ML21106A002
| ML21106A002 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/2021 |
| From: | Robert Beall NRC/NMSS/DREFS/RRPB |
| To: | |
| Robert Beall | |
| Shared Package | |
| ML20289A534 | List:
|
| References | |
| 10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31 | |
| Download: ML21106A002 (15) | |
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 controls for organization of Subpart F, which is to define requirements on plant SSCs, responsibilities of plant personnel, 1) Plant SSCs (e.g., configuration control, testing) and plant programs during the operating life of each advanced (§ 53.700 - 53.740) nuclear plant such that the first and second tier safety criteria 2) Plant personnel (e.g., licensing, training) (§ 53.XXX) and defined in §§ 53.210 and 53.220 are satisfied. Each licensee 3) Plant programs (e.g., radiation protection, EP, security) shall maintain the capabilities and reliabilities of facility (§ 53.800 - 53.900).
structures, systems, and components to ensure that the safety The requirements are needed to ensure that the advanced functions identified in § 53.230 will be performed if called upon nuclear plant is maintained and operated such that the first and during normal operations and licensing basis events. Each second tier safety criteria are met during all modes of normal licensee shall ensure that plant personnel have adequate operation.
knowledge and skills to perform their assigned duties that support the performance of the safety functions identified in
§ 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.
§ 53.800 PROGRAMS This section requires programs to be developed, implemented Programs must be provided for each advanced nuclear plant and maintained to help ensure that design features and human such that, when combined with associated design features and actions have the necessary capabilities and reliabilities needed human actions, the plant will satisfy the first and second tier to meet safety criteria in Subpart B throughout the operating life safety criteria defined in §§ 53.210 and 53.220. Programs must of each advanced nuclear plant.
also support continued assurance that the safety functions
identified in § 53.230 are maintained during normal operations The following sections describe requirements for selected and licensing basis events. The required plant programs must programs.
include but are not necessarily limited to the programs described in the following sections of this Subpart.
§ 53.810 Radiation Protection (a) Each licensee under this part must develop and implement a Paragraph (a) requires a radiation protection program Radiation Protection Program for operations that is associated with the requirements in Subpart B for normal commensurate with the scope and extent of licensed activities operations and protection of plant workers and is based on under this part and includes measures for limiting and language in 10 CFR 20.1101. Subparts H and I will describe monitoring radioactive plant effluents and limiting and the submittal requirements for the radiation protection program.
monitoring the dose to individuals working with radioactive materials in accordance with 10 CFR part 20.
(b) Each licensee under this part must develop, implement, and Paragraph (b) requires a program similar to that required in maintain a program for the control of radioactive effluents and 10 CFR 50.36a, Technical specifications on effluents from for keeping the doses to members of the public from radioactive nuclear power plants, and related requirements in standard effluents as low as reasonably achievable. The program shall technical specifications for offsite dose calculation manuals.
be contained in an Offsite Dose Calculations Manual (ODCM),
shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded.
The ODCM shall:
(i) contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and (ii) contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by § 53.xyz [Subpart J].
(c) [Additional provisions may be added if needed]
§ 53.820 Emergency Preparedness [work in progress] This section will discuss emergency plans, which for Parts 50 (a) Each licensee under this part must develop and implement and 52 licensees will be addressed by 10 CFR 50.160 or the an Emergency Preparedness Program for operations that is
commensurate with the risks posed by the licensing basis requirements in appendix E to 10 CFR Part 50 and, for nuclear events as analyzed in accordance with § 53.450. power reactor licensees, the planning standards of § 50.47(b).
Developing preliminary program requirements for The development of this program is being coordinated with onsite and offsite emergency preparedness programs ongoing activities.
in coordination with activities related to proposed rule Emergency Preparedness Requirements for Small Modular Reactors and Other New Technologies.
§ 53.830 Security Programs [work in progress] This section will discuss several security-related programs.
(a) Each licensee under this part must develop and implement The proposal for Part 53 will build on the consequence-based Information Security Program approach developed for the limited-scope rulemaking on Physical Security Program physical security requirements for advanced nuclear reactors by Cyber Security Program applying a performance-based, graded approach to a range of Access Authorization Program security areas, including physical security, cyber security, Material Control and Accounting information security, transportation security, fitness for duty, and access authorization.
Developing preliminary security program requirements for future release and discussion The staff is developing preliminary language for these programs for discussion during future meetings with stakeholders.
§ 53.840 Quality Assurance This section addresses quality assurance for operations and is (a) Each licensee under this part is responsible for the derived from the criteria in Appendix B to 10 CFR Part 50.
establishment and execution of the quality assurance program Paragraph (a) preliminary rule language refers to generally (QAP). The licensee may delegate to others, such as accepted consensus codes and standards to provide flexibility contractors, agents, or consultants, the work of establishing and for the possible development of guidance that could identify executing the quality assurance program, or any part thereof, potentially acceptable alternatives to NQA-1.
but shall retain responsibility for the QAP. The authority and duties of persons and organizations performing activities affecting the safety-related functions of structures, systems, and components shall be clearly established and delineated in the QAP. A written QAP manual must be developed and used to guide the conduct of the program in accordance with generally accepted consensus codes and standards. QA activities must be based upon written procedures and address the following:
- 1. Use of qualified personnel and written procedures
- 2. Procurement of goods and services
- 3. Handling, shipping and storage of materials and components
- 4. Testing and inspections
- 5. Corrective actions
- 6. Document control
- 7. Configuration control
- 8. Design control
- 8. Record keeping
- 9. Auditing (b) The results of the QA activities must be documented, along with any recommendations, and provided to facility management for action. Audits, both planned and unannounced, of the QAP must be performed to assess its effectiveness, with the results documented and provided to facility management.
(c) [Additional provisions may be added if needed]
§ 53.850 Integrity Assessment Programs The purpose of this section is to require licensees to actively Each licensee under this part must develop and implement an assess possible degradation of SSCs from the effects of aging, integrity assessment program to monitor, evaluate and manage: fatigue, and environmental conditions.
(a) the effects of plant aging on SR and NSRSS SSCs as well as any NSS SSCs whose failure could affect the performance of The purpose of the Aging Management Program addressed in plant safety functions. The program may refer to surveillances, Paragraph (a) is to ensure that the integrity and performance of tests, and inspections conducted for specific SSCs in the SR and NSRSS SSCs remain consistent with the original accordance with other requirements in this part or conducted in design and that safety margins are maintained over the life of accordance with applicable accepted consensus codes and the plant sufficient to compensate for uncertainties. The standards; program should be based upon a comprehensive and integrated (b) cyclic or transient load limits to ensure SSCs are maintained evaluation of the aging mechanisms applicable to the design, within the applicable design limits; identification of the SSCs the aging mechanisms apply to, the (c) degradation mechanisms related to chemical interactions, allowances provided in the design of the SSCs for aging effects operating temperatures, effects of irradiation, and other and plans, schedules and procedures for determining if environmental factors to ensure the capabilities and reliabilities degradation has been caused by aging. Risk insights may be of SSCs satisfy the functional design criteria of §§ 53.410(b) used to prioritize the aging examinations based upon the and 53.420(b). importance of the SSC to safety and the time frame when the
aging effects could be of concern. Risk insights may also be used to determine when to perform the aging examinations (e.g., full power, shutdown, refueling) so as to minimize the risk to the plant and the public. Acceptance criteria for determining whether or not corrective action is needed should be developed for use in examining the aging effects. The results of the aging examinations should be provided to those responsible for determining what, if any, corrective action is needed and when it should be done. The aging examination results and corrective actions taken should be documented and retained over the life of the plant as described in Subpart J.
Paragraphs (b) and (c) require programs similar to the aging management program described above to address cyclic/transient loadings and environmental effects that could result in the degradation of materials or related SSCs. This preliminary language is written assuming the programs for aging management, cyclic loads, and environmental effects are incorporated into an overall integrity assessment program.
§ 53.860 Fire Protection This section addresses requirements for a fire protection plan (a)(1) Each licensee under this part must have a fire protection supporting operations and is based on the requirements in plan that describes the overall fire protection program for the 10 CFR 50.48 for operating reactors.
facility, identifies the various positions within the licensee's organization that are responsible for the program, states the A future iteration of this section will be considered in concert authorities that are delegated to each of these positions to with specific fire protection requirements within the preliminary implement those responsibilities, and outline the plans for fire language in Subpart C for design and analyses and in protection, fire detection and suppression capability, and Subpart E on construction and manufacturing.
limitation of fire damage.
(2) The fire protection plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as: administrative controls and personnel requirements for fire prevention and manual fire suppression activities; automatic and manually operated fire detection and suppression systems; and the means to limit fire damage to SR and NSRSS structures, systems, or components
so that the capability to meet the requirements of § 53.210 is ensured.
(b)(1) Each licensee under this part must develop a performance-based or deterministic fire protection program that seeks to meet the safety criteria outlined in §§ 53.210 and 53.220, related safety functions outlined in § 53.230, and defense in depth as outlined in § 53.250 with specific fire protection measures related to fire prevention, fire detection, and fire suppression..
(2) The fire protection program must comply with the following:
(i) SR and NSRSS structures, systems, and components must be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
(ii) Noncombustible and fire- resistant materials shall be used wherever practical throughout the facility, particularly in locations with SR and NSRSS structures, systems, or components.
(iii) Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on SR and NSRSS structures, systems, and components.
(iv) Firefighting systems shall be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.
§ 53.870 Inservice Inspection/Inservice Testing This section addresses requirements for inservice inspection (a) Each applicant and licensee under this part must develop and inservice testing programs, which are historically an and implement programs for In-Service Inspection (ISI) and In- important activity conducted in accordance with ASME codes Service Testing (IST) prior to receiving an operating license. and regulations in 10 CFR 50.55a, Codes and standards.
The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be A topic for discussion is whether this requirement could be supplemented by risk insights that identify the most important incorporated into § 53.720 Maintaining capabilities and SSCs to plant safety. The types of testing and inspections and availability of structures, systems and components.
their frequency should be informed by the risk insights so as to
maintain the SSC reliability and performance consistent with the design. Risk insights must also be used to determine when to conduct the inspections and tests (e.g., full power, shutdown, refueling) so as to minimize risk to the plant and the public. The ISI/IST program must be documented in a written manual and managed by qualified personnel reporting to the Plant Manager.
(b) Prior to starting plant operation, baseline inspections and testing must be performed using the same techniques as will be used for future inspections and testing. These inspection and testing results must be used as benchmarks for evaluating the results of future inspections and tests. Sufficient room and support must be provided to accommodate the personnel, ISI/IST equipment and shielding necessary to perform the inspections and testing. Acceptance criteria for determining whether or not corrective action is needed must be developed (or taken from the codes and standards used in the design) for evaluating the results of the inspections and tests. The results of the inspections and testing must be provided to the Plant Manager who is responsible for determining what, if any, corrective action is needed and when it should be done. The ISI/IST results and corrective actions must be documented and retained over the life of the plant.
§ 53.880 Criticality Safety Program This section addresses requirements for criticality safety.
(a) Each licensee under this part must have a criticality safety program. The program must address the requirements in A topic for discussion is whether the alternatives to 10 CFR 70.24 of this chapter for maintaining a monitoring 10 CFR 70.24 provided in 10 CFR 50.68, Criticality accident system capable of detecting a criticality, having emergency requirements, are appropriate and useful in Part 53.
procedures, and providing radiation protection for plant workers.
§ 53.890 Facility Safety Program The first iteration preliminary rule language and discussion table Each licensee must establish and implement a facility safety for the proposal to require an ongoing facility safety program program (FSP) that routinely and systematically evaluates was made available in December 2020 under ADAMS potential hazards, operating experience related to plant SSCs, accession no. ML20349A267. The FSP concept is being
human actions, and programmatic controls affecting the safety considered, in part, to address the need to periodically assess functions required by § 53.230, and the resulting changes in possible risk reduction measures considering technology risks to the public from operation of the facility over its operating changes, economic costs, operating experience, and new or lifetime. An FSP must include a risk-informed, performance- revised hazard information. The staff notes that similar based process to proactively identify new or revised internal or programs have been incorporated into risk-informed regulations external hazards to the facility and performance issues related issued by other federal agencies.
to plant SSCs, human actions, and programmatic controls and must consider measures to mitigate or eliminate the resulting This second iteration does not include significant changes, but risks using the criteria defined in § 53.892. The FSP must be the staff is considering possible ways to make the sections implemented and supported by a written FSP as required in shorter by reducing the level of detail in the regulations. Details
§ 53.894. could be provided by future guidance documents.
§ 53.892 Facility Safety Program Performance Criteria (a) Each licensee for an advanced nuclear plant must take measures as may be appropriate when considering potential risks to public health and safety, technology changes, economic costs, operating experience, new or revised hazard assessments, or other factors included in the FSP plan required by § 53.894. Performance objectives for design features and programmatic controls must be established such that the risks to public health and safety from an advanced nuclear plant due to normal operation or licensing basis events must not be a significant addition to other societal risks.
(1) Each licensee must assess risk reduction measures related to the release or potential release of radioactive materials in plant effluents during normal operation whenever such a release could result in a member of the public receiving an annual radiation dose in excess of 0.3 millirems from liquid effluents or 1 millirem from gaseous effluents. The assessment and risk reduction measures must maintain doses to members of the public as low as is reasonably achievable taking into account the state of technology, the economics of improvements in relation to the state of technology, operating experience, and the economics of improvements in relation to benefits to the public health and safety.
(2) Each licensee must assess potential risk reduction measures related to licensing basis events, identified hazards, or other specific contributors to the overall cumulative risk from unplanned events as follows:
(i) For new or revised hazards, plant features, or other contributors to licensing basis events with an estimated upper bound frequency above one in one thousand years, licensees must consider risk reduction measures whenever the estimated radiation dose to a member of the public exceeds 2.5 millirem and the estimated frequency weighted cumulative dose to nearby populations increases by [5 person-rem].
(ii) For new or revised hazards, plant features, or other contributors to licensing basis events with an estimated lower bound frequency below one in one thousand years, licensees must consider risk reduction measures whenever the estimated frequency weighted cumulative dose to nearby populations increases by [5 person-rem] and either the frequency of a member of the public receiving a radiation dose with the potential for immediate health effects approaches five in one hundred million years or a radiation dose with the potential to cause latent health effects approaches two in ten million years.
(iii) For new or revised hazards, plant features, or other contributors to licensing basis events with an estimated dose to a member of the public less than or equal to a threshold value used for operational flexibilities in accordance with § 53.470, licensees must consider risk reduction measures whenever changes to the estimated consequences reduce the margin to the subject threshold value by more than ten percent and the estimated frequency weighted cumulative dose to nearby populations increases by [5 person-rem].
(iv) The assessment and risk reduction measures must maintain doses to members of the public as low as is reasonably achievable taking into account the state of technology, the economics of improvements in relation to the state of technology, information available on potential hazards,
operating experience, and the economics of improvements in relation to benefits to the public health and safety.
(b) Risk reduction measures taken at advanced nuclear plants whose licenses refer to certified designs or manufacturing licenses must also follow the change control and reporting provisions of 10 CFR part 52 or subpart H of this part related to changes to standardized designs.
§ 53.894 Facility safety program plan (a) General. Each licensee must adopt and implement an FSP using a written FSP plan that, at a minimum, contains the elements in this section. This FSP plan must be approved by NRC under the process required in § 53.896.
(b) Scope. (1) Each licensee must set forth in its FSP plan a statement describing the facility or facilities covered by the plan. The description must include the facility, personnel, programmatic controls, and facility environs that influence the assessments used in assessing potential risks in accordance with subparts B and C of this part and potential reduction measures using the performance criteria in § 53.892. The scope of the program plan must consider new or revised information related to:
(i) The performance of SSCs in terms of their capability and availability to perform the required safety functions required by
§ 53.230 during normal operation and licensing basis events and assessing potential risk reduction measures using the performance criteria in § 53.892; (ii) The role of personnel in making decisions, operating plant SSCs, or otherwise supporting the safety functions required by
§ 53.230 and assessing potential risk reduction measures using the performance criteria in § 53.892; (iii) The programmatic controls required within this part or otherwise implemented by a licensee to ensure capabilities and availabilities of SSCs and personnel performing the safety functions required by § 53.230 and assessing potential risk reduction measures using the performance criteria in § 53.892;
(iv) Natural and manmade hazards with the potential to affect plant SSCs or personnel supporting the safety functions required by § 53.230 and assessing potential risk reduction measures using the performance criteria in § 53.892; and (v) Operating experience related to plant SSCs, personnel, or programmatic controls supporting the safety functions required by § 53.230 and assessing potential risk reduction measures using the performance criteria in § 53.892.
(2) The methods used to analyze the technologies identified under paragraph (f)(1)(i) of this section against the criteria provided in § 53.892.
(3) Each licensee must set forth in its FSP plan a description of its overall safety philosophy and intended safety culture to be practiced by its management, employees and contractors; and (4) Each licensee must identify the required participants in the FSP plan, which will include managers, employees, and contractors that directly support facility operations; maintain, inspect, or change plant SSCs or programmatic controls; or assess potential risk reduction measures as required by
§ 53.892.
(c) Implementation. Each licensee must describe in its FSP plan the process the licensee will use to implement and maintain its FSP. As part of the licensees implementation process, the licensee must describe roles and responsibilities of each position that has significant responsibility for implementing the FSP, including those held by employees and other persons utilizing or providing significant services as identified by the licensee pursuant to paragraph (b)(3) of this section.
(d) Facility safety program training:
(1) Each manager, employee, and contractor identified under paragraph (b)(3) of this section will be trained on the licensees FSP.
(2) Each licensee must establish and describe in its FSP plan the licensees facility safety program training plan. An FSP training plan must set forth the procedures by which managers, employees, and contractors identified under paragraph (b)(3) of
this section will be trained on the licensees FSP. An FSP training plan must help ensure that all personnel who are responsible for implementing and supporting the FSP understand the goals of the program, are familiar with the elements of the program, and have the requisite knowledge and skills to fulfill their responsibilities under the program.
(3) For each position identified pursuant to paragraph (b)(3) of this section, the training plan must describe the frequency and content of the FSP training that the position receives.
(4) Training under this subpart F may include, but is not limited to, classroom, computer-based, or correspondence training.
(5) The licensee must keep a record of all training conducted under this part and update that record as necessary. The FSP training plan must set forth the process used to maintain and update the necessary training records required by this part.
(6) The FSP training plan must set forth the process used by the licensee to ensure that it is complying with the training requirements set forth in the training plan.
(e) Risk-informed hazard management program. Each licensee must establish a risk-informed hazard management program as part of the licensees FSP. The risk-informed hazard management program must be fully described in the FSP plan. The risk-informed hazard management program must establish:
(1) The processes or procedures used in the risk-informed hazard analysis to identify internal and external hazards having the potential to increase the frequency or consequences of radiological releases from normal operation or licensing basis events; (2) The processes or procedures used in the risk-informed hazard analysis to analyze identified hazards and support assessments against the criteria provided in § 53.892; (3) The methods used to identify and implement actions that mitigate or eliminate hazards based on assessments against the criteria provided in § 53.892.
(4) The methods used to ensure changes to the facility design or operations do not adversely affect measures in place to mitigate or eliminate hazards or that such changes have been assessed pursuant to the appropriate change control and have been incorporated into models used for assessments against the criteria provided in § 53.892.
(5) The methods used to maintain records of identified hazards and risks and the mitigation or elimination of the identified hazards and risks throughout the life of the facility.
(6) The position title(s) of the individual(s) responsible for administering the risk-informed hazard management program.
(f) Technology assessment program. Each licensee must establish a technology assessment program as part of the licensees FSP. The technology assessment program must be fully described in the FSP plan. The technology assessment program must establish:
(1) The methods used to identify and analyze current, new, or novel technologies that will mitigate or eliminate internal or external hazards and resulting risks from the release of radioactive materials from a facility during normal operations or licensing basis events; (2) The methods used to analyze the technologies identified under paragraph (f)(1) of this section against the criteria provided in § 53.892.
(3) The methods used to identify and implement actions related to technologies identified under paragraph (f)(1) of this section based on assessments against the criteria provided in
§ 53.892.
(4) The methods used to maintain records of technology assessments throughout the life of the facility.
(5) The position title(s) of the individual(s) responsible for administering the technology assessment program.
(g) Internal facility safety program assessment. (1) The licensee must describe in the FSP plan methods to annually confirm:
(i) The FSP is fully implemented and effective;
(ii) The licensees overall safety philosophy and intended safety culture are being implemented and effective; (iii) The facility safety program training program is implemented and effective; (iv) The facility continues to meet the performance criteria set forth in § 53.230 and effectively consider risk reduction measures using the performance criteria set forth in § 53.892.
(2) As part of its FSP plan, the licensee must describe the processes used to:
(i) Conduct internal FSP assessments; (ii) Internally report the findings of the internal FSP assessments to a management level so that the required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are provided; (iii) Develop, track, and review recommendations as a result of the internal FSP assessments; (iv) Develop improvement plans based on the internal FSP assessments; and (v) Manage revisions and updates to the FSP plan based on the internal facility safety program assessments.
§ 53.896 Review, Approval, and Retention of Facility safety Program Plans (a) Initial Filing. Each applicant for a license under this part must include its FSP plan as part of the application.
(b) Approval. The NRC will review the FSP plan to determine if the elements prescribed in this part are sufficiently addressed in the applicants submission. Approval of an FSP plan under this part does not constitute approval of the specific actions the licensee will implement under its FSP plan pursuant to § 53.894 and must not be construed as establishing an NRC standard regarding those specific actions.
(c) Review of amendments. Need to work out nature of reviews, notices, opportunities for hearing, etc. on amendments to the FSP plan. (CRITERIA X OK report w/in x days, NRC respond
within y days). CRITERIA Y follow process in § 53.xyz (license amendment)
§ 53.900 Procedures and Guidelines This section addresses requirements for developing, (a) Each licensee under this part must have a program for implementing, and maintaining procedures (e.g., operations and developing, implementing, and maintaining an integrated set of emergency operating procedures) and guidelines (e.g., accident procedures, guidelines, and related supporting activities to management guidelines). The program requirements include support normal operations and responding to possible an integration of such procedures as was performed as an unplanned events. industry response to the accident at Fukushima Daichi.
(b) The program required by paragraph (a) of this section must include but is not necessarily limited to development, A topic for discussion is whether this requirement would be implementation, maintenance, and supporting activities of the better incorporated into sections related to staffing and the following procedures and guidelines: human-system interfaces.
(1) Plant operations (2) Maintenance under § 53.730 (3) Program requirements under this subpart (4) Emergency operating procedures if human intervention is needed to respond to licensing basis events with a frequency greater than one in ten thousand years accounting for uncertainties (5) Accident management guidelines if human intervention is needed to respond to licensing basis events with a frequency less than one in ten thousand years accounting for uncertainties (c) [Additional provisions may be added if needed]