ML21083A031
| ML21083A031 | |
| Person / Time | |
|---|---|
| Issue date: | 03/29/2021 |
| 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: ML21083A031 (17) | |
Text
THIS SECOND ITERATION OF PRELIMINARY RULE LANGUAGE IS BEING RELEASED TO SUPPORT INTERACTIONS WITH STAKEHOLDERS AND THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS). THIS LANGUAGE HAS BEEN SUBJECT TO ONLY LIMITED 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 ITERATION OF PRELIMINARY RULE LANGUAGE AND WILL CONTINUE TO PROVIDE OPPORTUNITIES FOR PUBLIC PARTICIPATION AS PART OF THE PART 53 RULEMAKING ACTIVITIES.
THIS SECOND ITERATION IS PROVIDED IN THE SAME GENERAL STRUCTURE AS THE ORIGINAL RELEASE OF SUBPART B, TECHNOLOGY-INCLUSIVE SAFETY REQUIREMENTS, AND SUBPART C, DESIGN AND ANALYSIS REQUIREMENTS. SOME FEEDBACK HAS SUGGESTED BROADER CHANGES TO THE STRUCTURE AND CONTENT OF PART 53. THE STAFF IS CONTINUING TO REVIEW ALL OF THE COMMENTS AND SUGGESTIONS RECEIVED TO DATE BUT IS ISSUING THIS SECOND ITERATION TO SUPPORT ONGOING DISCUSSIONS RELATED TO KEY CONCEPTS.
Subpart B, Technology-Inclusive Safety Requirements 2nd Iteration (Redline/Strikeout) of Discussion Preliminary Rule Language In sum, Subpart B defines the safety criteria that other areas of Part 53 (e.g., design & analysis (Subpart C), operations (Subpart F)) will use as performance metrics.
This 2nd iteration discussion table explains both the general purpose the preliminary proposed rule language, as well as the changes made since the last published version.
§ 53.200 Safety Objectives. This section provides the overall qualitative safety goals.
Each advanced nuclear plant must be designed, constructed, operated, and decommissioned such that there is The change is to revise the first objective from providing reasonable assurance of adequate protection of to limit the reasonable assurance of adequate protection to limiting the possibility of an immediate threat to the public health and safety possibility of an immediate threat to the public health and safety.
and the common defense and security.. In addition, each This language generally aligns with standards the Commission advanced nuclear plant must take such additional measures to has used for determining the content of technical specifications.
protect public health and minimize danger to life or property as The change also revises the second objective from protect may be reasonableappropriate when considering technology public health and minimize danger to as may be appropriate changes, economic costs, operating experience, or other when considering potential risks to public health and safety.
factorspotential risks to public health and safety. These safety The purpose of these objectives is clarified by adding the objectives shall be carried out by meeting the safety criteria
identified in the assessments performed under the facility safety statement that they will be carried out by meeting the safety program required by § 53.800this subpart. criteria identified in this subpart (§§ 53.210 and 53.220).
This change resulted from stakeholder comments and internal NRC discussions regarding the difficulties in using the Atomic Energy Act (AEA) Sections 182 and 161 authorities as the safety objectives for part 53, and in turn as the bases for the two-tier safety criteria framework. Instead, the use of adequate protection is expected to be used in its traditional role as an NRC regulatory finding, which is presumed through compliance with NRC regulations including part 53 or other license requirements. While Sections 182 and 161 of the AEA will be cited as enabling legislation within the rule package (e.g., in the Federal Register Notice), the staff does not foresee incorporating language from the AEA into the safety objectives or tiers in part 53.
§ 53.2120 First Tier Safety Criteria. The first-tier safety criteria are metrics that establish a level of (a) Normal operations. Design features and programmatic safety or backstop based on current requirements in (a) Part 20 controls must be provided for each advanced nuclear plant to limits on doses to members of the public and (b) the Sections ensure the contribution to total effective dose equivalent to 50.34, 52.79, 100.11 reference values related to the safety individual members of the public from normal plant operation assessment of a site and radiological consequences from a does not exceed 0.1 rem (1 mSv) in a year and the contribution major accident in terms of possible dose to an individual at to dose in any unrestricted area does not exceed 0.002 rem defined distance and duration of exposure. Note that the term (0.02 millisievert) in any one hour.the public dose limits provided programmatic controls is intended to include human actions in Subpart D to 10 CFR part 20. governed by procedures and training.
(b) Unplanned events. Design features and programmatic controls must be provided for each advanced nuclear plant such The change in paragraph (a) is to reference part 20 instead of that analyses of licensing basis events in accordance with summarizing the part 20 requirements in part 53.
§ 53.240, including treatment of uncertainties, demonstrate with high confidence that events with an upper bound frequency This change resulted from stakeholder comments to reference greater than approximately once per 10,000 years meet the vs. repeat part 20 requirements.
following:
(1) An individual located at any point on the boundary of This 2nd iteration maintains the requirements for normal the exclusion area for any 2-hour period following the onset of operations in paragraph (a) of the first tier criteria, specifically the the postulated fission product release would not receive a requirements from Subpart D to part 20. The staff will revisit paragraph (a) as the remaining subparts are developed if the
radiation dose in excess of 25 rem (250 mSv) total effective dose distinction between first and second tier safety criteria for normal equivalent; and operations is not used within future subparts. If a distinction is (2) An individual located at any point on the outer not used in future subparts, then normal operations in the first boundary of the low population zone who is exposed to the and second tiers may instead be treated in a separate section in radioactive cloud resulting from the postulated fission product Subpart B similar to protection of plant workers (see § 53.260). If release (during the entire period of its passage) would not the distinction is used, then this paragraph in § 53.210(a) will receive a radiation dose in excess of 25 rem (250 mSv) total likely remain effective dose equivalent.1 (c) Design features and programmatic controls beyond The change in paragraph (b) is an editorial change to replace those needed for paragraphs (a) and (b) of this section must be with high confidence to including treatment of uncertainties. A provided for each advanced nuclear plant to satisfy additional footnote similar to that included in §§ 50.34(a)(1)(ii)(D)(1) and requirements established by the NRC for ensuring reasonable 52.79(a)(1)(vi)(A) is also inserted to describe the use of 25 rem assurance of adequate protection of the public health and safety as a reference value for evaluating plant design features with and maintaining common defense and security. respect to postulated accidents.
- 1. A whole body dose of 25 rem has been stated to correspond numerically to Paragraph (c) was deleted since the rule is no longer the once in a lifetime accidental or emergency dose for radiation workers which, establishing a direct connection between the first tier safety according to NCRP [National Council on Radiation Protection and Measurements] recommendations at the time could be disregarded in the criteria and the adequate protection standard within Section 182 determination of their radiation exposure status (see NBS Handbook 69 dated of the AEA.
June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under The NRC staff is also considering renaming §§ 53.210 and accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design 52.220 to replace the use of First Tier and Second Tier in the features with respect to postulated reactor accidents, to assure that these section titles and throughout Part 53. However, the basic designs provide assurance of low risk of public exposure to radiation, in the structure would remain. The use of this terminology has caused event of an accident. confusion due to its similarity to the Tier 1 and Tier 2 terminology used in the Part 52 design certification appendices. The staff welcomes stakeholder input on this topic.
§ 53.2320 Second Tier Safety Criteria. The second-tier safety criteria establish metrics consistent with (a) Normal operations. Design features and programmatic current requirements in Part 20 on maintaining doses as low as controls must be provided for each advanced nuclear plant to reasonably achievable (ALARA) for normal operations in ensure the estimated total effective dose equivalent to individual paragraph (a) through the use of design features and members of the public from effluents resulting from normal plant programmatic controls, as appropriate, taking into account the operation are as low as is reasonably achievable taking into factors listed in this paragraph.
account the state of technology, the economics of improvements in relation to the state of technology, operating experience, the The second-tier safety criteria for licensing basis events (i.e.,
economics of improvements in relation toand the benefits to the unplanned events) in paragraph (b) establishes the connection to
public health and safety and other factors included in the licensing basis events and defense in depth in subparagraph (1) assessments performed under the facility safety program and is taken from the NRC safety goals in subparagraph (2).
required by § 53.800. Performance objectives for design. Design These criteria will provide a basis for Subpart C and F.
features and programmatic controls must be established such that: [to be reworded for consistency with 10 CFR part 20 and The change in paragraph (a) is to replace the material taken from 40 CFR part 190]. Appendix I to part 50 with a placeholder to develop language (1) The calculated annual total quantity of all radioactive more directly related to part 20.
material above background to be released from each advanced nuclear plant to unrestricted areas will not result in an estimated This 2nd iteration maintains requirements for providing design annual dose or dose commitment from liquid effluents for any features and programmatic controls to ensure doses to the public individual in an unrestricted area from all pathways of exposure from normal plant operations are ALARA. The ALARA principle in excess of 3 millirems to the total body or 10 millirems to any is a longstanding element of NRC regulations for all licensees organ. (byproduct, source and special nuclear material as well as (2) The calculated annual total quantity of all radioactive utilization facilities) and the NRC staff is proposing to retain material above background to be released from each advanced ALARA in the second tier criteria in this preliminary rule nuclear plant to the atmosphere will not result in an estimated language. Some stakeholder comments seem to raise concerns annual air dose from gaseous effluents at any location near about the level of effort associated with preparing licensing ground level which could be occupied by individuals in applications and NRC reviews for new nuclear power designs.
unrestricted areas in excess of 10 millirads for gamma radiation This issue is addressed in part within the NRC Advanced or 20 millirads for beta radiation. Reactor Content of Application Project (ARCAP). Specifically, (b) Unplanned events. Design features and programmatic draft guidance for ARCAP Chapter 9 (ML20262H264) includes controls must be provided to: the following:
(1) Ensure plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability in lieu of providing detailed system descriptions and analysis of to address licensing basis events in accordance with § 53.240 estimated effluent releases as required by 10 CFR 50.34, 50.34a, and provide measures for defense-in-depth in accordance with 52.47, and 52.79, an application may demonstrate compliance with the
§ 53.250; and applicable regulations by describing a radiation protection program and an effluent release monitoring program that will ensure that effluent (2) Maintain overall cumulative plant risk from licensing release limits will be met during normal operations for the life of the basis events such that the risk to an average individual within the plant. Information related to physical systems can be limited to general vicinity of the plant receiving a radiation dose with the potential descriptions of layout and technologies used to limit the release of the for immediate health effects remains below five in 10 million various inventories of radioactive materials within the plant.
years, and below two in one million years forthe risk to such an individual receiving a radiation dose with the potential to cause As an additional matter, the language emphasizes that ALARA latent health effects remains below two in one million years. will be considered in light of the economics of improvements in relation to the state of technology. Consequently, commenters concern that the ALARA language will undermine the
predictability of the Part 53 licensing process by exposing applicants to an undefined and unwarranted set of requirements is offset by staffs recognition that such requirements must be considered in light of potential costs. See, e.g., Michigan v. EPA, 135 S. Ct. 2699 (2015).
The change in paragraph (b) is editorial.
This 2nd iteration maintains criteria in paragraph (b) for unplanned events that refer to the quantitative health objectives (QHOs) from the NRCs safety goal policy statement. The QHOs are a well-established measure used in NRC risk-informed decisionmaking and are a logical performance metric to support the risk management approaches to operations that will be reflected in Subpart F, Operations. The staff remains open to considering other reasonable alternatives to the use of the QHOs.
§ 53.230 Safety Functions. The safety functions and the design features and programmatic (a) The primary safety function is limiting the release of controls used to fulfill them will be the means to satisfy the first radioactive materials from the facility and must be maintained and second tier safety criteria. These requirements are similar in during routine operation and for licensing basis events over the concept to current requirements for LWRs to address the general life of the plant. design criteria and non-LWRs to provide principal design criteria; (b) Additional safety functions supporting the retention of however, these safety functions are broad in order to support any radioactive materials during routine operation and licensing basis technology.
eventssuch as controlling heat generation, heat removal, and chemical interactions--must be defined. The change in this section (paragraph (c)) is editorial to more
. (c)Design features and programmatic controls serve to clearly state the relationship between the safety functions and fulfill the primary safety function and additional safety functions safety criteria.
and must be maintained over the life of the plant. The primary and additional safety functions are required to meet the first and This 2nd iteration maintains the requirement (paragraph (b)) for second tier safety criteria and are fulfilled by the design features the identification of safety functions (e.g., controlling heat and programmatic controls specified throughout this part. generation, heat removal, and chemical interactions) that are needed to prevent the release of radionuclides instead of prescribing a specific list. This approach is taken in order to support technology inclusiveness, including the need to define
safety functions for inventories of radionuclides outside of primary reactor systems (e.g., waste gas systems if an unplanned release could exceed the safety criteria).
§ 53.240 Licensing Basis Events. This section establishes the requirement to identify and address Licensing basis events must be identified for each licensing basis events (i.e., unplanned events, both internal and advanced nuclear plant and analyzed in accordance with external hazards) to ensure estimates of offsite consequences
§ 53.[3x].450 to support assessments of the safety requirements are below the safety criteria and that SSCs, personnel, and ofin this subpart B. The licensing basis events must address programs address the safety functions. This section provides a combinations of malfunctions of plant SSCs, human errors, and starting point for requirements in Subparts C and F.
the effects of external hazards ranging from anticipated operational occurrences to highlyvery unlikely event sequences The change in this section is editorial.
that are notwith estimated frequencies well below the frequency of events expected to occur in the life of the advanced nuclear This 2nd iteration maintains the identification of licensing basis plant. The evaluation of licensing basis events must be used to events (LBEs) within this broad, higher-level subpart because of confirm the adequacy of design features and programmatic the historical and expected continuing importance of evaluating controls needed to satisfy first and second tier safety criteria of unplanned events as part of the licensing of advanced reactors.
this subpart and to establish related functional requirements for The possibility that this section could be addressed within plant SSCs, personnel, and programs. Subpart C can be considered as part of the later review of the technical requirements, once complete.
§ 53.250 Defense in Depth. This section establishes requirements based on the longstanding Measures must be taken for each advanced nuclear plant nuclear philosophy to ensure defense in depth to address to ensure appropriate defense in depth is provided to uncertainties.
compensate for epistemic and aleatory uncertainties such that there is high confidence that the safety criteria in this subpart B The changes in this section are editorial except for the addition of are met over the life of the plant. The epistemic and aleatory engineered design feature in the description of those items for uncertainties to be considered include those related to the state which defense in depth is required. This change reflects that the of knowledge and modeling capabilities, the ability of barriers to possible crediting of inherent characteristics within the design limit the release of radioactive materials from the facility during and analysis for advanced reactors and the reduced routine operation and for licensing basis events, and those uncertainties associated with such characteristics.
related to the reliability and performance of plant SSCs and personnel, and programmatic controls. Measures to compensate This 2nd iteration maintains defense in depth within this broad, for these uncertainties can include increased safety margins in higher-level subpart because of the historical and continued the design of SSCs and providing alternate means to accomplish importance of its role in addressing the risks associated with safety functions. No single engineered design or operational nuclear power plants. The staff notes that parts 50 and 52 do feature, human action, and or programmatic control, no matter not include a similar section on defense in depth because the how robust, should be exclusively relied upon to meet the safety defense-in-depth philosophy is incorporated into the prescriptive criteria of 10 CFR part§ 53.220(b) or the safety functions defined technical requirements for light water reactors (e.g., the general in accordance with § 53.230. design criteria in Appendix A to part 50). The possibility that this section could be addressed within Subpart C can be considered as part of the later review of the technical requirements, once complete.
§ 53.260 Protection of Plant Workers. The change in paragraph (b) is editorial; referencing part 20 (a) Design features and programmatic controls must exist instead of repeating the requirements in part 53.
for each advanced nuclear plant to ensure that radiological dose to plant workers does not exceed the occupational dose limits provided in subpart C to 10 CFR part 20. This 2nd iteration maintains the protection of plant workers within (b) The licensee As required by Subpart B to 10 CFR part Subpart B to capture occupational exposures within the 20, design features and programmatic controls must use, to the high-level safety requirements. Paragraph (b) maintains extent practical, procedures and engineering controls be based requirements for providing design features and programmatic upon sound radiation protection principles to achieve controls to achieve occupational doses ALARA. See discussion occupational doses and doses to members of the public that are under § 53.220.
as low as is reasonably achievable.
Subpart C, Design and Analysis Requirements 2nd Iteration (Redline/Strikeout) of Discussion Preliminary Rule Language This subpart addresses requirements for designing advanced nuclear plants and performing the supporting analyses, including the analyses of licensing basis events (§ 53.240).
§ 53.400 Design Objectives and Design Features. This section establishes the overall design features by referring Design features must be provided for each advanced to the underlying safety criteria and the related identification of nuclear plant such that, when combined with associated safety functions. Subsequent sections in this Subpart address programmatic controls and human actions, the plant will satisfy the need to define functional design criteria for the design the first and second tier safety criteria defined in §§ 53.220210 features.
and 53.230220. Design features must ensure that the safety functions identified in § 53.210230, of limiting the release of The changes in this section are editorial.
radioactive materials from the facility, is maintained during routine operations and licensing basis events by controlling the This 2nd iteration maintains this section and its role in helping to release of radioactive materials and by supporting other safety establish the general hierarchy of safety criteria, safety function, functions. design feature, and functional design criteria.
§ 53.410 Functional Design Criteria for First Tier Safety Functional design criteria are provided for each design feature to Criteria. ensure safety functions are met and the design complies with (a) Normal operations. Functional design criteria must be first tier and second tier safety criteria.
defined for each design feature required by § 53.400 to demonstrate compliance with the first tier safety criteria defined The changes in this section are editorial.
in § 53.220210(a). Corresponding programmatic controls, including monitoring programs, must be established to confirm This 2nd iteration maintains this section and its role in helping to that the established functional design criteria and the first tier establish the general hierarchy of safety criteria, safety function, safety criteria required in § 53.220210(a) are not exceeded design feature, and functional design criteria. Paragraph (b) will during normal operations. support later sections such as content of Technical (b) Unplanned events. Functional design criteria must be Specifications.
defined for each design feature required by § 53.400 relied upon to demonstrate compliance with the first tier safety criteria See discussions under § 53.210(a) and § 53.220(a) related to defined in § 53.220210(b). Corresponding programmatic ongoing discussions on treatment of normal operations.
controls and interfaces must be established in accordance with this and [other subparts to achieve and maintain the reliability and capability of SSCs relied upon to meet the established
functional design criteria and the first tier safety criteria required in § 53.220210(b), and to maintain consistency with analyses required by § 53.450.
§ 53.420 Functional Design Criteria for Second Tier Safety The changes in this section are editorial (largely deleting Criteria. repetitive text copied from Subpart B).
(a) Design features must be provided for each advanced nuclear plant such that, when combined with associated This 2nd iteration maintains this section and its role in helping to programmatic controls and human actions, the total effective establish the general hierarchy of safety function, design feature, dose equivalent to individual members of the public from and functional design criteria. Paragraph (b) will support later effluents resulting from normal plant operation are as low as is sections such as the identification and implementation of special reasonably achievable taking into account the state of treatment requirements.
technology, the economics of improvements in relation to the state of technology, operating experience, and benefits to the See discussions under § 53.210(a) and § 53.220(a) related to public health and safety, and other factors included in the ongoing discussions on treatment of normal operations and assessments performed under the facility safety program ALARA.
required by § 53.800, and the safety criteria and performance objectives in § 53.230(a). (a) Normal operations. Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with the second tier safety criteria in § 53.230220(a). Corresponding programmatic controls, including monitoring programs, must be established to confirm that the established functional design criteria and the safety criteria and performance objectives in § 53.230220(a) are not exceeded during normal operations.
(b) Design features must be provided for each advanced nuclear plant such that, when combined with associated programmatic controls and human actions, the analyses required by § 53.450 provide reasonable assurance that the estimated risks from unplanned events will be below the second tier safety criteria in § 53.230(b). (b) Unplanned events. Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with the second tier safety criteria in
§ 53.230(b).220(b) considering licensing basis events ranging from anticipated operational occurrences to very unlikely event sequences with estimated frequencies well below the frequency
of events expected to occur in the life of the advanced nuclear plant. Corresponding programmatic controls and interfaces must be established in accordance with this and other subparts to achieve and maintain the reliability and capability of SSCs relied upon to meet the second tier safety criteria in § 53.230220(b) and to maintain consistency with analyses required by § 53.450.
§ 53.430 Functional Design Criteria for Protection of Plant This section addresses design features and functional design Workers. criteria related to protection of plant workers.
Design features must be provided for each advanced nuclear plant such that, when combined with associated No changes in this section for the 2nd iteration.
programmatic controls and human actions, there is reasonable assurance the requirements for the protection of plant workers in The 2nd iteration maintains this section but expects that future
§ 53.260 will be met. Functional design criteria must be defined interactions related to ARCAP or other guidance will be needed for each design feature relied upon to demonstrate compliance to ensure an appropriate balance between initial design, with § 53.260. Corresponding programmatic controls, including licensing reviews, and performance monitoring.
monitoring programs, must be established to confirm that the worker protection criteria in § 53.260(a) are not exceeded. In See discussions under §§ 53.220 and 53.260.
addition, functional design criteria must be defined for each design feature to ensure that plant SSCs and associated programmatic controls, including monitoring programs, achieve occupational doses as low as is reasonably achievable as required by § 53.260(b).
§ 53.440 Design Requirements. This section addresses design requirements by defining the (a) The design features required to meet the first and means by which functional design criteria are met through second tier safety criteria defined in §§ 53.220210 and practices such as the use of generally accepted consensus 53.230220 shall be designed using generally accepted codes and standards and qualification of equipment/materials -
consensus codes and standards wherever applicable. including provisions similar to those in 10 CFR 50.43(e).
(b) The materials used for safety related and non-safety related but safety significant SSCs ([as will be defined in § Paragraph (c) addresses security by design from the Advanced 53.460)subpart A] must be qualified for their service conditions Reactor Policy Statement.
over the plant lifetime.
(c) Safety and security must be considered together in The changes in this section are editorial. The changes in the design process such that, where possible, security issues are paragraph (d) were made to more closely align with the language in 10 CFR 50.43(e) regarding the demonstration of capabilities
effectively resolved through design and engineered security through combinations of analyses, testing, and operational features. experience.
(d) Design features must be demonstrated capable of accomplishing the safety functions defined in § 53.210 without adversely affecting other design features. The demonstration must be through analysis consistent with § 53.450fulfilling functional design criteria considering interdependent effects through analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof for the range of conditions under which the analysis required in § 53.450 assumes these features will function throughout the plants lifetime.
§ 53.450 Analysis Requirements. This section addresses analyses requirements.
(a) Requirement to have a probabilistic risk assessment.
A probabilistic risk assessment (PRA) of each advanced nuclear The change in this section is primarily to paragraph (b) and is plant [reminder - plant definition to include multi-module and intended to support alternative approaches to a PRA for activities multi-source] must be performed to identify potential failures, such as selection of licensing basis events, safety classification, degradation mechanisms, susceptibility to internal and external and evaluating defense in depth. The alternative language is hazards, and other contributing factors to unplanned events that worded similar to other requirements in terms of generally might challenge the safety functions identified in § 53.210.230 accepted to support possible standards or other guidance and to support demonstrating that each advanced nuclear plant documents. Possible examples of alternative approaches that meets the second tier safety criteria of § 53.220(b). might be found acceptable include design-related consensus (b) The probabilistic risk assessment (PRA) must: codes and standards or IAEA specific safety requirements.
(1) Be used in(b) Specific uses of analyses. The PRA, other generally accepted risk-informed approach for This 2nd iteration does maintain a requirement to have a PRA to systematically evaluating engineered systems, or combination support the licensing and regulatory programs being developed thereof must be used: in subsequent subparts (e.g., a risk management approach for (1) In determining the licensing basis events, as operations). The staff is engaged in ongoing discussions on how described in § 53.240, which must be considered in the design to to ensure the level of effort required for a PRA is commensurate determine compliance with the safety criteria in Subpart B of this with the complexity of the subject reactor design. This topic will part. continue to be discussed during the development of Part 53 and (2) Be used forFor classifying SSCs and human actions related guidance, including the development of regulatory according to their safety significance in accordance with § 53.460 guidance related to the non-light water reactor PRA standard.
and for identifying the environmental conditions under which the SSCs and operating staff must perform their safety functions.
(3) Be used inIn evaluating the adequacy of defense-in- Paragraph (c) (formerly (b)(6) & (b)(7)) was revised to refer to depth measures required in accordance with § 53.250. generally accepted standards or guidance instead of defining (4) AssessTo identify and assess all plant operating specific periodicity for upgrading analysis tools.
states where there is the potential for the uncontrolled release of radioactive material to the environment. Changes to paragraph (d) (formerly (c)) are editorial.
(5) ConsiderTo identify and assess events that challenge plant control and safety systems whose failure could lead to the Former paragraph (d) relocated to paragraph (g).
uncontrolled release of radioactive material to the environment.
These include internal events, such as human errors and Paragraph (e) added to further clarify requirements for analysis equipment failures, and external events, such as earthquakes, of licensing basis events, including anticipated operational identified in accordance with Subpart D of this part. occurrences. This addition was in response to comments from (6) Conform(c) Maintenance and upgrade of analyses. external stakeholders and ACRS members.
The PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination Addition to paragraph (f) to address event sequences from thereof must be maintained and upgraded in conformance with initiation to a safe stable end state for DBAs was in response to generally accepted methods, standards, and practices. comments from ACRS members.
(7) Be maintained and upgraded to cover initiating events and modes of operation contained in generally accepted Changes to paragraph (f) on design basis accidents and methods, standards, and practices in effect one year prior to paragraph (g) on other required analyses are to capture the each required PRA upgrade. The PRA must be upgraded every possible use of a systematic approach other than PRA to support two years until the permanent cessation event selection and design processes.
(d) Qualification of operations under Subpart G of this part.(c)analytical codes. The analytical codes used in modeling plant behavior duringin analyses of licensing basis events (e.g.
thermodynamics, reactor physics, fuel performance, mechanistic source term) must be qualified for the range of conditions for which they are to be used.
(d) If not addressed within the PRA under paragraph (b),
analyses must be performed to assess:
(1) measures provided to protect against, detect and suppress fires that could impact the ability of equipment to perform its safety function and challenge the safety criteria contained in §§ 53.220 and 53.230.
(2) measures provided to protect against aircraft impacts as required by 10 CFR 50.150, and
(3) measures to mitigate specific beyond design basis events as required by 10 CFR 50.155.
(e)(e) Analyses of licensing basis events. Analyses must be performed for licensing basis events ranging from anticipated operational occurrences to very unlikely event sequences with estimated frequencies well below the frequency of events expected to occur in the life of the advanced nuclear plant. The licensing basis events must be identified using insights from a PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof to systematically identify and analyze equipment failures and human errors. The analyses must address event sequences from initiation to a defined end state and demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the second tier safety criteria of § 53.220(b) and provide defense in depth as required by § 53.250.
(f) Analysis of design basis accidents. The analysis of licensing basis events required by § 53.240 and § 53.450(e) must include analysis of a set of design basis accidents that address possible challenges to the safety functions identified in accordance with § 53.210230. Design basis accidents must be selected from those unanticipated event sequences with an upper bound frequency of less than one in 10,000 years as identified using insights from a design-specific probabilisticPRA, other generally accepted risk assessment that-informed approach for systematically identifiesevaluating engineered systems, or combination thereof to systematically identify and analyzesanalyze equipment failures and human errors. The events selected as design basis accidents should be those that, if not terminated, have the potential for exceeding the safety criteria in § 53.220210(b). The design-basis accidents selected must be analyzed using deterministic methods assumingthat address event sequences from initiation to a safe stable end state and assume only the safety--related SSCs identified in
§ 53.460 and human actions addressed by § 53.8xx (reference
to concept of operations sections of Subpart F) are available to perform the safety functions identified in accordance with
§ 53.210230. The analysis must conservatively demonstrate compliance with the safety criteria in § 53.220210(b).
(g) Other required analyses. If not addressed within the PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof under paragraph (b), analyses must be performed to assess:
(1) measures provided to protect against, detect and suppress fires that could impact the ability of equipment to perform its safety function and challenge the safety criteria contained in §§ 53.210 and 53.220.
(2) measures provided to protect against aircraft impacts as required by 10 CFR 50.150, and (3) measures to mitigate specific beyond design basis events as required by 10 CFR 50.155.
§ 53.460 Safety Categorization and Special Treatment. This section addresses the safety classification and (a) SSCs and human actions must be classified determination of appropriate special treatments.
according to their safety significance. The categories must include Safety Related (SR), which are those SSCs and human The changes in this section are editorial (largely deleting text that actions relied upon to function in response to design basis is expected to be provided within the definitions section in accidents to meet the safety criteria in § 53.220(b); Non-Safety Subpart A).
Related but Safety Significant (NSRSS), which are those SSCs and human actions that perform a function that is necessary to This 2nd iteration does maintain the use of the safety-related achieve adequate defense-in-depth or are classified as risk designation. Some stakeholders have proposed a more flexible significant (i.e., whose failure contributes 1% or more to but less defined approach to SSC classification. This is an cumulative plant risk, as defined in § 53.230, or would cause a important topic for future discussions given the potential role of licensing basis event to exceed the safety criteria in § 53.220(b)); SSC classification throughout part 53. The staff is interested in and Non-Safety Significant (NSS), which are those SSCs not examples of international practices or other approaches that warranting special treatmentNon-Safety Related but Safety would not align with the preliminary rule language in Part 53.
Significant (NSRSS), and Non-Safety Significant (NSS), as defined in subpart A of this part.
(b) For SR and NSRSS SSCs and human actions, the conditions under which they must perform their safety function in
§ 53.210230 must be identified. Special Treatment (e.g.,
functional design criteria and programmatic controls) must be established in accordance with this and [other Subparts to provide appropriate confidence that the SSCs will perform under the service conditions and with the reliability assumed in the analysis performed in accordance with § 53.450 to provide reasonable assurance of meeting the safety criteria in
§§ 53.220210(b) and 53.230220(b).
(c) Human actions to prevent or mitigate licensing basis events must be capable of being reliably performed under the postulated environmental conditions present and be addressed by programs established in accordance with Subpart F of this part to provide confidence that those actions will be performed as assumed in the analysis performed in accordance with § 53.450 to provide reasonable assurance of meeting the safety criteria in
§§ 53.220210(b) and 53.230220(b).
§ 53.470 Application of Analytical Safety Margins to This section addresses the potential for operational flexibility Operational Flexibilities. through the applicants or licensees adoption of more restrictive Where an applicant or licensee so chooses, design criteria in order to obtain safety margin for application to other criteria more restrictive than those defined in § 53.230220(b) areas - such as emergency planning zones. The section may be adopted to support operational flexibilities (e.g., establishes requirements to use the more restrictive design goal emergency planning requirements under Subpart F of this part). similar to the second tier safety criteria and to ensure analysis, In such cases, applicants and licensees must ensure that the design features, and programmatic controls are established functional design criteria of § 53.420(b), the analysis accordingly, if operational flexibilities are sought.
requirements of § 53.450, and identification of special treatment of SSCs and human actions under § 53.460 reflect and support This 2nd iteration maintains this section on applying and the use of alternative design criteria to obtain additional subsequently controlling analyses and designs in cases where analytical safety margins. Licensees must ensure that measures analytical margins have been used to gain operational taken to provide the analytical margins supporting operational flexibilities. The staff remains interested in feedback on the flexibilities are incorporated into design features and possible uses of analytical safety margins and how that concept programmatic controls and are maintained within programs can be best addressed within Part 53.
required in other Subparts.
§ 53.480 Design Control Quality Assurance. This section addresses quality assurance for design and analysis (a) Measures must be established to assure that the activities and is derived from Criterion III in Appendix B to 10 design criteria, analysis, categorization and special treatment of CFR Part 50. Paragraph (a) preliminary rule language refers to SSCs as required by § 53.460 are correctly translated into generally accepted consensus codes and standards to provide specifications, drawings, procedures, and instructions. These flexibility for the possible development of guidance that could measures must include provisions to assure that appropriate identify potentially acceptable alternatives to NQA-1.
quality standards are specified and included in design documents and that deviations from such standards are The change in this section is editorial.
controlled. Measures must also be established for the selection and review for suitability of application of materials, parts, equipment, and processes needed to meet the safety criteria identified per §§ 53.220210 and 53.230220 in accordance with § 53.xxx (construction and procurement subpart).Subpart E of this part. The QA program must conform with generally accepted consensus codes and standards.
(b) Measures must be established for the identification and control of design interfaces in accordance with § 53.490.
(c) The design control measures must provide for verifying or checking the adequacy of design in a manner commensurate with its safety significance, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The verifying or checking process must be performed in accordance with appropriate quality standards.
Design changes, including field changes, must be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another qualified organization.
§ 53.490 Design and Analyses Interfaces. This section requires applicants/licensees to identify, control, and Measures must be established for the identification and maintain interfaces (i.e., integration) between design and control of interfaces between (a) the plant design and supporting analyses activities and other activities, such as configuration analyses required by this Subpart and (b) the activities controls in Subpart F and the proposed facility safety program.
addressed by other Subparts over the life of each advanced nuclear plant. These measures must include procedures for the No changes in the section for the 2nd iteration.
review, approval, release, distribution, and revision of documents involving design interfaces such that design decisions are made in an integrated fashion considering all aspects of the facility impacted by the design or operational change prior to its implementation. Changes to design features and related programmatic controls over the lifetime of an advanced nuclear plant must be considered along with the state of technology, the economics of improvements in relation to the state of technology, operating experience, and benefits to the public health and safety, and other factors included in the assessments performed under the facility safety program required by § 53.800.