ML21106A001
| ML21106A001 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/2021 |
| 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: ML21106A001 (6) | |
Text
THIS PRELIMINARY 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.
Subpart F - Requirements for Operation Subpart F is part of the Part 53 structure and format described in previous public meetings. Subpart F defines the requirements during the operation phase of an advanced nuclear plant to ensure the safety criteria (Subpart B) and other areas of Part 53 (e.g., design & analysis (Subpart C)) continue to be satisfied throughout the plants lifetime.
§ 53.700 Operational objectives. This section provides the overall objectives and general Each licensee shall define, implement, and maintain organization of Subpart F, which is to define requirements on:
controls for plant SSCs, responsibilities of plant personnel, and plant programs during the operating life of each advanced 1) Plant SSCs (e.g., configuration control, testing) nuclear plant such that the first and second tier safety criteria (§ 53.700 - 53.740) defined in §§ 53.210 and 53.220 are satisfied. Each licensee 2) Plant personnel (e.g., licensing, training) (§ 53.XXX) and shall maintain the capabilities and reliabilities of facility 3) Plant programs (e.g., radiation protection, security) structures, systems, and components to ensure that the safety (§ 53.800 - 53.900).
functions identified in § 53.230 will be performed if called upon during normal operations and licensing basis events. Each These requirements are needed to ensure that the advanced licensee shall ensure that plant personnel have adequate nuclear plant is maintained and operated such that the first and knowledge and skills to perform their assigned duties that second tier safety criteria are met during all modes of normal support the performance of the safety functions identified in operation.
§ 53.230. Each licensee shall implement plant programs during operations sufficient to ensure that the safety functions identified in § 53.230 will be performed if called upon during normal operations and licensing basis events.
§ 53.710 Transition from construction/manufacturing This section requires a transition plan from construction to to operation. operations. This paragraph may be revised once the remainder The applicant or licensee shall prepare a transition plan of Part 53 is complete to account for ITAAC-related issues.
from construction to operations for each advanced nuclear plant. The plan must identify all testing or verifications required A possible discussion topic is whether these requirements to: would be more logically addressed as a startup testing program in the programs-related sections of this Subpart, rather than 1
(a) Before plant operation, demonstrate that the SR addressed separately as illustrated in the preliminary proposed and NSRSS SSCs have been appropriately constructed or rule language in this section.
manufactured to further ensure those SSCs have the capabilities needed to perform or support the safety functions of
§ 53.230 and satisfy the first and second tier safety criteria defined in §§ 53.210 and 53.220; (b) Demonstrate that plant personnel are appropriately licensed, trained, and otherwise capable and available to support the safety functions of § 53.230 and satisfy the first and second tier safety criteria defined in §§ 53.210 and 53.220; (c) Demonstrate that all programs, procedures, and controls have been prepared and implemented to support the safety functions of § 53.230 and satisfy the first and second tier safety criteria defined in §§ 53.210 and 53.220.
§ 53.720 Maintaining capabilities and availability This section provides the requirements for maintaining of structures, systems, and components. capabilities, availability, and reliability of SSCs to support Controls must be provided for each advanced meeting the first and second tier safety criteria for unplanned nuclear plant such that the capabilities and reliability of SSCs, events that are described in Subpart B. The basic structure of when combined with associated programmatic controls and this section is that controls for safety-related (SR) SSCs are human actions, provide reasonable assurance that the safety provided by technical specifications (paragraph (a)) and controls criteria defined in §§ 53.210(b) and 53.220(b) will be met. for non-safety related but safety significant (NSRSS) SSCs are required to be addressed with licensee-controlled documents and procedures (paragraph (b)).
(a) Technical Specifications must be developed and This paragraph defines the required limits to be included in implemented that define conditions or limitations on plant Technical Specifications (TS). The submittal and control of TS operations that are necessary to provide reasonable assurance as a key licensing document will be addressed in Subparts H, that SR SSCs fulfill the safety functions identified in § 53.230 Licenses, Certifications and Approvals, and I, Maintaining and and that satisfy the first tier safety criteria of § 53.210(b). The Revising Licensing Basis Information. The general content and technical specifications must describe the following control of TS under Part 53 will be similar to the requirements in requirements: Parts 50 and 52. The requirements for TS include limits on the (1) Limits on the inventory of radioactive materials inventories of radioactive materials, plant operating limits, and within the reactor system and supporting systems with the specific requirements for each SR SSC, including limiting potential, individually or collectively, to cause a release conditions for operation and required surveillances. The exceeding the safety criteria in § 53.210(b) as a result of a proposed requirements for TS also include sections on 2
design basis accident analyzed in accordance important design attributes (similar to design features in with § 53.450(e). § 50.36 but different term may be needed if design feature (2) Operating limits for the facility that if exceeded becomes a defined term within Part 53 to mean something could lead to a failure to perform a required safety function different than in Part 50), administrative controls, and necessary to meet the safety criteria in § 53.210(b). decommissioning (when applicable).
(3) For each SSC classified as SR in accordance with § 53.460, technical specifications must define: This first iteration of preliminary language for this section of Part (i) Limiting conditions for operation. Limiting 53 does not include the concept of safety limits or associated conditions for operation are the lowest functional capability limiting safety system settings from 10 CFR 50.36. As or performance levels of SR SSCs required to provide discussed in SECY-18-0096, systematic assessments and more reasonable assurance that the design basis accidents mechanistic approaches to evaluating source terms support an analyzed in accordance with § 53.450(e) would not give alternative approach to establishing barrier-based safety limits.
rise to an immediate threat to the public health and safety An example provided in that paper is a comparison of (1) the as represented by the first tier safety criteria of § traditional specified acceptable fuel design limits (SAFDLs) that 53.210(b). When a limiting condition for operation is not are generally used as performance measures or safety limits for met, the licensee must shut down the plant or follow any reactor protection systems in LWRs to protect a specific barrier remedial action permitted by the technical specifications from potential failure mechanisms (e.g., departure from nucleate until the condition will be met. boiling) and (2) the specified acceptable radionuclide release (ii) Surveillance requirements. Surveillance design limit (SARRDL) concept, which establishes limits on the requirements relate to test, calibration, or inspection to possible increase in circulating radionuclide inventory during assure that the necessary quality of systems and normal operations or an anticipated operational occurrence as components is maintained and that the limiting conditions part of an integrated or functional containment approach. The for operation will be met. SARRDL could be addressed under subparagraph (2) as an (4) Design attributes. Design attributes to be included operating limit within this preliminary construct of requirements are those attributes of the facility such as materials of for TS.
construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not This first iteration of preliminary language for this section of Part covered in categories described in paragraphs (a)(1)-(3) of this 53 does not include the criteria for identifying limiting conditions section. for operation (LCOs) from 10 CFR 50.36. Instead, the staff (5) Administrative controls. Administrative controls are proposes to maintain the concepts from Subparts B and C and the provisions relating to organization and management, define TS LCOs as providing limits on safety-related SSCs, procedures, recordkeeping, review and audit, and reporting which are those associated with protecting against an necessary to assure operation of the facility in a safe immediate threat to public health and safety and the first tier manner. Each licensee must submit any reports to the safety criteria in § 53.210(b). 10 CFR 50.36(c)(2)(ii) provides Commission pursuant to approved technical specifications as the criteria for limiting conditions for operation and includes specified in § 53.40. criterion (D) A structure, system, or component which operating 3
(6) Decommissioning. This paragraph applies only to experience or probabilistic risk assessment has shown to be advanced nuclear plants that have submitted the certifications significant to public health and safety. In this preliminary required by subpart G of this part. Technical specifications construct for Part 53, risk significant SSCs are addressed involving limiting conditions for operation; surveillance through a combination of TS for the safety-related SSCs and requirements; design features; and administrative controls will the introduction of paragraph (b) of this section for NSRSS be developed on a case-by-case basis. SSCs.
Note that guidance, Part 53 rule language, or addressing some designs within a revised Part 50 are being explored as possible ways to accommodate deterministic approaches for performing the design and analysis described in Subpart C. Developing approaches within this rulemaking for different design philosophies is a continuing topic of discussion.
(b) Controls on plant operations, including Paragraph (b) defines the required controls to be developed and availability controls, must be developed and implemented for NSRSS SSCs. Configuration management implemented to provide reasonable confidence that and other special treatments provide reasonable confidence the configurations and special treatments for that the capabilities and reliabilities of SSCs are maintained NSRSS SSCs provide the capabilities and reliabilities consistent with the underlying risk assessments. The staff required to satisfy the second tier safety criteria of notes that these or similar controls are needed to implement a
§ 53.220(b). The controls must: performance-based approach and to gain operational flexibilities (1)(i) Identify who within the advanced nuclear plant has in areas such as replacing the single-failure criterion with a authority to make configuration changes; probabilistic (reliability) approach, clarifying the appropriate (ii) Establish processes to make configuration changes classification and controls over SR SSCs and NSRSS SSCs, to the advanced nuclear plants system; and and supporting the staffing and program topics in this Subpart.
(iii) Establish processes to ensure that all departments of the advanced nuclear plant affected by the configuration As mentioned in the paragraph (a) discussion, guidance or changes are formally notified and approve of the change; and other changes may be needed to address deterministic (2) Describe the means by which the special treatments approaches with different supporting analysis, safety for each NSRSS SSC will be provided and maintained over the classification schemes, and design approaches (e.g., inclusion operating life of the advanced nuclear plant. [examples would of the single failure criterion). This topic also carries through include appropriate surveillances, reliability assurance other configuration control and program requirements that programs, etc.] differentiate between SR and NSRSS SSCs based on risk-informed, performance-based concepts.
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§ 53.730 Maintenance, repair and inspection programs. This section provides the requirements for developing and (a) A program to control maintenance activities and implementing a program to: (a) control maintenance activities; monitor the performance or condition of SR and SS SSCs must (b) take corrective action when performance issues are be developed and implemented to provide reasonable identified; (c) conduct routine evaluations of effectiveness; and assurance that the safety criteria defined in §§ 53.210(b) and (d) assess and manage risks resulting from maintenance 53.220(b) of this part will be met. activities. These requirements are similar to those included in (b) Whenever a licensee determines through activities 10 CFR 50.65 (maintenance rule) with the scope revised to related to maintenance, repair, and inspection of SSCs, the remain consistent with the safety criteria in Subpart B.
activities under § 53.720, or otherwise that the performance or condition of a NSRSS SSC does not meet established special treatment requirements or performance goals related to capabilities or reliabilities, the licensee must take appropriate corrective action.
(c) Performance and condition monitoring activities and associated goals and preventive maintenance activities must be evaluated at least every 24 months. The evaluations must take into account, where practical, industry-wide operating experience. Adjustments must be made where necessary to ensure that the objective of preventing failures of SSCs through maintenance is appropriately balanced against the objective of minimizing unavailability of SSCs due to monitoring or preventive maintenance.
(d) Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee must assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to SSCs that a risk-informed evaluation process determines are necessary to provide reasonable assurance that the performance measures defined in
§§ 53.210(b) and 53.220(b) of this part will be met.
§ 53.740 Design control. This section provides the requirements for assessing design or The potential for adverse effects on safety, security, EP, procedural changes during the operation phase of a facility to operations, or other items related to plant safety must be ensure safety functions continue to be satisfied and the assessed during the design process and before implementing 5
design or operational changes. This includes planned and interfaces between various design features, programs, and emergent changes, such as physical modifications, procedural procedures are appropriately considered.
changes, changes to operator actions or security assignments, maintenance activities, system reconfigurations, access modifications or restrictions, and changes to the emergency plan and security plan or their implementation. Accordingly, measures must be established for the identification and control of interfaces among plant activities. These measures must include procedures for the 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.
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