ML22020A001
| ML22020A001 | |
| Person / Time | |
|---|---|
| Issue date: | 02/01/2022 |
| From: | O'Driscoll J NRC/NMSS/DREFS/RRPB |
| To: | |
| James O'Driscoll | |
| Shared Package | |
| ML22020A000 | List: |
| References | |
| 10 CFR Part 50, 10 CFR Part 52, NRC-2009-0196, Part 50/52, RIN 3150-AI66 | |
| Download: ML22020A001 (133) | |
Text
ADAMS Accession No. ML22020A001 ACRS Subcommittee Briefing:
Content of Proposed Rulemaking to Align Licensing Processes and Incorporate Lessons Learned from New Reactor Licensing February 1, 2022 1
Todays Meeting
- Provide an update on the effort since the last ACRS meeting on this rulemaking (meeting transcript and slides: ADAMS Accession No. ML21075A211)
- Walk through the content of the proposed rule, including proposed rule language
- Discuss the estimates of costs and savings
- Provide an update on next steps and the rulemaking schedule
- Receive ACRS members perspectives 2
OPENING REMARKS Vicki Bier - Subcommittee Chairman Rob Taylor - Deputy Office Director for New Reactors in the Office of Nuclear Reactor Regulation 3
NRC STAFF PRESENTATION 4
NRC Staff Presenters Jim ODriscoll, NMSS Rulemaking Project Manager Omid Tabatabai, NRR Senior Project Manager 5
Purpose of the Rulemaking
- Implement Commission direction in SRM-SECY-15-0002, Proposed Updates of Licensing Policies, Rules, and Guidance for Future New Reactor Applications, to:
- Align Parts 50 and 52 reactor licensing processes
- Improve clarity
- Incorporate lessons learned in recent licensing proceedings
- Reduce unnecessary burden on applicants and staff 6
Rulemaking Process Identify Regulatory Proposed Final Rule need for Basis Rule rulemaking
- Described in
- Analyze
- Proposed rule *Final rule text SECY-15-0002 alternatives for text resolution
- Commissions
- Public meeting direction in
- Public meeting
- 75-day public SRM-SECY
- 75-day public comment period 0002 comment period Opportunities for public participation 7
Staffs Milestones of Rulemaking Activities October 1, 2018 Started scoping and outreach January 15, 2019 Held public meeting July 11, 2019 Internal alignment on scope of RB Issuance of Commission Information August 27, 2019 Paper SECY-19-0084 September 20, 2019 Held ACRS meeting 8
Staffs Milestones of Rulemaking Activities (contd)
November 21, 2019 Held public meeting February 2020 First draft of RB inputs completed April 29, 2020 Held public meeting January 2021 Published RB and FRN March 2021 Held public meetings 9
Staffs Milestones of Rulemaking Activities (contd)
May 2021 End of public comment period May 2021 First drafts of proposed rule technical inputs completed Final drafts of proposed rule technical November 2021 inputs completed December 2021 Commenced management review 10
Next Steps May 2022 Complete management concurrence May 2022 Forward the proposed rule to the Commission for approval March 2024 Forward the final rule to the Commission for approval 11
Federal Register Notice
- A proposed rule FRN supports public participation in the rulemaking process
- NRCs proposed rules also contain a preliminary cost/benefit analysis of the proposed changes
- Organization of the proposed rule:
- Preamble (explanation of proposed changes)
- Proposed rule language 12
Preamble TABLE OF CONTENTS:
I. Obtaining Information and Submitting Comments A. Obtaining Information B. Submitting Comments II. Background III. Discussion A. Applying the Severe Accident Policy Statement to New Part 50 License Applications B. Probabilistic Risk Assessment Requirements C. Three Mile Island Requirements D. Description of Fire Protection Design Features and Fire Protection Plans E. Operator Licensing F. Physical Security and Fitness-for-Duty Requirements G. Emergency Planning H. The Part 52 Licensing Process I. Environmental J. Applicability of Other Processes to the Part 52 Licensing Process K. Miscellaneous Topics IV. Specific Requests for Comments V. Section-by-Section Analysis VI. Regulatory Flexibility Certification VII. Regulatory Analysis VIII. Backfitting and Issue Finality A. Current and Future Applicants B. Existing Design Certifications C. Existing Licenses D. Draft Regulatory Guidance IX. Cumulative Effects of Regulation X. Plain Writing XI. National Environmental Policy Act XII. Paperwork Reduction Act Statement XIII. Criminal Penalties XIV. Voluntary Consensus Standards XV. Availability of Guidance XVI. Public Meeting XVII. Availability of Documents 13
Scope of the Proposed Rule
- Number of technical areas: 11
- Number of items in scope: 61
- Items with rulemaking recommendation: 60
- Number of items with rulemaking and guidance development or revision: 18
- Number of guidance documents with rule: 14
- Number of 10 CFR Parts affected by rulemaking: 9 14
Alignment of Parts 50 and 52
- The proposed rule addresses four areas where the NRCs policies and direction for new reactors have resulted it requirements and guidance for Part 52 applicants only:
- Application of Severe Accident Policy Statement (1)
- Probabilistic Risk Assessment Requirements (3)
- Three Mile Island Requirements (1)
- Fire Protection Design Features and Plans (1) 15
Lessons Learned from Recent Experience
- The proposed rule covers topics for which the NRCs recent experience with new light water reactor licensing has resulted in lessons learned Operator Physical Fitness Emergency Licensing Security For Duty Planning (5) (2) (4) (7)
Part 52 Applicability of Environmental Other Processes Miscellaneous Licensing Topic to the 10 CFR Topics Process Part 52 Process (1) (9)
(21) (5) 16
Relationship to Non-LWRs
- Cross-cutting item
- The item was added in response to public comments on the regulatory basis
- The goal of the discussion and proposed changes is to explain how this rulemaking activity fits with other licensing process efforts and rulemakings that relate to non-light water technology 17
Topics of Interest
- Part 52 Licensing Process (slides 19-39)
- Severe Accidents (slide 40)
- Probabilistic Risk Assessment Requirements (slide 41-44)
- Three Mile Island Requirements (slide 45)
- Emergency Planning (slide 48-52)
- Operators Licenses (slide 53-57)
- Miscellaneous Topics (slide 58-61) 18
Part 52 Licensing Process
- The NRC plans on providing changes and/or guidance to the following areas for Part 52 licensing:
- DC Renewal (III.H.1)
- Change Process (III.H.2)
- Design Scope and Standardization (III.H.3)
- References to SDAs (III.H.4)
- Content of Applications (III.H.5) 19
DC Renewal
- Changes: Remove 15-year duration of DCs and renewal requirements
- Affected regulations:
- §52.55, Duration of certification
- §52.57, Application for renewal
- §52.59, Criteria for renewal
- §52.61, Duration of renewal
- Part 52 Appendices A, D, and E
- Reduces unnecessary burden on NRC and industry
- Public comments:
- Supportive; resulted in applying the changes to SDAs and MLs
- Expired DCs removed from the regulations
- Cost/benefit: Cost-beneficial 20
Change Process
- Changes:
- Process for Making Changes to the Plant-Specific Design Control Document Organization and Section Numbering
- Include§50.59(c) Provisions in the Part 52 Change Process
- Approval Process for Changes While the Plant Is Being Constructed
- Standard Design Approval Variance Process
- Generic Standard Design Approval Change Process
- Referencing Manufacturing Licenses and Standard Design Approvals While They Are Under Review
- No change recommended:
- Move the§50.59-Like Change Process from Part 52 Appendices to Part 52, Subpart B 21
PS-DCD Organization and Numbering
- Change: Allow COL applicants to reorganize their PS-DCD without an exemption
- Affected regulations:
- §IV.A.2 of each DC Rule
- Reduces unnecessary burden on NRC and industry
- Public comments:
- Supportive; requested more permission to change Tier 1 information without prior staff approval
- No further changes to the regulations are proposed to address comments
- Cost/benefit: Rulemaking; qualitative 22
Include §50.59(c) Provisions in the Part 52 Change Process
- Change: Add provisions similar to§50.59(c)(4) to the change processes in each DC Rule
- Affected regulations:
- §VIII.B.5.a of each DC Rule
- Improves the alignment of the change processes in Part 50 and Part 52
- Public comments:
- Supportive; requested more permission to change Tier 1 information without prior staff approval
- No further changes to the regulations are proposed to address comments
- Cost/benefit: Rulemaking; qualitative 23
Approval Process for Changes While the Plant Is Being Constructed
- Change: Add provisions to allow licensees to continue construction while an LAR for a change is under review by the NRC
- Affected regulations:
- §VIII.B.5 of each DC Rule
- Eliminates construction delays related to NRC staff review of LARs
- Public comments:
- Requested removal of 45-day requirement to submit an LAR
- No further changes to the regulations are proposed to address comments
- Cost/benefit: Rulemaking; qualitative 24
Standard Design Approval Variance Process
- Change: Establish a new process to govern the request, review, and approval of changes to SDAs in an application
- Affected regulations:
- §52.93, Exemptions and variances
- §52.145, Finality of standard design approvals; information requests
- Establishes a clear regulatory process to make changes to a referenced SDA
- Public comments:
- New item; added in response to public comments on the regulatory basis
- Cost/benefit: Rulemaking; qualitative 25
Generic Standard Design Approval Change Process
- Affected regulations:
- §2.100, Scope of subpart
- §2.101, Filing of application
- §2.110, Filing and administrative action on submittals for standard design approval or early review of site suitability issues
- §52.3, Written communications
- §52.145, Finality of standard design approvals; information requests
- Establishes a clear regulatory process to make generic SDA changes
- Public comments:
- New item; added in response to public comments on the regulatory basis
- Cost/benefit: Rulemaking; qualitative 26
Referencing Manufacturing Licenses and Standard Design Approvals While They Are Under Review
- Affected regulations:
- §52.173, Referencing a manufacturing license application
- Regulatory Guide 1.206, Applications for Nuclear Power Plants
- Afford the same flexibility for MLs and SDAs as that currently afforded DCs and ESPs in the same situation
- Public comments:
- None; not in the regulatory basis (new item)
- Cost/benefit: Rulemaking; qualitative 27
Move the § 50.59-Like Change Process from Part 52 Appendices to Part 52, Subpart B
- Change: None
- Affected regulations: None
- Reduce the size and complexity of NRCs regulations
- Public comments:
- There were no public comments on this item
- Cost/benefit: N/A 28
Design Scope and Standardization
- Changes:
- Add Definitions of Tier 1, Tier 2, and Tier 2*
- Added Definition of Essentially Complete Design
- Restrictions on Changes to a DC or COL Referencing a DC for Reasons of Standardization
- Design Certification Rule Section IX 29
Add Definitions of Tier 1, Tier 2, and Tier 2*
- Change: Add standardized definitions of tiered information to Part 52
- Affected regulations:
- §52.1, Definitions
- Add clarity and efficiency; reduce scope of information designated as Tier 1
- Public comments:
- Requested the item apply only to DC applications
- Cost/benefit: Cost-beneficial 30
Clarify the Phrase Essentially Complete Design
- Change: Add standardized definition of essentially complete design to Part 52
- Affected regulations:
- §52.1, Definitions
- Add clarity and efficiency; reduce scope of information needed for review
- Public comments:
- Review September 24, 2021 NEI letter to NRC
- Cost/benefit: Cost-beneficial 31
Restrictions on Changes to a DC or COL Referencing a DC for Reasons of Standardization
- Change: Remove requirements to justify proposed exemptions on the basis of standardization
- Affected regulations:
- §52.63(b)(1), Finality of standard design certifications
- §52.93(c),
- §52.171(b)(2), Finality of manufacturing licenses; information requests
- Reduce unnecessary burden; simplify requirements to justify changes
- Public comments:
- None
- Cost/benefit: Cost-beneficial 32
Design Certification Rule Section IX
- Affected regulations:
- §52.99, Inspection during construction
- Part 52 Appendix D,Section IX
- Eliminates redundant regulations
- Public comments:
- ITAAC closure reporting during construction; the staff agrees and proposes an additional change as a result of the comment
- Cost/benefit: Rulemaking; qualitative 33
References to Standard Design Approvals
- Change: Revise to clarify that more than one SDA can be referenced in a license application
- Affected regulations:
- §52.73(a), Relationship to other subparts
- §52.79(c), Contents of applications; technical information in final safety analysis report
- §52.133(a), Relationship to other subparts
- §52.153(b), Relationship to other subparts
- Clarifies regulations
- Public comments:
- None
- Cost/benefit: Rulemaking; qualitative 34
Content of Applications
- Changes:
- Modify Requirements to Evaluate Conformance with the Standard Review Plan
- Align Requirements for Timely Completion of Construction
- Clarify Applicable Regulatory Parts for Certified Designs
- Clarify the Requirements for Environmental Qualification Program for Manufacturing Licenses 35
Modify Requirements to Evaluate Conformance with the Standard Review Plan
- Affected regulations:
- §50.34(h), Contents of applications; technical information
- §52.17(a)(1)(xii), Contents of applications; technical information
- §52.47(a)(9), Contents of applications; technical information
- §52.79(a)(41), Contents of applications; technical information in final safety analysis report
- §52.137(a)(9), Contents of applications; technical information
- §52.157(f)(30), Contents of applications; technical information in final safety analysis report
- Reduces unnecessary burden on license applicants
- Public comments:
- None
- Cost/benefit: Cost-beneficial to industry 36
Align Requirements for Timely Completion of Construction
- Change: Amend § 50.100 to resolve an inconsistency with respect to § 50.55 on timely completion of construction
- Affected regulations:
- §50.100, Revocation, suspension, modification of licenses, permits, and approvals for cause
- Clarifies the regulations
- Public comments:
- None
- Cost/benefit: Rulemaking; qualitative 37
Clarify Applicable Regulatory Parts for Certified Designs
- Change: Add a reference to part 52 as one of the applicable regulatory parts for appendices D and E
- Affected regulations:
- Part 52 Appendices D and E,Section V
- Clarifies the regulations
- Public comments:
- None
- Cost/benefit: Rulemaking; qualitative 38
Clarify the Requirements for Environmental Qualification Program for Manufacturing Licenses
- Change: Add a conforming change to the regulations
- Affected regulations:
- §52.157, Contents of applications; technical information in final safety analysis report
- Clarifies the regulations
- Public comments:
- None; new item
- Cost/benefit: Rulemaking; qualitative 39
Severe Accident Treatment Requirements
- Section III.A of the FRN
- Change: Add requirements for future Part 50 applicants to address severe accidents; require an evaluation on how a proposed change could affect ex-vessel severe accidents
- Affected regulations:
- §50.34, Contents of applications; technical information
- §50.59, Changes, tests, and experiments
- Aligns severe accident requirements in Part 50 and Part 52; aligns change process in Part 50 and Part 52 for this issue
- Public comments:
- Five comments; comments resulted in changes to staffs recommendation
- Cost/benefit: Quantitative; increased cost for industry, decreased cost for NRC 40
Probabilistic Risk Assessment Requirements
- Changes:
- Require PRA in the design and operating phases for Part 50 plants (III.B.1)
- Allow greater use of risk-informed categorization of structures, systems, and components (III.B.2)
- Require maintaining and simplify the schedule for upgrading the plant-specific PRA (III.B.3) 41
Use of Probabilistic Risk Assessment in Design
- Change: Extend the current PRA requirements in Part 52 to apply to Part 50 power reactor license applicants
- Affected regulations:
- §50.34(a), Preliminary safety analysis report
- §50.34(b), Final safety analysis report
- Aligns Parts 50 and 52 on the use of PRA in the design of the facility and ensures that similar risk information is supplied in applications for new power reactor CPs or OLs under Part 50
- Public comments:
- Ten comments; expressed concern over changes and need for clarification on how to meet requirements
- In response to comments, NRC changed the cost model to reflect the significant effort required to complete an upgrade prior to loading fuel
- Cost/benefit: Development of PRA, rulemaking; qualitative 42
Risk-Informed Categorization of Structures, Systems and Components
- Change: Allow DC applicants, power reactor CP holders, and COL holders to risk-inform the categorization of structures, systems, and components
- Affected regulations:
- §50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors
- Public comments:
- Five comments; questioned the benefit of the rulemaking and expressed concern of undue burden
- No change to recommendations in response to comments
- Cost/benefit: Rulemaking; qualitative 43
Maintaining and Upgrading the Plant-Specific PRA
- Changes:
- Make regulation applicable to those license holders under Part 50 that are required to develop a PRA
- Simplify the schedule for upgrading the PRA
- Affected regulations:
- §50.71, Maintenance of records, making of reports
- Promote a more stable and equitable regulatory environment
- Public comments:
- Four comments; generally supportive of the changes and recommended some clarifications in regulatory basis
- No change to recommendations in response to comments
- Cost/benefit: Cost for Maintenance of PRA for part 50 plants, rulemaking; some reduced cost from fewer part 52 exemption requests 44
Three Mile Island Requirements
- Section III.C of the FRN
- Changes:
- Apply TMI requirements to new power reactor applications submitted under Part 50
- Delete requirements that are included in other regulations or are no longer needed or applicable
- Affected regulations:
- §50.34(f), Additional TMI related requirements
- Aligns Parts 50 and 52 on requirements related to the TMI accident and ensures consistency in new reactor licensing reviews
- Public comments:
- Three comments; generally supportive, additional revisions recommended
- Staff recommended several changes based on comments
- Cost/benefit: Reduced application requirements, rulemaking; quantitative 45
Questions 46
Lunch Break 11:30 AM - 1:00 PM 47
Emergency Planning
- Changes:
- Emergency Plan Change Process (III.G.1)
- Emergency Preparedness Exercises (III.G.2)
- Significant Impediments to Development of Emergency Plans (III.G.3)
- Offsite Contacts, Arrangements, and Certifications (III.G.4) 48
Emergency Plan Change Process
- Change: Align the introductory text of
§50.54 with the text in §50.54(q)(2)
- Affected regulations:
- §50.54, Conditions of licenses
- Eliminates uncertainty and clarifies intent of regulations
- No public comments
- Cost/benefit: Cost-beneficial; assessed qualitatively 49
Emergency Preparedness Exercises
- Changes: Revise paragraphs IV.F.2.a.(i) through (iii) and j of Part 50 Appendix E to provide clarifications and revisions for subsequent exercises and exercise planning
- Affected regulations:
- Part 50 Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities
- Aligns the requirements for Part 52 licensees and Part 50 licensees
- No public comments
- Cost/benefit: Cost-beneficial 50
Significant Impediments to Development of Emergency Plans
- Changes: Distinguish the siting requirements from EP considerations in §52.17 and clarify that NRC consultation with FEMA is for the purpose of evaluating whether significant impediments identified by the applicant can be mitigated or eliminated by the measures proposed by the applicant
- Affected regulations:
- §52.18, Standards for review of applications
- Eliminate uncertainty and clarify intent of regulations
- Public comments:
- Supportive and recommended a flexible approach
- Further changes to RG 4.7 in response to the comment
- Cost/benefit: Cost-beneficial 51
Offsite Contacts, Arrangements, and Certifications
- Change: Clarify what information is required regarding contacts, arrangements, and certifications with Federal, State, and local governmental authorities in the SSAR
- Affected regulations:
- § 52.17, Contents of applications; technical information
- Clearly defines the differing requirements for ESP applications
- Public comments:
- Editorial correction to regulatory basis
- No change to recommendations in response to this comment
- Cost/benefit: Cost-beneficial 52
Operators Licensing
- Changes:
- Criteria for Simulation Facilities (III.E.1)
- Plant Walkthrough (III.E.2)
- Continuing Training for Operator License Applicants (III.E.3)
- Waiver of Examination and Test Requirements (III.E.4) 53
Operators Licenses - Criteria for Simulation Facilities
- Changes:
- Amend criteria that a plant-referenced simulator must meet when it will be used to complete the control manipulations required by
§ 55.31(a)(5) by operator license applicants at nuclear power plants that are under construction
- Amend the definitions of plant-referenced simulator and reference plant in § 55.4, Definitions, to clarify that these terms are also applicable to simulators that model nuclear power plants that are under construction
- Affected regulations:
- § 55.46(c), Plant-referenced simulators
- § 55.4, Definitions
- Reduces administrative burden
- No public comments
- Cost/benefit: Cost-beneficial 54
Plant Walkthrough
- Change: Amend the plant walkthrough requirement to give facility licensees of new reactors under construction the option of using suitable alternatives to in-plant testing while the plant is under construction
- Affected regulations:
- §55.45, Operating tests
- Promotes a more efficient and effective operator licensing process at cold plants
- No public comments
- Cost/benefit: Cost-beneficial 55
Continuing Training for Operator License Applicants
- Change: Establish a new requirement for facility licensees at cold plants to maintain the knowledge, skills, and abilities of operator license applicants who have successfully completed the NRC initial licensing examination
- Affected regulations:
- §55.31, How to apply
- Promotes a more efficient and effective operator licensing process at cold plants
- Public comments:
- Recommended more specific guidance
- NRC thus is proposing a new Section H, Continuing Training for Applicants at New Reactors under Construction, in NUREG-1021
- Cost/benefit: Rulemaking; qualitative 56
Waiver of Examination and Test Requirements
- Change: Include provisions for licensing of operators at subsequent new units at multiunit sites
- Affected regulations:
- §55.47, Waiver of examination and test requirements
- Promotes a more efficient and effective operator licensing process at cold plants
- No public comments (not in regulatory basis)
- Cost/benefit: Cost-beneficial 57
Miscellaneous Topics
- The NRC is also proposing changes in several other areas of the proposed rulemaking as follows:
- Status of ITAAC Completion (III.K.7)
- Reporting Requirements for:
- ECCS Errors (III.K.4)
- Completion of Construction (III.K.8)
- Conditions of Licenses (III.K.9)
- Discontinuing Priority Ranking Model for GIs (III.K.6) 58
Status of ITAAC Completion
- Change: Revise the language regarding acceptance criteria from have been met to are met
- Affected regulations:
- §52.97(a)(2)
- Consistency with the requirements in the AEA and
§52.103(g), which state that the acceptance criteria in the COL are met
- No public comments
- Cost/benefit: Negligible cost impact 59
Reporting Requirements for ECCS Errors
- Change: Amend §50.46(a)(3)(i) and (iii) to relax certain reporting requirements related to those SDAs and DCs that are not referenced in any application for the construction or operation of a reactor
- Affected regulations:
- §50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors
- Defer the annual reporting until the standard design is referenced in a license or license application
- No public comments
- Cost/benefit: Cost-beneficial 60
Reporting Requirements for Completion of Construction
- Change: Require that all future Part 50 power reactor licensees and Part 52 COL holders promptly notify the NRC of the successful completion of power ascension testing
- Affected regulations:
- §50.71(i)
- Enables NRC to begin assessing Part 171 annual fees without issuance of license conditions
- No public comments
- Cost/benefit: Cost-beneficial 61
Conditions of License
- Change: Clarify the applicability of conditions of operating licenses for non-power production and utilization facilities
- Affected regulations:
- §50.54, Conditions of licenses
- Clarify the requirements of applicability
- No public comments; not discussed in regulatory basis
- Cost-beneficial 62
Discontinuing Priority Ranking Model for GIs
- Change: Amend regulations to reflect discontinuance of the priority ranking model to identify significant generic issues (GIs) in favor of a risk-informed method to identify significant GIs that an applicant should address in its submittal
- Affected regulations:
- §§ 52.47(a)(21), 52.79(a)(20), 52.137(a)(21), and 52.157(f)(28)
- Must propose technical resolutions of all GIs identified since July 21, 1999; unresolved safety issues; and medium- and high-priority generic safety issues identified before July 21, 1999, that are relevant to design
- Public comments:
- New guidance (beyond RG 1.174) needed for applicants/licensees to implement
- No change to recommendations in response to this comment
- Cost/benefit: Cost-beneficial; assessed qualitatively 63
Estimates of Costs and Savings
- Total net averted costs to industry and the NRC between $16.1 million and $25.5 million
- To account for sensitivity to plant-specific conditions, the NRC staff performed an uncertainty analysis, which found that the chance of net averted costs is greater than 99%
- Rulemaking would yield nonquantifiable benefits as well (regulatory efficiency, public confidence) 64
Questions 65
Recap and Next Steps
- Complete concurrence on draft proposed rule
- Submit the proposed rule to the Commission
- Plan for additional public meeting(s) during the proposed rule phase 66
Rulemaking Schedule Submit proposed Issue final rule to the rule Commission May 2022 October 2024 67
Contact Information Jim ODriscoll, Project Manager Division of Rulemaking, Environmental, & Financial Support Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Email: James.ODriscoll@nrc.gov Phone: 301-415-1325 Omid Tabatabai, Senior Project Manager Division of New Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Email: Omid.Tabatabai@nrc.gov Phone: 301-415-6616 68
How to Stay Informed and Involved
- The meeting summary will be posted soon
- Search regulations.gov on the docket ID NRC-2009-0196
- Press the Subscribe button to receive alerts about updates to the docket 69
SUPPORTING INFORMATION 70
Abbreviations ACRS Advisory Committee on Reactor LWR Light-Water Reactor Safeguards ML Manufacturing License ADAMS Agencywide Documents Access and NEI Nuclear Energy Institute Management System NMSS Office of Nuclear Material Safety and AEA Atomic Energy Act of 1954, as amended Safeguards CFR Code of Federal Regulations NRC Nuclear Regulatory Commission COL Combined License NRR Office of Nuclear Reactor Regulation CP Construction Permit OL Operating License DC Design Certification PRA Probabilistic Risk Assessment ECCS Emergency Core Cooling System PS-DCD Plant-Specific Design Certification EP Emergency Planning Document ESP Early Site Permit RB Regulatory Basis FEMA Federal Emergency Management Agency RG Regulatory Guide FFD Fitness For Duty SDA Standard Design Approval FRN Federal Register Notice SECY Office of the Secretary FSAR Final Safety Analysis Report SRM Staff Requirements Memorandum GI Generic Issue SRP Standard Review Plan ITAAC Inspections, Tests, Analyses, and SSAR Site Safety Analysis Report Acceptance Criteria SSC Structure, System, and Component LAR License Amendment Request TMI Three Mile Island 71
ECCS Acceptance Criteria-Rule Language 72
DC Renewal- Rule Language
§ 52.55 Duration of certification. Referencing a design certification application.
(a) [Reserved]Except as provided in paragraph (b) of this section, a standard design certification issued under this subpart is valid for 15 years from the date of issuance.
(b) [Reserved]A standard design certification continues to be valid beyond the date of expiration in any proceeding on an application for a combined license or an operating license that references the standard design certification and is docketed either before the date of expiration of the certification, or, if a timely application for renewal of the certification has been filed, before the Commission has determined whether to renew the certification. A design certification also continues to be valid beyond the date of expiration in any hearing held under
§52.103 before operation begins under a combined license that references the design certification.
(c) An applicant for a construction permit or a combined license may, at its own risk, reference in its application a design for which a design certification application has been docketed but not granted
§ 52.57 [Reserved]Application for renewal.
§ 52.59 [Reserved]Criteria for renewal.
§ 52.61 [Reserved]Duration of renewal.
Each renewal of certification for a standard design will be for not less than 10, nor more than 15 years.
73
DC Renewal- Rule Language (contd)
- Part 52 DC appendices A, D and E (example is Appendix A):
VII. [Reserved]Duration of This Appendix This appendix may be referenced for a period of 15 years from June 11, 1997, except as provided for in 10 CFR 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
VIII. Processes for Changes and Departures A. * *
- B. Tier 2 information.
- 1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).
- 2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
- 3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under §§ 52.55 or 52.61, unless:
- a. * *
- 74
SDA and ML Renewal- Rule Language Subpart E Standard Design Approvals
§52.147 [Reserved]Duration of design approval Subpart F Manufacturing Licenses
§52.173 Duration of manufacturing license. Referencing a manufacturing license application.
A manufacturing license issued under this subpart may be valid for not less than 5, nor more than 15 years from the date of issuance. A holder of a manufacturing license may not initiate the manufacture of a reactor less than 3 years before the expiration of the license even though a timely application for renewal has been docketed with the NRC. Upon expiration of the manufacturing license, the manufacture of any uncompleted reactors must cease unless a timely application for renewal has been docketed with the NRC. An applicant for a construction permit or a combined license may, at its own risk, reference in its application a design for which a manufacturing license application has been docketed but not granted.
§52.177 [Reserved]Application for renewal.
§52.179 [Reserved]Criteria for renewal.
§52.181 [Reserved]Duration of renewal.
75
Removal of Expired DCs-Rule Language Appendix B to Part 52[Reserved]Design Certification Rule for the System 80 + Design Appendix C to Part 52[Reserved]Design Certification Rule for the AP600 Design 76
PS-DCD Organization and Numbering
/Relocate Requirements IV. Additional Requirements and Restrictions A. An applicant for a combined license that wishes to reference this appendix shall, in addition to complying with the requirements of 10 CFR 52.77, 52.79, and 52.80, comply with the following requirements:
- 1. Incorporate by reference, as part of its application, this appendix;
- 2. Include, as part of its application:
- a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the U.S. ABWR design, as modified and supplemented by the applicant's exemptions and departures;
- b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
- c. Plant-specific technical specifications, consisting of the generic and site-specific technical specifications, that are required by 10 CFR 50.36 and 50.36a;
- d. Information demonstrating that the site characteristics fall within the site parameters and that the compliance with the site parameters and interface requirements have been met;
- e. Information that addresses the COL action items; and
- f. Information required by 10 CFR 52.47 that is not within the scope of this appendix.
77
Include § 50.59(c) Provisions in the Part 52 Change Process VIII. Processes for Changes and Departures B. Tier 2 information.
5
- a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the TS, or requires a license amendment under paragraph B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD. The provisions in B.5.b of this section do not apply to proposed departures when applicable regulations establish more specific criteria for accomplishing such departures.
78
Approval Process for Changes While the Plant Is Being Constructed VIII. Processes for Changes and Departures B. Tier 2 information.
5* *
- h.
(1) Notwithstanding the provisions of B.5.a of this section, during the period when the plant is being constructed prior to a Commission finding under 10 CFR 52.103(g), a licensee may construct an SSC in accordance with a proposed departure from Tier 2 or Tier 2* information, excluding the Tier 2* departures covered under B.6.b of this section, of the plant-specific DCD for a COL covered by 10 CFR 52.98(c)(1), without first obtaining a license amendment provided:
- i. A licensee must submit the request for a license amendment required to authorize the departure from Tier 2 or Tier 2* of the plant-specific DCD within 45 days after the licensee approves the design of an SSC as changed and begins construction of the SSC; ii. A licensee is not permitted to begin construction in accordance with a design that departs from Tier 2 or Tier 2* unless the SSC under construction is located within the restricted area defined in 10 CFR Part 20 and described in the FSAR, as updated; and iii. If the NRC does not approve a request for a license amendment as submitted, the licensee is obligated to construct the facility in accord in the FSAR, as updated, including the plant-specific DCD.
(2) Notwithstanding the provisions of B.5.a of this section regarding the license amendment required under B.5.b of this section, after 10 CFR 52.103(g) finding by the Commission, a prior approval from NRC is required before implementing a proposed departure from Tier 2 information under B.5.b of this section. A prior approval is required for departures from Tier 2* in accordance with B.5.a and B.6.b of this section.
- * *
- 79
Standard Design Approval Variance Process Subpart C Combined Licenses
§ 52.93 Exemptions and variances.
(c) An applicant for a construction permit, combined license, or manufacturing license who has filed an application referencing a standard design approval issued under subpart E of this part may include in the application a request for a variance from one or more design characteristics, site parameters, terms and conditions, or approved design of the reactor or major portions thereof. In determining whether to grant the variance, the NRC staff shall apply the same technically relevant criteria as were applicable to the application for the original or amended standard design approval. Once a construction permit, combined license, or manufacturing license is issued, a referenced standard design approval is subsumed, to the extent referenced, into the construction permit, combined license, or manufacturing license 80
Standard Design Approval Variance Process (contd)
Subpart E Standard Design Approvals
§ 52.145 Finality of standard design approvals; information requests.
(c) Upon issuance of a construction permit, operating license, combined license, or manufacturing license, any referenced standard design approval is subsumed, to the extent referenced, into the construction permit, operating license, combined license, or manufacturing license.
(d) An applicant for a construction permit, operating license, combined license, or manufacturing license referencing one or more standard design approvals may include in its application a request for a variance from one or more from one or more provisions of the standard design approval, or from the associated final safety analysis report. In determining whether to grant the variance, the NRC staff shall apply the same technically relevant criteria applicable to the application for the original standard design approval.
81
Generic Standard Design Approval Change Process Subpart A - Procedure for Issuance, Amendment, Transfer, or Renewal of a License, and Standard Design Approval
§ 2.100 Scope of subpart.
This subpart prescribes the procedure for issuance of a license; amendment of a license at the request of the licensee; transfer and renewal of a license; and issuance of a standard design approval, and amendment of a standard design approval at the request of the standard design approval holder under subpart E of part 52 of this chapter.
§ 2.101 Filing of application.
(a)
(1) An application for a permit, a license, a license transfer, a license amendment, a license renewal, or a standard design approval, or a standard design approval amendment mustshall be filed with the Director, Office of Nuclear Reactor Regulation, or the Director, Office of Nuclear Material Safety and Safeguards, as prescribed by the applicable provisions of this chapter. A prospective applicant may confer informally with the NRC staff before filing an application.
82
Generic Standard Design Approval Change Process (contd)
§ 2.110 Filing and administrative action on submittals for standard design approval or early review of site suitability issues.
(a)
(1) A submittal for a standard design approval or standard design approval amendment under subpart E of part 52 of this chapter shall be subject to §§ 2.101(a) and 2.390 to the same extent as if it were an application for a permit or license.
(2) Except as specifically provided otherwise by the provisions of appendix Q to parts 50 of this chapter, a submittal for early review of site suitability issues under appendix Q to parts 50 of this chapter shall be subject to §§ 2.101(a)(2) through (4) to the same extent as if it were an application for a permit or license.
(b) Upon initiation of review by the NRC staff of a submittal for an early review of site suitability issues under Appendix Q of part 50 of this chapter, or for a standard design approval or standard design approval amendment under subpart E of part 52 of this chapter, the Director, Office of Nuclear Reactor Regulation, shall publish in the Federal Register a notice of receipt of the submittal, inviting comments from interested persons within 60 days of publication or other time as may be specified, for consideration by the NRC staff and ACRS in their review.
(c)
(1) Upon completion of review by the NRC staff and the ACRS of a submittal for a standard design approval or standard design approval amendment, the Director, Office of Nuclear Reactor Regulation, shall publish in the Federal Register a determination as to whether or not the design is acceptable, subject to terms and conditions as may be appropriate, and shall make available at the NRC Web site, http://www.nrc.gov, a report that analyzes the design.
83
Generic Standard Design Approval Change Process (contd)
General Provisions
§ 52.3 Written communications.
(b) * * *
(1) Applications for amendment of permits, approvals, and licenses; reports; and other communications. All written communications (including responses to: generic letters, bulletins, information notices, regulatory information summaries, inspection reports, and miscellaneous requests for additional information) that are required of holders of early site permits, standard design approvals, combined licenses, or manufacturing licenses issued under this part must be submitted as follows, except as otherwise specified in paragraphs (b)(2) through (b)(7) of this section: to the NRC's Document Control Desk (if on paper, the signed original), with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector, if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part 84
Generic Standard Design Approval Change Process (contd)
Subpart E Standard Design Approvals
§ 52.145 Finality of standard design approvals; information requests.
(e) The holder of a standard design approval may not make changes to the design of the nuclear power reactor or major portions thereof without prior NRC staff approval. The request for a change to the design must be in the form of an amendment application as specified in § 52.3 of this chapter. The application shall fully describe the changes desired, following the form prescribed for original applications insofar as it applies. The NRC staffs review of the amendment application will be guided by the applicable considerations governing the issuance of the initial approval. Upon completion of its review of the application for the amendment the NRC staff shall publish a determination in accordance with § 52.143.
85
Referencing Manufacturing Licenses and Standard Design Approvals While They Are Under Review Subpart F Manufacturing Licenses
§52.173 Duration of manufacturing license. Referencing a manufacturing license application.
A manufacturing license issued under this subpart may be valid for not less than 5, nor more than 15 years from the date of issuance. A holder of a manufacturing license may not initiate the manufacture of a reactor less than 3 years before the expiration of the license even though a timely application for renewal has been docketed with the NRC. Upon expiration of the manufacturing license, the manufacture of any uncompleted reactors must cease unless a timely application for renewal has been docketed with the NRC. An applicant for a construction permit or a combined license may, at its own risk, reference in its application a design for which a manufacturing license application has been docketed but not granted.
86
Definitions of Tier 1, Tier 2, and Tier 2*
Part 52 - Licenses, Certifications, and Approvals for Nuclear Power Plants General Provisions
§ 52.1 Definitions.
Tier 1 means, for design certifications issued after [EFFECTIVE DATE OF FINAL RULE], the qualitative and functional-level portion of the design-related information contained in the generic design control document that is approved and certified by a standard design certification. The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. For design certifications issued prior to [EFFECTIVE DATE OF FINAL RULE], see the definition of this term in the applicable appendix to this part.
Tier 2* means for design certifications issued after [EFFECTIVE DATE OF FINAL RULE], the portion of the Tier 2 information, designated as such in the generic design control document, that is subject to the change process in section VIII.B.6 of the appendices to this Part. This designation expires for some Tier 2* information under section VIII.B.6. For design certifications issued prior to [EFFECTIVE DATE OF FINAL RULE], see the definition of this term in the applicable appendix to this part.
Tier 2 means, for design certifications issued after [EFFECTIVE DATE OF FINAL RULE], the portion of the design-related information contained in the generic design control document that is approved, but not certified, by a standard design certification. Compliance with Tier 2 is required, but generic changes to, and plant-specific departures from, Tier 2 are governed by section VIII of the applicable appendix to this part. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in section VIII of the applicable appendix to this part. Regardless of these differences, an applicant or licensee must meet the requirement in section III.B of the applicable appendix to this part to reference Tier 2 when referencing Tier 1.
For design certifications issued prior to [EFFECTIVE DATE OF FINAL RULE], see the definition of this term in the applicable appendix to this part.
87
Clarify the Phrase Essentially Complete Design Part 52 - Licenses, Certifications, and Approvals for Nuclear Power Plants General Provisions
§ 52.1 Definitions.
Essentially complete design means a nuclear power plant design of adequate scope and detail to enable the Commission to reach a conclusion that the applicant has satisfied all applicable regulations associated with the design. The scope of an essentially complete design describes all structures, systems, and components required for safe operation of the plant in all modes.
Interface requirements are specified for features that are not included within the scope of the design. The level of detail is sufficient to allow resolution of all technical issues using an approach informed by the safety significance of the plants structures, systems, and components.
88
Restrictions on Changes to a DC or COL Referencing a DC for Reasons of Standardization
§ 52.63 Finality of standard design certifications.
(b)
(1) An applicant or licensee who references a design certification rule may request an exemption from one or more elements of the certification information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 52.7. In addition to the factors listed in § 52.7, the Commission shall consider whether the special circumstances that § 52.7 requires to be present outweigh any decrease in safety that may result from the reduction in standardization caused by the exemption. The granting of an exemption on request of an applicant is subject to litigation in the same manner as other issues in the operating license or combined license hearing.
89
Restrictions on Changes to a DC or COL Referencing a DC for Reasons of Standardization (contd)
§ 52.93 Exemptions and variances.
(cd) An applicant for a combined license who has filed an application referencing a nuclear power reactor manufactured under a manufacturing license issued under subpart F of this part may include in the application a request for a departure from one or more design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure.
§ 52.171
- manufacturing Finality of
- licenses; information requests.
(b) * * *
(2) An applicant or licensee who references or uses a nuclear power reactor manufactured under a manufacturing license under this subpart may request a departure from the design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. The granting of a departure on request of an applicant is subject to litigation in the same manner as other issues in the construction permit or combined license hearing.
90
Design Certification Rule Section IX Appendix D to Part 52Design Certification Rule for the AP1000 Design IX. [Reserved]Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
§ 52.99 Inspection during construction.
(a) Licensee schedule for completing inspections, tests, or analyses. The licensee shall submit to the NRC, no later than 1 year after issuance of the combined license or at the start of construction as defined at 10 CFR 50.10(a), whichever is later, its schedule for completing the inspections, tests, or analyses in the ITAAC. The licensee shall submit updates to the ITAAC schedules every 6 months thereafter and, within 1 year6 months of its scheduled date for initial loading of fuel, the licensee shall submit updates to the ITAAC schedule every 3060 days until the final notification is provided to the NRC under paragraph (c)(1) of this section.
91
References to Standard Design Approvals Subpart C Combined Licenses
§ 52.73 Relationship to other subparts.
(a) An application for a combined license under this subpart may, but need not, reference a standard design certification, standard design approval, or more than one standard design approval, provided each referenced standard design approval is for different portions of the same reactor design, or manufacturing license issued under subparts B, E, or F of this part, respectively, or an early site permit issued under subpart A of this part. * *
- 92
References to Standard Design Approvals (contd)
§ 52.79 Contents of applications; technical information in final safety analysis report.
(c) If the combined license application references a one or more standard design approvals, then the following requirements apply:
(1) The final safety analysis report need not contain information or analyses submitted to the Commission in connection with theeach design approval referenced, provided, however, that the final safety analysis report must either include or incorporate by reference theeach standard design approval final safety analysis report and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the characteristics of the site fall within the site parameters specified in theeach design approval. In addition, the plant-specific PRA information must use the PRA information for theeach design approval and must be updated to account for site-specific design information and any design changes or departures.
(2) The final safety analysis report must demonstrate that all terms and conditions that have been included in the each design approval will be satisfied by the date of issuance of the combined license.
(3) The final safety analysis report must demonstrate that the interfaces between a referenced standard design approval and the balance of the nuclear power plant not described in a referenced standard design approval conform to the descriptions, analyses, and evaluations stated in the referenced standard final safety analysis report.
(4) If the combined license application references more than one standard design approval, the final safety analysis report must demonstrate that the interfaces between each of the referenced standard design approvals conform to the descriptions, analyses, and evaluations stated in each referenced standard final safety analysis report.
93
References to Standard Design Approvals (contd)
Subpart E Standard Design Approvals
§ 52.133 Relationship to other subparts.
(a) This subpart applies to a person that requests a standard design approval from the NRC staff separately from an application for a construction permit filed under 10 CFR part 50 or a combined license filed under subpart C of this part. An applicant for a construction permit, operating license, or combined license, or manufacturing license may reference a standard design approval, or more than one standard design approval, provided each referenced standard design approval is for different portions of the same reactor design Subpart F Manufacturing Licenses
§ 52.153 Relationship to other subparts.
(b) Subpart B of this part governs the certification by rulemaking of the design of standard nuclear power facilities. Subpart E of this part governs the NRC staff review and approval of standard designs for a nuclear power facility. A manufacturing license applicant may reference a standard design certification or a standard design approval, or more than one standard design approval, provided each referenced standard design approval is for different portions of the same reactor design. These subparts may also be used independently of the provisions in this subpart 94
Modify Requirements to Evaluate Conformance with the Standard Review Plan Part 50 - Domestic Licensing of Production and Utilization Facilities
§ 50.34 Contents of applications; technical information.
(h) [RESERVED]
95
Modify Requirements to Evaluate Conformance with the Standard Review Plan (contd)
Subpart A Early Site Permits
§ 52.17 Contents of applications; technical information.
(a) * * *
(1) * * *
(xii) [Reserved]
Subpart B Standard Design Certifications
§ 52.47 Contents of applications; technical information.
(a) * * *
(9) [RESERVED]
§ 52.79 Contents of applications; technical information in final safety analysis report.
(a) * * *
(41) [Reserved]
96
Modify Requirements to Evaluate Conformance with the Standard Review Plan (contd)
Subpart E Standard Design Approvals
§ 52.137 Contents of applications; technical information.
(a) * * *
(9) [RESERVED]
Subpart F Manufacturing Licenses
§ 52.157 Contents of applications; technical information in final safety analysis report.
(f) * * * *
(30) [RESERVED]
97
Align Requirements for Timely Completion of Construction
§ 50.100 Revocation, suspension, modification of licenses, permits, and approvals for cause.
A license, permit, or standard design approval under parts 50 or 52 of this chapter may be revoked, suspended, or modified, in whole or in part, for any material false statement in the application or in the supplemental or other statement of fact required of the applicant; or because of conditions revealed by the application or statement of fact of any report, record, inspection, or other means which would warrant the Commission to refuse to grant a license, permit, or approval on an original application (other than those relating to §§50.51, 50.42(a),
and 50.43(b)); or for failure to manufacture a reactor, or construct or operate a facility in accordance with the terms of the permit or license, provided, however, that failure to make timely completion of the proposed construction or alteration of a facility under a construction permit under part 50 of this chapter or a combined license under part 52 of this chapter shall be governed by the provisions of §50.55(b); or for violation of, or failure to observe, any of the terms and provisions of the act, regulations, license, permit, approval, or order of the Commission.
98
Clarify Applicable Regulatory Parts for Certified Designs Appendix D to Part 52Design Certification Rule for the AP1000 Design V. Applicable Regulations A.1. Except as indicated in paragraph B of this section, the regulations that apply to the AP1000 design are in 10 CFR parts 20, 50, 52, 73, and 100, codified as of January 23, 2006, that are applicable and technically relevant, as described in the FSER (NUREG-1793) and Supplement No. 1 Appendix E to Part 52Design Certification Rule for the ESBWR Design V. Applicable Regulations A. Except as indicated in paragraph B of this section, the regulations that apply to the ESBWR design are in 10 CFR parts 20, 50, 52, 73, and 100, codified as of October 6, 2014, that are applicable and technically relevant, as described in the FSER (NUREG-1966) and Supplement No. 1.
99
Clarify the Requirements for Environmental Qualification Program for Manufacturing Licenses
§ 52.157 Contents of applications; technical information in final safety analysis report.
(f) * * * *
(6) A description of the program, and its implementation, required by § 50.49(a) of this chapter for the environmental qualification of electric equipment important to safety and t The list of electric equipment important to safety that is required by 10 CFR 50.49(d);
100
Severe Accident Treatment Requirements Part 50 - Domestic Licensing of Production and Utilization Facilities
§ 50.34 Contents of applications; technical information.
(a) * * *
(15) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a permit to construct a light-water power reactor under this part shall submit a description and analysis of design features for the prevention and mitigation of severe accidents.
(b) * * *
(13) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a license to operate a light-water power reactor under this part shall submit a description and analysis of design features for the prevention and mitigation of severe accidents.
101
Severe Accident Treatment Requirements (contd)
§ 50.59 Changes, tests, and experiments.
(c) * * *
(2) * * *
(vii) Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or (viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.;
(ix) For a power reactor licensed after [EFFECTIVE DATE OF FINAL RULE], result in a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously evaluated and determined to be not credible could become credible; or (x) For a power reactor licensed after [EFFECTIVE DATE OF FINAL RULE], result in a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously evaluated.
102
Probabilistic Risk Assessment Requirements
§ 50.34 Contents of applications; technical information.
(a) * * *
(3) * * *
(iii) Information relative to materials of construction, general arrangement, and approximate dimensions, sufficient to provide reasonable assurance that the final design will conform to the design bases with adequate margin for safety,.
(iv) A description and analysis of the fire protection design features for the plant necessary to comply with General Design Criterion 3 of appendix A to this part.
(4) * * *Analysis and evaluation of ECCS cooling performance and the need for high point vents following postulated loss-of-coolant accidents must be performed, as applicable, in accordance with the requirements of § 50.46 and § 50.46a of this part.
for facilities for which construction permits may be issued after December 28, 1974.
(14) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a permit to construct a power reactor under this part shall submit a description of the plant-specific probabilistic risk assessment (PRA) and its results.
(15) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a permit to construct a light-water power reactor under this part shall submit a description and analysis of design features for the prevention and mitigation of severe accidents.
(16)On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a permit to construct a light-water power reactor under this part shall submit a discussion on proposed technical resolutions of all generic issues identified since July 21, 1999, unresolved safety issues, and medium- and high-priority generic safety issues identified before July 21, 1999, that are relevant to the design. These issues are based upon the applicant's review of publicly available information published up to 6 months before the docket date of the application (for example, the issues listed in NRCs NUREG-0933, "Resolution of Generic Safety Issues,").
103
Probabilistic Risk Assessment Requirements (contd)
(b) * * *
(4) * * *Analysis and evaluation of ECCS cooling performance following postulated loss-of-coolant accidents shall be performed, as applicable, in accordance with the requirements of § 50.46 for facilities for which a license to operate may be issued after December 28, 1974.
(6) * * *
(viii) A description and analysis of the fire protection design features for the plant necessary to comply with § 50.48 and a description of the fire protection program required by § 50.48 of this chapter and its implementation.
(9) Provided the terms of § 50.61(b)(1) apply, a description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in § 50.61 (b)(1) and (b)(2).
(13) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a license to operate a light-water power reactor under this part shall submit a description and analysis of design features for the prevention and mitigation of severe accidents.
(14) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a license to operate a power reactor under this part shall submit a description of the plant-specific probabilistic risk assessment and its results..
(15) On or after [EFFECTIVE DATE OF FINAL RULE], applicants for a license to operate a light-water power reactor under this part shall submit a discussion on proposed technical resolutions of all generic issues identified since July 21, 1999, unresolved safety issues, and medium- and high-priority generic safety issues identified before July 21, 1999, that are relevant to the design. These issues are based upon the applicant's review of publicly available information published up to 6 months before the docket date of the application (for example, the issues listed in NRCs NUREG-0933, "Resolution of Generic Safety Issues.").
104
Probabilistic Risk Assessment Requirements (contd)
§ 50.69 Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors (b) * * *
(1) A holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder of a renewed LWR license under part 54 of this chapter; an applicant for a construction permit or operating license under this part; or an applicant for a design approval, a combined license, or manufacturing license under part 52 of this chapter; may voluntarily comply with the requirements in this section as an alternative to compliance with the following requirements for RISC-3 and RISC-4 SSCs::
(1) This section describes alternative requirements for the SSCs of a light-water reactor plant. Holders of an operating license or construction permit under this part, a combined license or manufacturing license under part 52 of this chapter, or a renewed license under part 54 of this chapter may voluntarily comply with the requirements in this section. Compliance with the requirements in this section may be proposed when applying for a standard design certification or approval, manufacturing license, or combined license under part 52 of this chapter and when applying for a construction permit or operating license under this part. For RISC-3 and RISC-4 SSCs, the requirements in this section are an alternative to compliance with the following:
105
Probabilistic Risk Assessment Requirements (contd)
§ 50.71 Maintenance of records, making of reports (h)
(1) No later than the scheduled date for initial loading of fuel, each holder of an operating license for a power reactor under this part or a combined license under subpart C of 10 CFR part 52 of this chapter shall develop a level 1 and a level 2 probabilistic risk assessment (PRA).
(2) Each holder of a combined license licensee required to develop a PRA shall maintain and upgrade the PRA required by paragraph (h)(1) of this sectionto reflect the as-built, as-operated facility. In addition, the licensee mustThe upgraded the PRA mustto cover initiating events and modes of operation contained in NRC-endorsed consensus standards on PRA that are endorsed by the NRC. The upgrade must be completed within five years of NRC endorsement of the standardin effect one year prior to each required upgrade. The PRA must be maintained and upgraded every four years until the permanent cessation of operations under § 50.82(a)(1) or § 52.110(a) of this chapter.
(3) Each holder of a combined license licensee required to develop a PRA shall, no later than the date on which the licensee submits an application for a renewed license, upgrade the PRA required by paragraph (h)(1) of this section to cover all modes and all initiating events.
106
Three Mile Island Requirements (f) Additional TMI-related requirements. In addition to the requirements of paragraph (a) of this section, each applicant for a light-water-reactor construction permit or manufacturing license whose application was pending as of February 16, 1982, shall meet the requirements in paragraphs (f)(1) through (3) of this section. This regulation applies to the pending applications by Duke Power Company (Perkins Nuclear Station, Units 1, 2, and 3), Houston Lighting & Power Company (Allens Creek Nuclear Generating Station, Unit 1), Portland General Electric Company (Pebble Springs Nuclear Plant, Units 1 and 2), Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2), Puget Sound Power & Light Company (Skagit/Hanford Nuclear Power Project, Units 1 and 2), and Offshore Power Systems (License to Manufacture Floating Nuclear Plants). The number of units that will be specified in the manufacturing license above, if issued, will be that number whose start of manufacture, as defined in the license application, can practically begin within a 10-year period commencing on the date of issuance of the manufacturing license, but in no event will that number be in excess of ten. The manufacturing license will require the plant design to be updated no later than 5 years after its approval. Paragraphs (f)(1)(xii), (2)(ix), and (3)(v) of this section, pertaining to hydrogen control measures, must be met by all applicants covered by this regulation. However, the Commission may decide to impose additional requirements and the issue of whether compliance with these provisions, together with 10 CFR 50.44 and criterion 50 of appendix A to 10 CFR part 50, is sufficient for issuance of that manufacturing license which may be considered in the manufacturing license proceeding. In addition, eEach applicant for a design certification, design approval, combined license, or manufacturing license under part 52 and each applicant for a power reactor construction permit or power reactor operating license under part 50 of this chapter shallmust demonstrate compliance with the technically relevant portions of the requirements in paragraphs (f)(21) andthrough (f)(3) of this section, except for paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v).
107
Three Mile Island Requirements (contd)
(1) [Reserved]
(2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. The information must be of the type customarily required to satisfy §10 CFR 50.35(a)(2) or to address unresolved generic safety issues.
(i) [Reserved] Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's. (Applicable to construction permit applicants only) (I.A.4.2.)
(ii) Establish a program, to begin during construction and follow into operation, for integrating and expanding current efforts to improvedeveloping and maintaining plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, and operator training. , and coordination with INPO and other industry efforts. (Applicable to construction permit applicants only) (I.C.9)
(iii) Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts. (I.D.1)
(iv) Provide a plant safety parameter display console system that will display for displaying to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded. (I.D.2) 108
Three Mile Island Requirements (contd)
(v) Provide for automatic indication of the bypassed and operable status of safety systems. (I.D.3)
(vi) [Reserved]Provide the capability of high point venting of noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1)
(vii) Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain accident source term11 radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment. (II.B.2)
(viii) Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain accident source term11 radioactive materials without radiation exposures to any individual exceeding 5 rems to the whole body or 50 rems to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, radioiodines and cesiums, and nonvolatile isotopes),
hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations. (II.B.3)
(ix) [Reserved]
109
Three Mile Island Requirements (contd) auxiliary feedwater system flow indication in the control room. (Applicable to PWR's only)
(II.E.1.2)
(xiii) Provide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish and maintain natural circulation in hot standby conditions with only onsite power available. (Applicable to PWR's only) (II.E.3.1)
(xiv) Provide containment isolation systems that: (II.E.4.2)
(A) Ensure all non-essential systems are isolated automatically by the containment isolation system, (B) For each non-essential penetration (except instrument lines) have two isolation barriers in series, (C) Do not result in reopening of the containment isolation valves on resetting of the isolation signal, (D) Utilize a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, (E) Include automatic closing on a high radiation signal for all systems that provide a path to the environs.
110
Three Mile Island Requirements (contd)
(xv) Provide a capability for containment purging/venting designed to minimize the purging time consistent with ALARA principles for occupational exposure. Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions. (II.E.4.4)
(xvi) [Reserved]Establish a design criterion for the allowable number of actuation cycles of the emergency core cooling system and reactor protection system consistent with the expected occurrence rates of severe overcooling events (considering both anticipated transients and accidents). (Applicable to B&W designs only). (II.E.5.1)
(xvii) Provide instrumentation to measure, record and readout in the control room:
(A) containment pressure, (B) containment water level, (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents at all potential, accident release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples. (II.F.1)
(xviii) Provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's. (II.F.2) 111
Three Mile Island Requirements (contd)
(xix) Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage. (II.F.3)
(xx) Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A) Level indicators are powered from vital buses; (B) motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety and (C) electric power is provided from emergency power sources. (Applicable to PWR's only). (II.G.1)
(xxi) Design auxiliary heat removal systems such that necessary automatic and manual actions can be taken to ensure proper functioning when the main feedwater system is not operable. (Applicable to BWR's only). (II.K.1.22)
(xxii) [Reserved]
(xxiii) [Reserved]
(xxiv) [Reserved]
(xxv) [Reserved]
(xxvi) Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) accident source term11 radioactive materials following an accident. Applicants shall submit a leakage control program, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency. (III.D.1.1) 112
Three Mile Island Requirements (contd)
(xxvii) Provide for monitoring of in plant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions. (III.D.3.3)
(xxviii) Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in an accident source term11 release, and make necessary design provisions to preclude such problems.
(III.D.3.4)
(3) * * *
(h) [RESERVED] Conformance with the Standard Review Plan (SRP).
113
Emergency Planning Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities I. Introduction * * *
- IV. Content of Emergency Plans F. Training
- 2. * * *
(a) * * *
(i) For an operating license issued under this part, this exercise must be conducted within 2 years before the issuance of the first operating license for full power (one authorizing operation above 5 percent of rated thermal power) of the first reactor and shall include participation by each State and local government within the plume exposure pathway EPZ and each state within the ingestion exposure pathway EPZ. If the full participation exercise is conducted more than 1 year prior to issuance of an operating licensee for full power, an exercise which tests the licensee's onsite emergency plans must be conducted within 1one year before issuance of an operating license for full power. This exercise need not have State or local government participation.
(ii) For a combined license issued under part 52 of this chapter, this exercise must be conducted within two 2 years beforeof the scheduled date for initial loading of fuel. If the first full participation exercise is conducted more than 1one year before the scheduled date for initial loading of fuel, an exercise which tests the licensee's onsite emergency plans must be conducted within one1 year before the scheduled date for initial loading of fuel. This exercise need not have State or local government participation. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of the first full participation exercise, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply. 114
Emergency Planning (contd)
(iii) For a combined license issued under part 52 of this chapter, if the applicantlicensee currently has an operating reactor at the site, an exercise, either full or partial participation, shall be conducted for each subsequent reactor constructed on the site.
The exercise for each subsequent reactor is not required if, in its application for a combined license, the licensee includes an analysis that shows that an exercise for the new reactor would not demonstrate any new features (e.g., emergency response organization, facilities, procedures) or capabilities beyond those in the emergency plan for the existing reactor(s). This exercise may be incorporated in the exercise requirements of Sections IV.F.2.b. and c. in this appendix. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of this exercise for the new reactor, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply.
115
Emergency Planning (contd)
(j) * * *
(iii) For each site, an 8-calendar-year exercise cycle shall begin in the calendar year of the completion of the first subsequent exercise conducted to meet the requirements in paragraphs 2.b and c of this section. In each 8-calendar-year exercise cycle, nuclear power reactor licensees shall vary the content of scenarios during exercises conducted under paragraph 2 of this section to provide the opportunity for the ERO to demonstrate proficiency in the key skills necessary to respond to the following scenario elements:
(1) Hostile action directed at the plant site; (2) No radiological release or an unplanned minimal radiological release that does not require public protective actions; (3) An initial classification of, or rapid escalation to, a Site Area Emergency or General Emergency; (4) Implementation of strategies, procedures, and guidance under § 50.155(b)(2);
and (5) Integration of offsite resources with onsite response. 116
Emergency Planning (contd)
(iv) The licensee shall maintain a record of exercises conducted during each 8-year exercise cycle that documents the content of scenarios used to comply with the requirements of section IV.F.2.j of this appendix.
(v) [Reserved]Each licensee shall conduct a hostile action exercise for each of its sites no later than December 31, 2015.
(vi) [Reserved]The first 8-year exercise cycle for a site will begin in the calendar year in which the first hostile action exercise is conducted. For a site licensed under 10 CFR part 52, the first 8-year exercise cycle begins in the calendar year of the initial exercise required by section IV.F.2.a of this appendix 117
Emergency Planning (contd)
The exercises conducted under paragraph 2 of this section by nuclear power reactor licensees must provide the opportunity for the ERO to demonstrate proficiency in the key skills necessary to implement the principal functional areas of emergency response identified in paragraph 2.b of this section. Each exercise must provide the opportunity for the ERO to demonstrate key skills specific to emergency response duties in the control room, TSC, OSC, EOF, and joint information center. For each site, an 8-calendar-year exercise cycle shall begin in the calendar year of the completion of the first subsequent exercise conducted to meet the requirements in paragraph 2.b and c of this section. Additionally, iIn each 8-calendar-year exercise cycle, nuclear power reactor licensees shall vary the content of scenarios during exercises conducted under paragraph 2 of this section to provide the opportunity for the ERO to demonstrate proficiency in the key skills necessary to respond to the following scenario elements: hostile action directed at the plant site, no radiological release or an unplanned minimal radiological release that does not require public protective actions, an initial classification of or rapid escalation to a Site Area Emergency or General Emergency, implementation of strategies, procedures, and guidance under § 50.155(b)(2), and integration of offsite resources with onsite response. The licensee shall maintain a record of exercises conducted during each 8-year exercise cycle that documents the content of scenarios used to comply with the requirements of this paragraph. Each licensee shall conduct a hostile action exercise for each of its sites no later than December 31, 2015. The first 8-year exercise cycle for a site will begin in the calendar year in which the first hostile action exercise is conducted.
For a site licensed under 10 CFR part 52, the first 8-year exercise cycle begins in the calendar year of the initial exercise required by section IV.F.2.a of this appendix.
118
Operators Licenses Part 55Operators Licenses
§ 55.4 Definitions.
Plant-referenced simulator means a simulator modeling the systems of the reference plant with which the operator interfaces or, for a plant that is being constructed, will interface, in the control room, including operating consoles, and which permits use of the reference plant's procedures.
Reference plant means the specific nuclear power plant from which a simulation facility's control room configuration, system control arrangement, and design data are derived. The reference plant may or may not be actually constructed.
119
Operators Licenses (contd)
§ 55.45 Operating tests.
(a) * * *
(b) Implementation - Administration. (1) The operating test will be administered in a plant walkthrough and in either - A a simulation facility that the Commission has approved for use after application has been made by the facility licensee under § 55.46(b);, (2) A a plant-referenced simulator (§ 55.46(c)), (3) T or the plant, if approved for use in the administration of the operating test by the Commission under § 55.46(b).
(2) If a facility is under construction, suitable alternatives may be used in lieu of the plant walkthrough portion of the operating test.
[52 FR 9460, Mar. 25, 1987, as amended at 53 FR 43421, Oct. 27, 1988; 62 FR 59276, Nov. 3, 1997; 66 FR 52667, Oct. 17, 2001]
§ 55.46 Simulation facilities.
(a) * * *
(c) Plant-referenced simulators.
(1) * * *
(2) * * *:
(i) The plant-referenced simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that either replicate the most recent core load in the nuclear power reference plant for which a license is being sought; and, or, prior to initial fuel load, replicate the intended initial core load for the nuclear power reference plant for which a license is being sought; and, 120
Operators Licenses (contd)
§ 55.47 Waiver of examination and test requirements.
(a) * * *
(1) Has had extensive actual operating experience at a comparable facility, as determined by the Commission, within two years before the date of application; (2) Hhas discharged his or her responsibilities competently and safely and is capable of continuing to do so; and (3) Hhas learned the operating procedures for and is qualified to operate competently and safely the facility designated in the application.
(2) The Commission may accept as proof of the applicant's past performance a certification of an authorized representative of the facility licensee or of a holder of an authorization by which the applicant was previously employed. The certification must contain a description of the applicant's operating experience, including an approximate number of hours the applicant operated the controls of the facility, the duties performed, and the extent of the applicant's responsibility.
(3) The Commission may accept as proof of the applicant's current qualifications a certification of an authorized representative of the facility licensee or of a holder of an authorization where the applicant's services will be utilized.
(b) On application, the Commission may waive any or all of the requirements for a written examination and operating test for a licensee who applies for a license to operate one or more subsequent units at a multiunit site, licensed collectively or individually, if it finds that the -
(1) Subsequent unit(s) is/are approved to be, or was/were constructed to, the same standard design or modular design, as defined in § 52.1 of this chapter, as the unit(s) on which the applicant is already licensed, or the subsequent unit(s) is/are otherwise essentially identical to the unit(s) on which the applicant is already licensed, and (2) The applicant has been sufficiently trained on the differences between the units.
121
Operators Licenses (contd)
§ 55.31 How to apply.
(a) * * *
(1) * * *
(4) (i) Provide evidence that the applicant has successfully completed the facility licensee's requirements to be licensed as an operator or senior operator and of the facility licensee's need for an operator or a senior operator to perform assigned duties. An authorized representative of the facility licensee shall certify this evidence on Form NRC-398. This certification must include details of the applicant's qualifications, and details on courses of instruction administered by the facility licensee, and describe the nature of the training received at the facility, and the startup and shutdown experience received. In lieu of these details, the Commission may accept certification that the applicant has successfully completed a Commission-approved training program that is based on a systems approach to training and that uses a simulation facility acceptable to the Commission under § 55.45(b) of this part; (ii) If the NRC Form 398 is submitted before the facility licensee is required to have the requalification program described in § 55.59 in effect as described under § 50.54(i-1) of this chapter, describe how the applicants knowledge, skills, and abilities will be maintained sufficient to safely perform the functions of an operator or senior operator after the applicant passes the written examination described in §§ 55.41 and 55.43 and the operating test described in §55.45, and prior to participation as a licensed operator in the requalification program. In lieu of this description, the Commission will accept a statement that the applicant will participate in a Commission-approved continuing training program developed by using a systems approach to training within 3 months of receiving notice that the applicant has passed the written examination and operating test;
- * * *
- 122
Status of ITAAC Completion
§ 52.97 Issuance of combined licenses.
(a)
(1) * * *
(2) The Commission may also find, at the time it issues the combined license, that certain acceptance criteria in one or more of the inspections, tests, analyses, and acceptance criteria (ITAAC) in a referenced early site permit or standard design certification have been are met. This finding will finally resolve that those acceptance criteria have been met, those acceptance criteria will be are deemed to be excluded from the combined license, and findings under §52.103(g) with respect to those acceptance criteria are unnecessary.
123
Reporting Requirements
§ 50.71 Maintenance of records, making of reports (i) Each licensee shall notify the Commission as specified in § 50.4 or § 52.3 of this chapter, of successfully completing power ascension testing or startup testing, as applicable, within 30 calendar days of completing the testing.
124
Conditions of License
§ 50.54 Conditions of licenses.
The following paragraphs of this section, with the exception of paragraphs (r) and (gg), and the applicable requirements of 10 CFR 50.55a, are conditions in every nuclear power reactor operating license issued under this part. The following paragraphs with the exception of paragraph (r), (s), and (u) of this section are conditions in every combined license issued under part 52 of this chapter, provided, however, that paragraphs (i) introductory text, (i)(1), (j), (k), (l),
(m), (n), (q)(2), (w), (x), (y), (z), and (hh) of this section are only applicable after the Commission makes the finding under § 52.103(g) of this chapter. The following paragraphs of this section, with the exception of paragraphs (a)(1) through (4), (m)(2) and (3), (o), (q)(6), (r), (s)(1), (u), (w)(1) through (4), (z), (bb), (ff), (gg)(1) and (2), and (jj) are conditions in every non-power production or utilization facility operating license issued under this part.
(j) Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor utilization facility shall be manipulated only with the knowledge and consent of an operator or senior operator licensed pursuant to part 55 of this chapter present at the controls.
(k) An operator or senior operator licensed pursuant to part 55 of this chapter shall be present at the controls at all times during the operation of the utilization facility.
125
Conditions of License (contd)
(m)
(1) A senior operator licensed pursuant to part 55 of this chapter shall be present at the utilization facility or readily available on call at all times during its operation, and shall be present at the utilization facility during initial start-up and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling, or as otherwise prescribed in the utilization facility license.
(n) The licensee shall not, except as authorized pursuant to a construction permit, make any alteration in the production or utilization facility constituting a change from the technical specifications previously incorporated in a license or construction permit pursuant to § 50.36 of this part.
126
Conditions of License (contd)
(q) * * *
(6) * * *
(s)* * *
(2)* * *
(ii) If after April 1, 1981, the NRC finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency (including findings based on requirements of appendix E, section IV.D.3) and if the deficiencies (including deficiencies based on requirements of appendix E, section IV.D.3) are not corrected within four months of that finding, the Commission will determine whether the reactor production or utilization facility shall be shut down until such deficiencies are remedied or whether other enforcement action is appropriate. In determining whether a shutdown or other enforcement action is appropriate, the Commission shall take into account, among other factors, whether the licensee can demonstrate to the Commission's satisfaction that the deficiencies in the plan are not significant for the plant in question, or that adequate interim compensating actions have been or will be taken promptly, or that there are other compelling reasons for continued operation.
(z) Each licensee with a utilization facility nuclear power reactor licensed pursuant to sections 103 or 104b. of the Act shall immediately notify the NRC Operations Center of the occurrence of any event specified in § 50.72 of this part.
127
Conditions of License (contd)
(ee)
(1) Each license issued under this part authorizing the possession of byproduct and special nuclear material produced in the operation of the licensed reactor production or utilization facility includes, whether stated in the license or not, the authorization to receive back that same material, in the same or altered form or combined with byproduct or special nuclear material produced in the operation of another reactor of the same licensee located at that site, from a licensee of the Commission or an Agreement State, or from a non-licensed entity authorized to possess the material.
128
Discontinuing Priority Ranking Model for GIs 129
Discontinuing Priority Ranking Model for GIs (contd) 130
References ADAMS Accession Document Title Number/FR Citation Transcript of the Advisory Committee on Reactor Safeguard 683rd Full Committee Meeting - March ML21075A211 4, 2021, Pages 1-288 (Open)
Summary of March 2, 2021 Category 3 Public Meeting to Discuss the Alignment of Licensing ML21076A098 Processes and Lessons Learned from New Reactor Licensing Rulemaking, dated March 19, 2021 86 FR 7513 - Regulatory Basis-Alignment of Licensing Processes and Lessons Learned From New 86 FR 7513 Reactor Licensing 04/29/2020 - Public Meeting to Discuss the Status of Rulemaking to Align Licensing Processes and ML20141L609 Apply Lessons Learned from New Reactor Licensing [NRC-2009-0196; RIN 3150-AI66]
85 FR 9328 - Revision of Fee Schedules; Fee Recovery for Fiscal Year 2020 85 FR 9328 2/14/20 - Letter to Petitioner M. Lorton on Behalf of Algignis, Inc.; Results of PRM Sufficiency ML20008D640 Review; Petition for Rulemaking for 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Pkg) 11/18/2019 - 84 FR 63565 - Miscellaneous Corrections 84 FR 63565 11/21/2019 - Category 3 Public Meeting Summary RE: Regulatory Basis: Rulemaking to Align ML19344C768 Licensing Processes and Apply Lessons Learned from New Reactor Licensing (NRC-2009-0196)
Transcript of the Advisory Committee on Reactor Safeguards Regulatory Policies & Practices-Part ML19294A009 50 52 Meeting - September 20, 2019 131
References (contd)
ADAMS Accession Document Title Number/FR Citation SECY-19-0084, Status of Rulemaking to Align Licensing Processes and Lessons Learned from New ML19161A169 Reactor Licensing (RIN 3150-AI66)
SECY-19-0034, Improving Design Certification Content ML19080A034 Summary of January 15, 2019 Public Meeting to Discuss the Proposed Rulemaking to Align the ML19023A046 Regulations in Parts 50 and 52 to Address Updates to the Licensing Processes and Lessons Learned for Future New Reactor Applications SECY-15-0002, Proposed Updates of Licensing Policies, Rules and Guidance for Future New ML13277A420 Reactor Applications SRM-SECY-15-002, Staff Requirements-SECY-15-002-Proposed Updates of Licensing Policies, ML15266A023 Rules and Guidance for Future New Reactor Applications Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants 60 FR 32138 SECY-89-013, Design Requirements Related to the Evolutionary Advanced Light Water Reactors, ML003707947 dated January 19, 1989 SECY-90-016, Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to ML003707849 Current Regulatory Requirements, dated January 12, 1990 SECY-93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light- ML003708021 Water Reactor (ALWR) Designs, dated April 2, 1993 Bipartisan Policy Center Report Recommendations on the New Reactor Licensing Process ML13059A240 132
References (contd)
ADAMS Accession Document Title Number/FR Citation NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for https://www.nrc.gov/rea Nuclear Power Plants: LWR Edition, with updates through 2007 ding-rm/doccollections/
nuregs/staff/sr0800/
Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in ML17317A256 Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated 2018 Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy ML090410014 of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated 2009 Regulatory Guide 1.201, Revision 0, Guidelines for Categorizing Structures, Systems, and ML061090627 Components in Nuclear Power Plants According to Their Safety Significance, dated 2006 Regulatory Guide 1.189 Revision 3, Fire Protection for Nuclear Power Plants. dated 2018 ML17340A875 Regulatory Guide 1.206, Revision 1, Applications for Nuclear Power Plants. dated 2018 ML18131A181 Regulatory Guide 5.84, Revision 0, Fitness-for-Duty for New Nuclear Power Plant ML15083A412 Construction Sites, dated July 2015 Draft Regulatory Guide 5040, Urine Specimen Collection and Test Result Review Under 10 84 FR 48750 CFR Part 26, Fitness-for-Duty Programs, dated September 16, 2019 NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, ML18093B087 and Information Requests. DT-19-15, dated 2019.
NRC NUREG-1409, Revision 1, Backfitting Guidelines., Draft Report for Comment, dated ML18109A498 2020 133