.210 through 50.290 are required to provide a description and The requirements identified here are consistent with analysis of design features deemed important to safety because they the Commission's severe accident policy statement prevent or mitigate accidents that could progress beyond design basis (50 FR 32138) while tying together existing accidents and events addressed by §§ 50.250 and 50.260. These requirements with the commensurate analysis.
events could include conditions not considered for design basis accidents, but that are considered in the overall design using best estimate methodology including consideration of uncertainties, in order to assess risk to the public health and safety. These events include those that would require analysis of design features for the prevention For LWRs, such accidents have generally been and mitigation of severe accidents. assumed to result in substantial meltdown of the core (ii) A light water reactor applicant must address how the design with subsequent release into the containment of prevents and mitigates severe accidents based on conditions derived appreciable quantities of fission products.
from operating experience and/or input from probabilistic risk assessments. e.g., challenges to containment integrity caused by (iii) An applicant with a non-light-water reactor design must use core-concrete interaction, steam explosion, high-engineering judgement and/or input from probabilistic risk assessments pressure core melt ejection, hydrogen combustion, to identify what constitutes severe accident conditions for their specific and containment bypass.
design and describe the measures provided in the design for preventing or mitigating such accidents. Severe accidents for non-LWRs are not defined to the (iv) Analyses of these accidents must show that the design same degree as LWRs; events evaluated in this demonstrates adequate defense-in-depth such that acceptable dose section should involve some level of fuel or core consequence criteria - including those in § 50.270(a)(2)(iv) below - are damage, based on the event criteria outlined in this met even in circumstances with fuel or core damage or potential for section.
large radiological releases from other sources in the facility.
(2)(i) The applicant must provide information regarding safety features that will be engineered in the facility and any barriers that must be protected during various accidents to limit the release of radioactive (iv) is consistent with the existing requirements related material released to the environment. to the 25 rem and Part 100 requirements, which are (ii) The applicant must perform an analysis and evaluation of the based on a core damage event.
severe accidents that could lead to fission product release, using the expected barrier leak rate(s) and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with 7
applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences.
(iii) The accident-specific fission product release to be used in the analyses required by § 50.270(a)(2)(ii), must consist of a mechanistic source term that is based on physically based models of the facility response.
(iv) Site characteristics must comply with Part 100 of this chapter.
(v) The minimum acceptance criteria for the analysis required in Minimum is a reference to the fact that applicants may this section are: elect to use more stringent acceptance criteria (which (A) An individual located at any point on the boundary of the would then replace these) in order to achieve exclusion area for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the additional flexibilities offered by Part 53 provisions postulated fission product release, would not receive a radiation dose in referenced below.
excess of 25 rem total effective dose equivalent (TEDE), and (B) An individual located at any point on the outer boundary of These criteria also apply to DBAs, consistent with the low population zone, who is exposed to the radioactive cloud existing requirements. NRC staff expects that the resulting from the postulated fission product release (during the entire severe accident conditions will bound those period of its passage) would not receive a radiation dose in excess of 25 conditions.
(vi) Analyses that show that the necessary systems and barriers remain effective during postulated accidents may be used as a surrogate for offsite dose calculations required in § 50.270(a)(2)(ii) for postulated accidents.
(vii) Applicants electing not to use mechanistic source terms to evaluate postulated accidents may use a bounding-type assessment assuming severe plant conditions and reliance on a given barrier such as a containment structure.
(b) As part of the overall safety design philosophy, the applicant Requires applicants consider defense-in-depth (no must demonstrate defense-in-depth such that no plausible scenario reliance on a single SSC/barrier) and mitigate against leads to dose consequences beyond the acceptance criteria identified in more severe potential scenarios. Provides avenues the § 50.270(a)(2)(iv). In performing the analyses required in this for crediting barrier mitigation and excluding some section, applicants are not required to evaluate scenarios that are not events, similar to international practical elimination physically possible or can be shown to occur at a sufficiently low concept. Staff expects there would be a frequency frequency with a high degree of confidence such that consideration of threshold for this exclusion for applicants leveraging a these events can be excluded as part of the residual risk of the facility. PRA. The residual risk portion is subject to change or further clarification.
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§ 50.280 Functional containment (a) As an alternative to § 50.54(o), the requirement that the These requirements replace containment-related containment remains intact in § 50.150(a)(i), the containment portion of regulatory requirements. They establish what
§§ 50.155(b)(1)(i), and 52.79(a)(12), non-LWR applicants may elect to constitutes a functional containment and makes provide a functional containment; that is, may designate a set of barriers functional containment SSC qualification taken together that effectively limit the physical transport and release of commensurate with the purpose of the component radionuclides to the environment across the full spectrum of events (safety related for AOOs/DBAs, special treatment for discussed above. As part of the approach under §§ 50.210 through BDBEs).
50.290, aspects of SSCs designated as part of the functional containment (and those that support these SSCs) used in the analyses of DBAs must be classified as safety related.
(b) If SSCs designated as part of the functional containment are Paragraph (b) reflects that the SSCs making up the relied on to mitigate events in § 50.260, the applicant must identify functional containment should be classified according appropriate treatments such that these SSCs function as assumed in to their role in addressing AOOs, DBAs, beyond-their role as part of the functional containment. To support defense-in- design basis accidents, or severe accidents.
depth, acceptance criteria (including the dose consequence criteria) related to the performance of these SSCs must be met without exclusive reliance on any single element of the design.
§ 50.290 Design requirements Applicants must apply the following provisions of this section, as Depending on how this proposed rule language applicable: develops, this section may be folded into § 50.230 (a) Technical specifications - In lieu of the four criteria for limiting and/or portions of § 50.230 may be relocated here.
conditions for operation (LCO) listed in § 50.36(c)(2)(ii), applicants may provide LCOs for § 50.36(c)(2)(ii)(B) and (C) only, provided these LCO criteria (A) relates to reactor coolant pressure criteria identify appropriate requirements on systems that perform other boundary; LCO criteria (D) is based on PRA and safety functions such as cooling to maintain the integrity of required operating experience. This provision would remove systems and barriers. consideration of those criteria from being required, (b) Provided applicants comply with §§ 50.220 through 50.290, provided barrier requirements are captured. This applicants using the provisions of this section need not comply with the serves to catch additional Part 50 regulations that following regulations: [to be revised as needed] conflict with this section and could change as the Part (c) Reserved 53 provisions are added.
Areas from Part 53 being explored for use in this alternative framework-In utilizing the provisions of §§ 50.200 et seq., applicants may choose to use the following regulations:
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Special treatment - In addressing the requirements associated with paragraph (e) of this section, applicants are required to identify appropriate treatments for SSCs relied on to mitigate these events. In identifying these treatments, applicants may use the framework set forth in § 53.YYY.
Siting considerations - In lieu of [appropriate set of 50/52 siting requirements], applicants may apply
§ 53.5XX to determine site boundary areas and populations considerations.
Emergency preparedness requirements - In lieu of §§ 50.54(q), 50.54(t), [other appropriate 50/52 EP requirements], applicants may apply § 53.5XX to determine emergency preparedness requirements.
Security requirements - As an alternative to the requirements set forth in §§ 50.34(c), 52.79(a)(35), and
[other appropriate requirements as applicable], applicants may apply § 73.YY in lieu of the requirements necessary to satisfy the cited physical security requirements.
(additional references to Part 53 - here or elsewhere) 10