ML21049A277

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NRR E-mail Capture - Transmittal of Information Related to Technology Inclusive Content of Application Project (Ticap) and Advanced Reactor Content of Application Project (Arcap) to Support Upcoming Public Meeting Scheduled for February 25,
ML21049A277
Person / Time
Issue date: 02/18/2021
From: Joseph Sebrosky
NRC/NRR/DANU
To: Afzali A, Austgen K, Draffin C
Nuclear Energy Institute, Southern Company Services, US Dept of Energy, Office of Nuclear Energy
References
Download: ML21049A277 (34)
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Text

From: Sebrosky, Joseph Sent: Thursday, February 18, 2021 9:59 AM To: Afzali, Amir; AUSTGEN, Kati; Cyril Draffin Cc: NICHOL, Marcus; TSCHILTZ, Michael; Shams, Mohamed; Smith - NRR, Brian; Sanfilippo, Nathan; Segala, John; Lauron, Carolyn

Subject:

Transmittal of Information Related to Technology Inclusive Content of Application Project (TICAP) and Advanced Reactor Content of Application Project (ARCAP) to Support Upcoming Public Meeting Scheduled for February 25, 2021 Attachments: ARCAP ISG - Initial Startup 18-21 version.pdf; ARCAP ISG - Organization 18-21 version.pdf; ARCAP Chapter Crosswalk-2-17 version.pdf Amir Afzali Southern Company Services Licensing and Policy Director - Next Generation Reactors Kati Austgen Sr. Project Manager, New Reactors Nuclear Energy Institute Cyril Draffin Senior Fellow, Advanced Nuclear United States Nuclear Industry Council Mr. Afzali, Ms. Austgen, and Mr. Draffin, The purpose of this email is to provide you with the attached files to support the upcoming February 25, 2021, public meeting on construction permit guidance, and advanced reactor stakeholder topics. These files include:

  • Advanced Reactor Content of Application Project (ARCAP) draft interim staff guidance (ISG) for Chapter 11, Organization
  • ARCAP draft ISG for Chapter 12, Initial Startup Program
  • A file titled, ARCAP Chapter Crosswalk 2-17 version, that provides a table with the staffs consideration for a proposed ARCAP structure. A version of this table was previously provided and discussed during a public meeting on October 22, 2020.

The attached files will be referenced in the NRC staff presentations during the February 25, 2021, public meeting. This email will be captured in ADAMS and the email will be made publicly available so that interested stakeholders will have access to the information prior to the meeting. The public meeting notice will be updated to provide a link to the attached documents once the documents are publicly available in ADAMS.

If you have questions regarding the attached documents please contact me.

Sincerely, Joe Sebrosky

Senior Project Manager Advanced Reactor Policy Branch Office of Nuclear Reactor Regulation 301-415-1132

Hearing Identifier: NRR_DRMA Email Number: 1028 Mail Envelope Properties (MN2PR09MB5385C5D90174627B3B5B2EB6F8859)

Subject:

Transmittal of Information Related to Technology Inclusive Content of Application Project (TICAP) and Advanced Reactor Content of Application Project (ARCAP) to Support Upcoming Public Meeting Scheduled for February 25, 2021 Sent Date: 2/18/2021 9:59:01 AM Received Date: 2/18/2021 9:59:01 AM From: Sebrosky, Joseph Created By: Joseph.Sebrosky@nrc.gov Recipients:

"NICHOL, Marcus" <mrn@nei.org>

Tracking Status: None "TSCHILTZ, Michael" <mdt@nei.org>

Tracking Status: None "Shams, Mohamed" <Mohamed.Shams@nrc.gov>

Tracking Status: None "Smith - NRR, Brian" <Brian.Smith@nrc.gov>

Tracking Status: None "Sanfilippo, Nathan" <Nathan.Sanfilippo@nrc.gov>

Tracking Status: None "Segala, John" <John.Segala@nrc.gov>

Tracking Status: None "Lauron, Carolyn" <Carolyn.Lauron@nrc.gov>

Tracking Status: None "Afzali, Amir" <AAFZALI@southernco.com>

Tracking Status: None "AUSTGEN, Kati" <kra@nei.org>

Tracking Status: None "Cyril Draffin" <cyril.draffin@usnic.org>

Tracking Status: None Post Office: MN2PR09MB5385.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1638 2/18/2021 9:59:01 AM ARCAP ISG - Initial Startup 18-21 version.pdf 362231 ARCAP ISG - Organization- 2-18-21 version.pdf 412356 ARCAP Chapter Crosswalk-2-17 version.pdf 158221 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

This draft staff white paper has been prepared and is being released to support ongoing public discussions. The guidance found in this draft white paper uses an interim staff guidance (ISG) format. The staff is considering using the ISG format in the near future to provide guidance to facilitate the near-term review of advanced reactor applications.

This paper has not been subject to NRC management and legal reviews and approvals, and its contents are subject to change and should not be interpreted as official agency positions.

DANU [XX]-ISG-[YYYY-##]

Advanced Reactor Content of Application Initial Startup Program Interim Staff Guidance February X, 2021 1

MLxxxxxxxxx TAC: xxxxxx OFFICE QTE [PGCB PM] [NRR Technical [NRR Technical Lead Lead/Author] Branch Chief]

NAME DATE OFFICE [Other NRR Division [Other NRC Division [Regional Offices, as OGC Directors, as Directors, as appropriate]

appropriate] appropriate]

NAME DATE OFFICE [PGCB LA] [NRR Technical Lead Division Director]

NAME DATE INTERIM STAFF GUIDANCE ADVANCED REACTOR CONTENT OF APPLICATION INITIAL STARTUP PROGRAM DANU-ISG-YYYY-##

PURPOSE The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff is providing this interim staff guidance (ISG) to facilitate the review of advanced reactor content of application guidance that applies to Title 10 of the Code of Federal Regulations (10 CFR) Part 53, Licensing and Regulation of Advanced Nuclear Reactors. Portions of the ISG can also be used to support non-light water reactors (non-LWRs), stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under 10 CFR Part 50 or for a design certification (DC), combined license (COL), a standard design approval (SDA) or a manufacturing license (ML) under 10 CFR Part 52.

The guidance in this ISG supports the development of the portion of an advanced reactor application associated with the Initial Startup Program.

BACKGROUND The goal of the 10 CFR Part 53 rulemaking effort is to develop the regulatory infrastructure to support the licensing of advanced nuclear reactors. The term advanced nuclear reactor, for purposes of this rulemaking, means a nuclear fission or fusion reactor with significant improvements compared to commercial nuclear reactors under construction as of January 2019.

This rulemaking would revise the NRC's regulations by adding a risk-informed, technology-inclusive regulatory framework for advanced nuclear reactors, in response to a growing interest in possible licensing and deployment of advanced nuclear reactors and the related requirements of the Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439) as amended by the Energy Act of 2020. The rule language for 10 CFR Part 53 is under development and, as such, the guidance found in this document is subject to change based on the outcome of this rulemaking. Key documents related to the Part 53 rulemaking, including preliminary proposed rule language and stakeholder comments, can be found at Regulations.gov under Docket ID NRC-2019-0062.

This edition of the ISG is based on the advanced reactor content of application project (ARCAP) whose purpose is to develop technology-inclusive, risk-informed and performance-based application guidance. The ARCAP is broader and encompasses the industry-led technology-inclusive content of application project (TICAP). The guidance found in this ISG supplements the guidance found in DANU-ISG-YYYY-##, Advanced Reactor Content of Application Guidance, which provides a roadmap for developing all portions of an application. The guidance in this ISG is limited to the portion of an advanced reactor application associated with the Initial Startup Program (ISP) of the reactor plant applicant.

As stated above, the Part 53 regulation is under development. As the 10 CFR Part 53 requirements are finalized, this ISG guidance will be supplemented, as necessary, to provide guidance reflecting any differences in requirements between Parts 50/52 and Part 53.

RATIONALE Note - this section will be updated with additional stakeholder interactions - expected during the monthly ARCAP meetings.

APPLICABILITY This ISG is applicable to non-light water reactors (non-LWRs), stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under 10 CFR Part 50 or for a design certification (DC), combined license (COL), a manufacturing license (ML) or a Standard Design Approval (SDA) under 10 CFR Part 52.

GUIDANCE The ISP consists of preoperational testing (i.e., tests conducted following construction and construction related testing, but prior to initial fuel load) and initial startup testing (i.e., tests conducted during and after initial fuel load, up to and including initial power ascension). The primary objective of the ISP is to demonstrate, to the extent possible, that the safety-related (SR), safety-significant (SS) SSCs operate in accordance with the design and as described in the safety analysis report. Additional objectives of the ISP include:

x Providing assurance that the facility exhibits the performance and associated safety margins that are described in the design.

x Satisfying any license conditions (e.g., ITAAC) associated with the ISP.

x Obtaining data to validate the analytical models.

x Familiarizing the plants operating and technical staff with operation of the facility.

x Verifying the adequacy of the plant operating and emergency procedures The applicants plans for the ISP are required by 10 CFR 30.53(c) for radiation detection and monitoring instruments,10 CFR 50.34(b)(6)(iii) for plants applying for an Operating License (OL) under 10 CFR Part 50 and 10 CFR 52.79(a)(28) for plants applying for a Combined License (COL) under 10 CFR 52.79. For plants applying for a COL via 10 CFR 52.79, but referencing a certified design (DC) under 10 CFR 52.47, a Standard Design Approval (SDA) under 10 CFR 52.137or a design with a Manufacturing License (ML) under 10 CFR 52.157, the ISP includes the inspections, tests, analysis and acceptance criteria (ITAAC) associated with the DC, SDA or ML (see 10 CFR 52.47(b)(1) and 52.158(a), respectively).

The detailed description of the ISP can be included in the FSAR or in a separate document referenced in the FSAR. For a Construction Permit (CP) application, the information described in Phase 1 below should be provided with a commitment to provide the Phase 2 information at the OL stage. For a COL, DC, SDA and ML the information described in Phases 1 and 2 should be provided. The reviewer should review the completeness of the ISP information provided with respect to the license being requested and the guidance provided below.

The ISP is generally divided into two phases - the preoperational phase (prior to initial fuel loading) and initial startup testing (initial fuel loading and initial power ascension). If the application is for a CP, the ISP description can be limited to the Phase 1 testing along with a description of the scope, objectives and programmatic controls associated with the test program. For OL, DC, COL, SDA and ML applications, the application should include a description of the Phase 1 and 2 test programs along with a description of the scope, objectives and programmatic controls associated with the test programs.

Phase 1 - Preoperational Testing x Reactivity Control Functions:

> Reactivity control system performance x Heat Removal Functions:

3/4 Pressure boundary integrity 3/4 Normal heat removal and control system performance 3/4 Residual heat removal system integrity and performance x Containment of Radioactive Material:

3/4 Functional containment performance 3/4 Radiation and criticality monitoring system performance 3/4 Radioactive waste processing, handling and storage system performance x Testing required by consensus design codes and standards applied in the design (e.g.,

ASME, IEEE) for items such as pumps, valves, dynamic restraints, electrical equipment, as applicable.

x Flow induced system vibration and thermal expansion tests.

x Electrical system performance for normal and emergency power.

x Equipment identified as necessary for defense-in-depth.

x Instrumentation and control systems relied upon in the safety analysis to perform SR or SS functions.

x Fuel handling and storage system performance.

x Support system performance for SR and SS equipment (e.g. cooling).

Phase 2 - Initial Startup Testing x Initial fuel loading and reactor physics tests:

3/4 Initial criticality 3/4 Shutdown margin 3/4 Reactivity control system performance 3/4 Shutdown time 3/4 Manual scram function 3/4 Neutron monitoring instrumentation operation and calibration x Low power testing:

3/4 Reactivity control system worth 3/4 Neutron monitoring instrumentation operation and calibration 3/4 Neutron flux distribution 3/4 Neutron and gamma radiation surveys 3/4 Operability of alarms and low power protective features 3/4 Reactivity control system performance 3/4 Shutdown time x Power ascension testing:

3/4 Reactivity coefficients and power to flow characteristics 3/4 Neutron flux and power distribution 3/4 Reactivity control system influence on power distribution and core design limits 3/4 Reactivity control system performance 3/4 Shutdown time 3/4 Reactor coolant system performance 3/4 Flow induced vibration monitoring 3/4 Neutron and gamma radiation surveys 3/4 Neutron monitoring instrumentation and calibration 3/4 Operability of alarms and full power protective features 3/4 Plant response to various AOOs (e.g. turbine trip, loss of normal power) x Performance of residual heat removal system.

x Performance of liquid and gaseous waste systems.

x Performance of first-of-a-kind, inherent or passive safety features.

x Flow induced vibration and thermal expansion within design limits.

The ISP should be planned and conducted in an orderly fashion. Accordingly, the description of the ISP in the application should address the following programmatic items related to the development and conduct of the ISP:

x The ISP objectives, including the objectives of each phase of the program.

x The scope of each phase of the ISP.

x The organization and responsibilities for conduct and control of the testing program.

x A general schedule and sequence for conducting the tests, including established hold points.

x The extent to which the test program will use plant operating, emergency and surveillance procedures and technical specifications x The plan for interfacing with other ongoing activities so as to coordinate and avoid interferences.

x The prerequisites which must be in place prior to conducting each test, including implementation of the technical specifications.

x The information to be measured during each test.

x The acceptance criteria for each test and the conditions which would cause the test to be terminated prematurely.

x The review process and documentation to be applied for each test, including verification that any retesting has been satisfactorily completed.

x The review process and bases for concluding the ISP test results support safe operation of the plant.

In addition, the application should include a general description for each test, or group of similar tests (i.e., test abstract), to be conducted. The focus of the test descriptions should be on providing the bases for the tests and test conditions selected, instrumentation to be used, and a description of how the tests will confirm the performance of the SSCs. Each test description should include the acceptance criteria that define the performance, physical condition or analysis results that must be demonstrated to confirm the design characteristics and features perform consistent with the design. The ISP development should also take into consideration ISP experience at other facilities and include measures to avoid problems they have had.

In general, each test should directly, or indirectly through analysis, confirm SSC performance over the full range of operating conditions (normal operation, AOOs, DBEs, DBAs and BDBE conditions) over the plant lifetime. In addition, the performance of other SSCs containing radioactive material (e. g., spent fuel storage) should be confirmed in the ISP. Risk insights from the plants PRA and safety analysis should be used to identify the specific systems and components, test objectives, test conditions and test parameters selected so as to test the risk significant equipment and conditions. Thus, a graded approach to testing can be applied provided the test program provides reasonable assurance the SR and SS SSCs will perform satisfactorily. In addition, the test program should be sequenced and structured so that plant safety is never entirely dependent upon untested SSCs or temporary plant equipment.

The application should describe the responsibilities and guidelines for conduct of the ISP. In general, the applicant is responsible for all aspects of the ISP, although other parties (e.g.,

vendors) may conduct some of the testing. The applicants responsibilities include:

x Defining the qualifications of the personnel managing, conducting and reviewing the test program results.

x Using contractor or vendor personnel, as appropriate.

x Providing training as necessary to ensure that personnel are ready to perform their functions.

x Developing the testing objectives, schedule, sequence, prerequisites, procedures safety precautions and acceptance criteria.

x Managing, controlling and approving key aspects (e.g., prerequisites, procedures) of the test program.

x Establishing a plant review committee to review, evaluate and disposition the test results.

x Coordination with other elements of the plant organization (e.g. engineering, design, operations), as necessary, in planning, conducting and reviewing test results.

x Preparation, approval and retention of test reports.

The guidelines to be followed in conducting the ISP include:

x The tests should be conducted using detailed procedures approved by managers in the applicants ISP organization.

x The staff conducting the tests (including contractors, vendors or others) should have the appropriate training, experience and education determined necessary by management.

x The tests should not be run until there is verification that all prerequisites for the test have been completed or are in place.

x The test sequence should be established to ensure that testing is completed and operability confirmed on systems and equipment needed to support future testing.

x Where modifications have been made to SSCs, retesting should be conducted.

In reviewing the application, the reviewer needs to have reasonable assurance that the requirements to conduct an ISP, as stated in 10 CFR 30.53(c ), 10 CFR 50.34(b)(6)(iii), and 10 CFR 52.79(a)(28) are met for the design and technology under review. This determination should be based on whether the information provided in the application is sufficient to conclude:

1. The Phase 1 test program includes all SR and SS SSCs that can reasonably be tested at the preoperational stage.
2. The Phase 2 test program includes all SR and SS SSCs that were not tested in Phase 1.
3. The applicants responsibilities are clearly described.
4. The description in the application covers all of the overarching items listed previously for developing the ISP, or deviations are justified.
5. Risk insights have been used to select the most important parameters to be measured.
6. First-of-a-kind SSCs and inherent/passive safety features are identified and included in the test program.
7. For applications that use a COL, DC, SDA or ML, the ITAAC associated with these licenses are included in the ISP.
8. The parameters to be measured in the test program are sufficient to determine, directly or through analysis, the SSC performs as designed.
9. Information sufficient to validate the analytical codes has been collected.
10. The applicants process for reviewing test results and determining the acceptability of the results, or requiring a modification and retest, are clearly described and reasonable.

With positive answers to the above items, it can be concluded that the performance of each SR and SS feature of the design has been demonstrated and sufficient data exists to assess the analytical tools used in the safety analysis. Thus, there is reasonable assurance that the ISP is in compliance with the applicable regulations for a CP, COL, DC, SDA or ML.

IMPLEMENTATION The staff will use the information discussed in this ISG in performing safety evaluations of license applications submitted under 10 CFR 50, 52 or 53.

[Identify how the information will facilitate staff review of license amendments, license renewal applications, etc.]

BACKFITTING AND ISSUE FINALITY DISCUSSION

[OGC provides this discussion, but the staff can propose text for OGC consideration].

Example: The NRC staff issuance of this ISG is not considered backfitting as defined in 10 CFR 50.109(a)(1), nor is it deemed to be in conflict with any of the issue finality provisions in 10 CFR Part 52.

CONGRESSIONAL REVIEW ACT

[OGC provides this discussion to support issuance of the final ISG. However, the staff can propose text for OGC consideration].

Example: This ISG is a rule as defined in the Congressional Review Act (5 U.S.C. §§ 801-808).

However, the Office of Management and Budget has not found it to be a major rule as defined in the Congressional Review Act.

FINAL RESOLUTION By [insert date], this information will be transitioned into [identify the appropriate regulatory process (Standard Review Plan (SRP), Regulatory Guide (RG))]. Following the transition of this guidance to the [SRP, RG], this ISG will be closed. .

APPENDIX A. Resolution of Public Comments APPENDIX A Resolution of Public Comments A notice of opportunity for public comment on this Interim Staff Guidance (ISG) was published in the Federal Register (insert FR Citation #) on [date] for a 30-60 day comment period. [Insert number of commenters] provided comments which were considered before issuance of this ISG in final form.

Comments on this ISG are available electronically at the NRC's electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into ADAMS, which provides text and image files of NRC's public documents. Comments were received from the following individuals or groups:

Letter No. ADAMS No. Commenter Affiliation Commenter Name Abbreviation 1

2 3

4 5

The comments and the staff responses are provided below.

Comment 1: [Each comment summary must clearly identify the entity that submitted the comment and the comment itself].

NRC Response: Comment responses should begin with a direct statement of the NRC staffs position on a comment, e.g., the NRC staff agrees with the comment or the NRC staff disagrees with the comment.

x If the NRC staff agrees, explain why and provide a clear statement as to how the relevant language was revised or supplemented to address the comment. Include the following language at the end of the comment response: The final ISG was changed by <describe the change; if necessary by quoting the newly revised language>.

x If the NRC disagrees with a comment and no change was made to the generic communication, then explain why and provide the following language at the end of the comment response: No change was made to the final ISG as a result of this comment.

APPENDIX B References

This draft staff white paper has been prepared and is being released to support ongoing public discussions. This draft white paper uses an interim staff guidance (ISG) format because the staff is considering using this format to provide staff guidance in the near future to support the review of advanced reactor applications.

This paper has not been subject to NRC management and legal reviews and approvals, and its contents are subject to change and should not be interpreted as official agency positions.

DANU [XX]-ISG-[YYYY-##]

Advanced Reactor Content of Application Organization Interim Staff Guidance February X, 2021 1

MLxxxxxxxxx TAC: xxxxxx OFFICE QTE [PGCB PM] [NRR Technical [NRR Technical Lead Lead/Author] Branch Chief]

NAME DATE OFFICE [Other NRR Division [Other NRC Division [Regional Offices, as OGC Directors, as Directors, as appropriate]

appropriate] appropriate]

NAME DATE OFFICE [PGCB LA] [NRR Technical Lead Division Director]

NAME DATE INTERIM STAFF GUIDANCE ADVANCED REACTOR CONTENT OF APPLICATION ORGANIZATION DANU-ISG-YYYY-##

PURPOSE The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff is providing this interim staff guidance (ISG) to facilitate the review of advanced reactor content of application guidance that applies to Title 10 of the Code of Federal Regulations (10 CFR) Part 53, Licensing and Regulation of Advanced Nuclear Reactors. Portions of the ISG can also be used to support non-light water reactors (non-LWRs), stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under 10 CFR Part 50, for a combined license under 10 CFR Part 52 or for a design certification (DC) under 10 CFR Part 52.

The guidance found in this ISG supports the development of the portion of an advanced reactor application associated with an applicants Organization.

BACKGROUND The goal of the 10 CFR Part 53 rulemaking effort is to develop the regulatory infrastructure to support the licensing of advanced nuclear reactors. The term advanced nuclear reactor, for purposes of this rulemaking, means a nuclear fission or fusion reactor with significant improvements compared to commercial nuclear reactors operating on the date of enactment of the Energy Act of 2020 or under construction as of January 2019. This rulemaking would revise the NRC's regulations by adding a risk-informed, technology-inclusive regulatory framework for advanced nuclear reactors, in response to a growing interest in possible licensing and deployment of advanced nuclear reactors and the related requirements of the Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439) as amended by the Energy Act of 2020. The rule language for 10 CFR Part 53 is under development and as such the guidance found in this document is subject to change based on the outcome of this rulemaking. Key documents related to the Part 53 rulemaking, including preliminary proposed rule language and stakeholder comments, can be found at Regulations.gov under Docket ID NRC-2019-0062.

This edition of the ISG is based on the advanced reactor content of application project (ARCAP) whose purpose is to develop technology-inclusive, risk-informed and performance-based application guidance. The ARCAP is broader and encompasses the industry-led technology-inclusive content of application project (TICAP). The guidance found in this ISG supplements the guidance found in DANU-ISG-YYYY-##, Advanced Reactor Content of Application Guidance, which provides a roadmap for developing all portions of an application. The guidance in this ISG is limited to the portion of an advanced reactor application associated with the organization of the nuclear reactor plant applicant. Guidance regarding operational programs, conduct of operations, and procedures is not within the scope of this ISG but rather is addressed in TICAP ISG XX, Chapter 8, Programs.

The Part 53 regulation is under development and as such the guidance found in this document is subject to change based on the outcome of this rulemaking. As the 10 CFR Part 53 requirements are finalized this ISG guidance will be supplemented, as necessary, to provide guidance in the organizational and training areas to reflect any differences in requirements between Part 50/52 and Part 53. Key documents related to the Part 53 rulemaking, including preliminary proposed rule language and stakeholder comments, can be found at Regulations.gov under Docket ID NRC-2019-0062.

RATIONALE Note - this section will be updated with additional stakeholder interactions - expected during the monthly ARCAP meetings.

APPLICABILITY All holders of and applicants for a power reactor construction permit and operating license under 10 CFR Part 53. This ISG can also be used to support light water reactors (LWRs), non-LWRs, stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under 10 CFR Part 50, for a combined license (COL) or a design certification (DC) or standard design approval (SDA) under 10 CFR Part 52 or for an Early Site Permit (ESP) under 10 CFR Part 52.

GUIDANCE Design, Construction, Operating Organization - Key Management Positions An applicant should provide descriptions of the organizational structure and key management positions within the design, construction and operating organizations that are responsible for facility design, design review, design approval, construction management, testing, and operation of the plant. Acceptance criteria are based on meeting the relevant requirements of the following Commission regulations 1: 0F

  • 10 CFR 50.71 1 As the 10 CFR Part 53 requirements are finalized this ISG guidance will be supplemented, as necessary, to provide guidance in the organizational and training areas to reflect any differences in requirements between Part 50/52 and Part 53.

The applicant for a CP/OL or COL should provide the following information:

  • Organizational charts of the applicant's corporate-level management, technical support, and operations organizations, including organizational and management structure responsible for direction and support of design and construction of the proposed plant,
  • A general staffing plan for construction, startup testing,
  • Details of the interaction of design and construction within the applicant's organization and the manner by which the applicant will ensure close integration of the architect engineer (AE) and the nuclear reactor vendor,
  • Plans (preliminary for CP applicants) for the applicant's operations organization, including a general staffing plan for operations (OL and COL),
  • The relationship of the nuclear-oriented part of the organization to the rest of the corporate organization,
  • A description of the provisions for technical support for operations including interfaces between corporate, operations and the Technical Support Center (if applicable) (OL and COL),

For a DC or SDA application, the information provided should focus on the corporate level management and technical support organizations of the design organization.

For the design, construction and preoperational period (DC, SDA, CP/OL or COL), key management responsibilities in the following areas should be described:

  • Principal site-related engineering studies of the meteorology, geology, seismology, hydrology, demography, and environmental effects,
  • Design of safety-significant (i.e., safety-related and non-safety-related with special treatment) SSCs,
  • Review and approval of safety-related and safety-significant SSC design features,
  • Material and component specification review and approval,
  • Procurement of materials and equipment,
  • Management of construction activities,
  • Quality assurance activities for design and construction.

For the operational period (OL or COL), key management responsibilities in the following areas should be described

  • Nuclear, PRA, mechanical, structural, electrical, thermal-hydraulic, metallurgy and materials, and instrumentation and controls engineering (design and technical support)
  • Plant chemistry
  • Health physics
  • Fueling and refueling operations support
  • Maintenance support
  • Operations support
  • Fire protection
  • Quality assurance
  • Training
  • Safety review
  • Startup testing
  • Emergency planning
  • Security If the application is for more than one module/unit, the applicant should provide information addressing staffing plans that take into account the staggered timelines for additional modules/units scheduled to come on-line with respect to preoperational testing, fuel load, startup, and power ascension testing of each new module/unit. The applicant should describe the organizational arrangement and functions to meet the needs of the multiple modules/units.

The applicant should include in this discussion the extent to which the organizational arrangement and functions are shared between or among the modules/units addressed in the application and describe the organizational arrangement and functional divisions or controls that have been established to preserve integrity between individual modules/units and/or programs.

For plant sites with existing, operating nuclear modules/units, the applicant should include in this discussion the extent to which the organizational arrangement and functions are shared between the new and existing modules/units. In addition, the applicant should include a discussion of the organizational arrangement and functional divisions or controls that have been established to preserve integrity between the new and existing operational modules/units and/or programs.

NRC guidance regarding the operating organization is described in Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operation), which references the guidance in ANSI/ANS 3.2-2012, Managerial, Administrative and Quality Assurance Controls for the Operational Phase of Nuclear Power Plants.

Educational and Experience Requirements for Key Management Personnel The application should describe the educational and experience requirements for each key management position described above.

For a CP or COL application, the information should describe the applicants past experience in the design and construction of nuclear power plants or relevant non-power reactor design and construction experience with similar attributes to those found in the application. Experience in activities of similar scope and complexity should also be described.

The CP or COL applications should include information that demonstrates the ability of the technical staff to support or perform the safety-related activities specified in the application, as applicable. The applicant should describe the level of risk analysis experience available to perform necessary probabilistic risk assessments.

The reviewer should compare the education and experience of key personnel described above with the qualifications and experience guidance endorsed by RG 1.8, Applicable experience, i.e., work performed in a nuclear-fueled electric power production plant (commercial or military) during preoperational, startup-testing, or operational activities. Individual experience which may not be entirely applicable should be weighed against the requirements of the position.

Training for Plant Staff The NRC regulations listed below provide information pertaining to the training of nuclear power plant personnel. The OL/COL applicant should describe the training programs that are to be developed to meet these regulations. In describing compliance to these regulations, the applicant may reference in this section material discussed elsewhere in the application (i.e.,

external to the safety analysis report). 2 1F

  • 10 CFR 50.120, Training and Qualification of Nuclear Power Plant Personnel
  • 10 CFR Part 50, Licensing of Production and Utilization Facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

CP applicants should provide commitments to provide the information requested below in the OL application.

With respect to nuclear plant worker training, NRC guidance includes RG 1.8, "Qualification and Training of Personnel for Nuclear Power Plants. RG 1.8 endorses ANSI/ANS-3.1-1993, Selection, Qualification, and Training of Personnel for Nuclear Power Plants. The application should indicate the extent to which the applicable portions of this guidance is used and should justify any exceptions.

2 As the 10 CFR Part 53 requirements are finalized this ISG guidance will be supplemented, as necessary, to provide guidance in the organizational and training areas to reflect any differences in requirements between Part 50/52 and Part 53.

The training programs should focus on those tasks that are important to plant operation with regard to nuclear safety, defense-in-depth, or that are risk significant using a systems (or systematic) approach to training (SAT) as defined by 10 CFR 55.4.

The program description addressing the applicable sections of 10 CFR Part 26, Fitness for Duty Programs, should be provided in a CP/OL/COL application document separate from the safety analysis report.

Licensed Operator Training OL and COL application should provide a description and schedule of the training programs for reactor operators and senior reactor operators that meet the requirements in 10 CFR 55 with milestones for implementation during construction. Also describe the licensed operator requalification programs as required in 10 CFR 50.54(i-1) and 10 CFR 55.59, Requalification.

Describe the time when the operator requalification program will be in effect.

The licensed operator training program description should address the use of a simulator. NRC RG 1.149, Nuclear Power Plant Simulation Facilities for Use in Operator Training and License Examinations," is an acceptable approach for utilizing simulation facilities.

As an option to addressing the licensed operator training criterion, the applicant may provide a commitment to meet the guidelines of Nuclear Energy Institute (NEI)06-13A, Template for an Industry Training Program Description, for its licensed operator training program.

Non-licensed Personnel Training For OL and COL applicants, describe the training program for non-licensed nuclear plant personnel that meets the requirements of 10 CFR 50.120(b)(2) and (b)(3). Describe how the training program is derived from a systems approach to training as defined in 10 CFR 55.

The non-licensed plant staff training program should include, in addition to the technical training that is required for each non-licensed plant staff position, training in the following areas: physical security, emergency protection, radiological emergency, administrative procedures, radiation protection, fire protection, quality assurance, and fitness for duty (addressed in a separate application document).

The application should describe a program to periodically evaluate the non-licensed plant staff training programs by individuals other than those directly responsible for the training. This evaluation should include an assessment of program effectiveness in developing the trainees ability to meet performance requirements of the job. The program should be periodically revised and updated, to reflect the result of program evaluations, industry experience, and changes to the facility, procedures, regulations, and quality requirements.

The program descriptions should include the initial training, periodic retraining, and qualification that are required for non-licensed plant staff. These programs are to be established, implemented, and maintained 18 months prior to the scheduled date for initial fuel load.

As an option to addressing the non-licensed plant staff training criterion, the applicant may provide a commitment to meet the guidelines of Nuclear Energy Institute (NEI)06-13A, Template for an Industry Training Program Description, for its non-licensed plant staff training program.

Basis/number of operating shift crews, their staffing and responsibilities 3 2F Operator staffing requirements are addressed in:

  • 10 CFR 50.54(m), addressing minimum requirements per shift for on-site staffing of nuclear power modules/units by operators and senior operators licensed under 10 CFR Part 55.

Describe the functions, responsibilities, and authorities of the following plant positions (OL or COL):

  • operations supervisors,
  • operating shift supervisors/managers,
  • shift technical advisors,
  • reactor operators and senior operators,
  • non-licensed operators.

For each position listed above, describe the interfaces with offsite personnel or key management positions. Such interfaces include defined lines of reporting responsibilities (e.g.,

from the plant manager to the immediate superior), lines of authority, communication channels, and roles in risk-informed evaluations and decision making. Provide a description of the authority that may be granted to operations supervisors; to operating crew shift supervisors/managers, including the authority to issue standing or special orders; and to reactor operators and senior operators.

The application should describe the shift position titles, applicable operator licensing requirements for each, and the minimum numbers of personnel planned for each shift for all combinations of modules/units proposed to be at the station in either operating or safe shutdown mode. The applicant should also describe shift crew staffing plans unique to refueling operations. In addition, the application should describe the proposed means of assigning shift responsibility for implementing the radiation protection and fire protection programs on a round-the-clock basis where appropriate.

If the station contains, or there are plans that it contains, power generating facilities other than those specified in the application (e.g., fossil-fueled units), the applicant should describe interfaces with the organizations operating the other facilities. The description should include any proposed sharing of personnel between the units, a description of the duties of the shared personnel, and the proportion of the time these shared personnel will be assigned to the nonnuclear units.

For CP applicants, plans for staffing the operating organization may not be fully developed and staffed. It is acceptable if these plans are not fully developed, provided that the applicant either makes commitments or includes a license condition to ensure that the staffing plans are included in the OL application.

If an exemption is necessary from the licensed operator staffing requirements described in 10 CFR 50.54(m), the applicant should provide a basis for this exemption utilizing the guidance 3 As the 10 CFR Part 53 requirements are finalized this ISG guidance will be supplemented, as necessary, to provide guidance in the organizational and training areas to reflect any differences in requirements between Part 50/52 and Part 53.

contained in NUREG-1791, Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m).

However, there are practical and prescriptive limitations in NUREG-1791 that prevent it from addressing certain considerations associated with advanced reactor designs. Notably, NUREG-1791 does not address reducing licensed operator staffing levels to zero, such as might be the case for a fully autonomous advanced reactor plant design. For fully autonomous plant, extensive preapplication activities are paramount.

Acceptance Criteria 4 3F

1. The applicant has described the assignment of plant operating responsibilities; the reporting chain up through the chief executive officer of the applicant; the proposed size of the regular plant staff; the functions and responsibilities of each major plant staff group; the proposed shift crew complement for single-module/unit or multiple-module/unit operation; the qualification requirements for key members of its plant staff.
2. The applicant is technically qualified, as specified in 10 CFR 50.34, 10 CFR 50.40,10 CFR 50.48, and 10 CFR Part 50 or Part 52 (as applicable).
3. The key positions for ensuring the safe operation of the plant are in the operating organization. On-shift personnel are able to provide initial facility response in the event of an emergency.
4. The applicant has adequately described the groups and key positions responsible for implementing the initial test program and providing technical support for the operation of the facility.
5. The applicant has committed that the experience and qualifications of key members of the management and technical support organizations meet or exceed those endorsed by RG 1.8, or justified exceptions.
6. The applicants organizational requirements conform to the guidance of RG 1.33, "Quality Assurance Program Requirements (Operation)" or has provided justified exceptions.
7. An adequate number of licensed operators will be available at all required times to satisfy the minimum staffing requirements of 10 CFR 50.54(j), (m), or the applicant has provided adequate justification for an exemption. Compliance with 10 CFR 50.54(i), (j),

(k), (l), and (m) requires the applicant to demonstrate/describe how the operating organization satisfies minimum requirements for operator supervision and the availability of licensed senior operators and licensed operators during specific reactor conditions and modes of operation. Any requests for exemptions from the licensed operator staffing requirements specified in 10 CFR 50.54(m) should be justified using the guidance set forth in NUREG-1791, Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m).

4 As the 10 CFR Part 53 requirements are finalized this ISG guidance will be supplemented to provide guidance in the organizational and training areas to reflect any differences in requirements between Part 50/52 and Part 53.

8. Engineering expertise on shift should be consistent with the Commissions Policy Statement on Engineering Expertise on Shift and the guidelines of Three Mile Island (TMI) Action Plan Item I.A.1.1 of NUREG-0737.
9. The applicant has described the role and function of the AE and the NSSS vendors during design and construction and has described organizational controls over the project-related activities of the AE and nuclear reactor vendors including preservation of documentation.
10. The applicant has identified and described the reporting responsibilities and authorities in the functional areas of radiation protection/health physics, quality assurance, and training. The reporting responsibilities and authorities ensure independence from normal operating pressures.
11. The applicant has defined the responsibilities of the operating organization related to activities important to the safe operation and maintenance of the facility. Functional areas, (e.g., maintenance, operations, training, etc.), are separately supervised and/or managed.
12. Sufficient managerial depth is available to provide qualified backup for overall station operation in the event of unexpected contingencies of a temporary nature.
13. The number of licensed and non-licensed personnel for onsite shift operating crews should be sufficient to avoid the routine use of overtime.
14. The training program for licensed operators meets the requirements of 10 CFR 55.
15. The training program for non-licensed nuclear plant personnel meets the requirements of 10 CFR 50.120(b)(2) and (b)(3).

IMPLEMENTATION The staff will use the information discussed in this ISG to determine the following:

[Identify how the information will facilitate staff review of license amendments, license renewal applications, etc.]

BACKFITTING AND ISSUE FINALITY DISCUSSION

[OGC provides this discussion, but the staff can propose text for OGC consideration].

Example: The NRC staff issuance of this ISG is not considered backfitting as defined in 10 CFR 50.109(a)(1), nor is it deemed to be in conflict with any of the issue finality provisions in 10 CFR Part 52.

CONGRESSIONAL REVIEW ACT

[OGC provides this discussion to support issuance of the final ISG. However, the staff can propose text for OGC consideration].

Example: This ISG is a rule as defined in the Congressional Review Act (5 U.S.C. §§ 801-808).

However, the Office of Management and Budget has not found it to be a major rule as defined in the Congressional Review Act.

FINAL RESOLUTION By [insert date], this information will be transitioned into [identify the appropriate regulatory process (Standard Review Plan (SRP), Regulatory Guide (RG))]. Following the transition of this guidance to the [SRP, RG], this ISG will be closed.

APPENDIX A. Resolution of Public Comments APPENDIX A Resolution of Public Comments A notice of opportunity for public comment on this Interim Staff Guidance (ISG) was published in the Federal Register (insert FR Citation #) on [date] for a 30-60 day comment period. [Insert number of commenters] provided comments which were considered before issuance of this ISG in final form.

Comments on this ISG are available electronically at the NRC's electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into ADAMS, which provides text and image files of NRC's public documents. Comments were received from the following individuals or groups:

Letter No. ADAMS No. Commenter Affiliation Commenter Name Abbreviation 1

2 3

4 5

The comments and the staff responses are provided below.

Comment 1: [Each comment summary must clearly identify the entity that submitted the comment and the comment itself].

NRC Response: Comment responses should begin with a direct statement of the NRC staffs position on a comment, e.g., the NRC staff agrees with the comment or the NRC staff disagrees with the comment.

x If the NRC staff agrees, explain why and provide a clear statement as to how the relevant language was revised or supplemented to address the comment. Include the following language at the end of the comment response: The final ISG was changed by <describe the change; if necessary by quoting the newly revised language>.

x If the NRC disagrees with a comment and no change was made to the generic communication, then explain why and provide the following language at the end of the comment response: No change was made to the final ISG as a result of this comment.

APPENDIX B References

ProposedARCAPDocumentStructure Legend PrimaryportionsderivedfromTICAP Primaryportionsderivedfromseparateongoingregulatoryactivities Version CombinationofnewTICAPandARCAP 2/17/2021 NewARCAPguidancebeingdeveloped

  • Guidancereferencedinthedevelopedcolumnisprovidedforconsiderationandmaynotalwaysbeapplicableforagivendesign.
  1.  Needed Disposition TICAP(Industryled) TICAP/ARCAP(NRCled)

Developed* AdditionalActivities**

Overviewoftechnology(sizeofthe Generaldescriptionoftheplantsystems Willthischapterincludegeneralsite NEI1804 TICAP Includesgenericdescriptionofsafety

reactorandplannedcommercial (Baselineoperatingparameters). characteristicstypicallydescribedfor RG1.233andRG1.232 ARCAPdeveloping casefordesign.

applicationofthedesignpower Generalsitecharacteristics sites(i.e.ARCAPCh.2)? Commission's2008 varioussubsections Commissionstatementcanbefoundat

production,industrialapplication,etc.) oIntroduction "PolicyStatementon FRNVol.73,No.199,10/14/2008 Generaldescriptionoftheplant oSiteCharacteristicsandSiteParameters Industrywouldprefertoseegeneralsite Adv.Reactors" NEI1804,RG1.232andRG1.233are

systemsandrolesthattheyplayin oGeographyandDemography informationinChapter1butnotthe TMIRequirements10 onlymentionedoncebutareapplicableto

normalandoffnormalconditions, oNearbyIndustrial,Transportation,and detailedinformationsupportingthe CFR50.34(f) allproposedARCAPdispositionscolored

includingrefueling MilitaryFacilities DesignBasisExternalHazardLevels.That NUREG0933GSIsand greenandblue.

ProposedFSARChapters

1 oRegionalClimatology,andLocal couldpotentiallygoinChapter2. USIs Meteorology,andAtmospheric

Dispersion(BasisforSection2.3below) IndustrynotesthatNRCARCAPis

oHydrologicalDescription workingonreducingthelevelofdetailin

oGeology,Seismology,andGeotechnical theSARrelatedtothesite(whichis

Engineering generallyacknowledgedtobetoo

detailed).

Page 1

  • SummaryofSafetyCaseFindings RG1.70for SECY200045

oOverviewofaffirmativeLMPbased geographicalformat "PopulationRelated

safetycasemethodology,including RG4.7(awaiting SitingConsiderationsfor

referencetoNEI1804andany Commissiondirection Adv.Rxs" deviationsfromtheapproved onSECY200045) DG4028"Volcanic

Ch.1GeneralPlantInformation,Site

methodology RG1.91 HazardsAssessmentsfor

Description,andOverviewoftheSafety

oSummaryofFSFs RG1.76 ProposedNPPs" Case

oSummaryofLBEswithfocusonDBAs RG1.221 RESGuidanceonRIPB

oSummaryofradiologicalconsequence RG1.27 ApproachtoSeismic

assessment RG1.23 Safety oSummaryofhowthedesignprovides RG1.145 NonLWRMELCOR

thatFSFsaremetkeyplantattributes RG1.111 DemonstrationProject anddesignfeaturesthatprovide RG1.194 RG1.59intheprocess

reasonableassuranceofadequate RG1.59 ofbeingupdated.In

2 protectionofpublichealthandsafety RG1.102 additionstaffdeveloping

oEvaluationofDIDcapabilities NUREG2115 RGassociatedwithdam

ANS/ANSI2.272020 safetyreviews Industryconsidersthatthedetailed ANS/ANSI2.292020 informationshouldbeprovidedin NUREG2213 subsequentchapters.AnLBEbyLBE ANSI/ANS2.262004 discussionisnotintended,forexample; ASCE/SEI4305 thatisprovidedinChapter3. RG1.132

RG1.138

USACEEM11102 1902

RG1.198 2.5-ExternalHazardsEvaluation Chapter2statestheapplicantmay RG1.200 RGendorsingnonLWR Thisusedtobethe"PRAOverview"

IndustrynotesthatNRCARCAPis provideinformationabout PRAStandard Section12inthepreviousversion,itis

workingonreducingthelevelofdetailin additionalgenericanalysisusedin TICAPCh.2andARCAP nowCh.2 theSARrelatedtothesite(whichis subsequentsubsections.Whatdoesthis Ch.12 generallyacknowledgedtobetoo mean?Provideexamplestoclarifythe

detailed). typesofanalysisandpotential

2.6-AnalysesofSystems,Components, subsections.

andMaterialsPerformance Beyondthoseexamplesalreadyprovided

2.1ProbabilisticRiskAssessment Needsadditionaldiscussion.Industry (e.g.,sourceterm),exampleswould

oOverviewofPRA notesthatSSCspecificinformationis likelybetechnologyspecific.See

oSummaryofKeyPRAFindings Ch.2-GenericAnalyses proposedtobeprovidedinChapters6 computercodeexampleunderSection

2.2-SourceTerm and7. 2.7.

2.3-Meteorology 2.7-AnalyticalCodes 2.4-OtherGenericAnalyses Needsadditionaldiscussion.Tothe

extentcodesneedtobediscussedinthe

SAR,thiscanbedoneintheLBEsection

(Chapter3)orviareferencetoatopical

report.Totheextentacomputercodeis

akeyanalyticaltoolforseveralanalyses,

itmightbeappropriatetoaddressitin

thischapter.

Page 2

3.1LicensingBasisEventSelection Wherewilltheaircraftandlossoflarge SECY160012, IAPStrategy2Code ThisChapternowincorporatesDBA

Methodology areaanalysisbedescribed?Needs AccidentSourceTerms Assessmentsupport Analysis,whichwaspreviouslyCh.6 3.2AnticipatedOperational additionaldiscussion.Theissueforthis andSitingForSmall TICAPCh.2and3 Occurrences chapterappearstobetheinclusionor ModularReactorsAnd NonLWRMELCOR

3.3DesignBasisEvents notofthoseeventsforwhich NonLightWater DemonstrationProject 3.4BeyondDesignBasisEvents assessementsarerequiredbutwhichdo Reactors. TICAPCh.3 3.5DesignBasisAccidents notevolvefromtheLMPprocess. RG1.217

3 Ch.3-LicenseBasisEventAnalysis Specifically,aircraftimpact(50.150)and NEI0713,

lossoflargearea(LOLA)(50.155).If Methodologyfor

theseassessmentsareincludedin PerformingAircraft

Chapter3,theyshouldbeinaspecific ImpactAssessmentsfor

sectionoutsideoftheLMPderived NewPlantDesigns events.Industryconsideringanew RG1.203 Section3.6entitled"SpecialEvent

Analyses."

4.1EvaluationofIntegratedPlantRisk 4.2DefenseinDepth Industryisconsideringdevelopingthe RG1.145 TICAPCh.4 ThisChapternowincorporates"Defense

4.2DefenseinDepth 4.2.1-PlantCapabilityDID necessaryguidanceforboth4.1and4.2. inDepth"whichwaspreviouslyCh7on

4 Ch.4-IntegratedEvaluations 4.2.2-ProgrammaticDID ItissuggestedthatARCAPcommenton theTable thatguidancewhenprovidedratherthan

developingitindependently.

5.1PrincipalDesignCriteriaandSafety *SRandNSRSToperatoractions-input Chapter5statesthatSRandNSRST SECY180096 TICAPCh.5,6,and7 RelatedSSCs todeterminingstaffinglevels,I&Csafety operatoractionswillbeidentified.What "Functional RGendorsingASMESec

  • RequiredSafetyFunctions categories,humanfactorsanalysis, istobedonewiththese? Containment III,Div5"HighTemp
  • RequiredFunctionalDesignCriteria training,etc Industryconsideringlinkingto PerformanceCriteriafor Materials" 5.2ComplementaryDesignCriteriaand *Validationofequipmentqualifications supportingplantprogramsthatprovide nonLWRs" RGendorsingASMESec

NonSafetyRelatedwithSpecial -wherewillthequalificationof reasonableassurancethattheactions RG1.201 XI,Div2"Reliability

TreatmentSSCs equipment(seismic,environmental)be willbeaccomplished. RG1.129 IntegrityManagement"

  • RiskSignificantSafetyFunctions addressed? RG1.100 FuelQualification

IndustrynotesthatEquipment NUREG0800(SRP)Sec. Guidance(whitepaper

qualificationisaspecialtreatment 4.2 andsubsequentNUREG)

Ch.5-SafetyFunctions,Design associatedwithSSCsseeChapters6 ATFISG202001 TopicalReportonTRISO

5 Criteria,andSSCCategorization and7. fuel

  • Descriptionoftheanalyticalcodes(TH, DRGforI&CReviews

reactorphysics,fuelperformance)used MSRFuelQualification

inthesafetyanalysisandhowtheywere Guidance validatedbeprovided.

Industryconsidersthattotheextent

informationoncomputercodesis

neededintheSAR,itwouldbeprovided

inChapter3withtheassociatedLBEorin

Chapter2,ifacrosscuttingtool.

Page 3

  • SummaryofDBEHLs *BasisforTechSpecallowableoutage *Additionaldiscussionsarestillneeded
  • SafetyRelatedDesignCriteria timesandproposedLCOs ontheLoDdocument(2page
  • Reliabilityandcapabilityperformance document).TheLoDconsiderationsmay

basedtargets applyonmultiplechapters.

  • SpecialTreatments
  • DBEHLrelateddesignrequirementsfor WRTSRdesigncriteria,howarethese

nonsafetyrelatedSSCs relatedtotheComplementaryDesign

  • SystemdescriptionsforSRSSCs CriteriaandthePDCs?
  • WilltheTICAPchaptersinclude

acceptancecriteriasimilartoARCAP?

Ch.6SafetyRelatedSSCCriteriaand

Industryconsidersthattotheextent

Capabilities reasonableandpracticalfortechnology inclusiveguidance,andwithanemphasis

onperformancebasedcriteria.

  • WherewilltheLBEcomparisontotheF Ccurvebedescribed?Chapter3.
  • Wherewillthedesignparametersfor

theSSCsbedescribed?Industrynotes

thatthisinformationwillbeprovidedin

TICAPChapters6and7.

Ch.7NSRSTSSCCriteriaand *SpecialTreatments *Basisforallowableoutagetimes Capabilities *SystemdescriptionsforNSRSTSSCs

  • HumanFactors *Maintenance NUREG0711 Thischapternowincorporatesthe
  • Training *ChangeControlNeedsadditional NUREG1275 previous"HumanFactors"chapter10
  • ReliabilityAssurance discussion.Itisnotclearwhatis fromthepreviousversion Applicableplantprogramsareverymuch intendedhere.Somethingbeyond

Ch.8PlantPrograms afunctionofthetechnologyandthe 50.59?

affirmativesafetycase.Thethreecited *ConductofOperations abovewereintendedtobeexamples.

9.1LiquidandGaseousEffluents NUREG0800(SRP) ARCAPCh.8 ARCAPteamdevelopeddraftguidance

9.2-ContaminationControl Secs.11.2,11.3,and thatdiscussesaperformancebased

Ch.9-ControlofRoutinePlant 9.3-SolidWaste 11.4 approach.Thedraftguidancehasbeen

8 RadioactiveEffluents,Plant RG1.109andRG1.111 wellreceivedbystakeholdersinpublic

Contamination,andSolidWaste RG4.21 meetings.Teamisfurtherrefiningthe

NEI0710A approach.ISGtobedeveloped

RG8.8 ARCAPCh.9 ARCAPteamtodevelopeddraftguidance

RG8.10 basedonFSARChapter8.ISGtobe

9 Ch.10-ControlofOccupationalDose

ANSI/ANS18.11999 developed

NEI0708A 11.1Description/responsibilitiesofkey

managementpositions 11.2Educational,trainingand

experiencerequirementsforkey

managementpositions Ch.11Organization 11.3Interfaceswithsupportgroups(e.g.

TechnicalSupportCenter,Corporate) 11.5Basis/numberofoperatingshift

crews,theirstaffingandresponsibilities Page 4

12.1Asbuiltverificationprogram NUREG0800(SRP)Sec.

(ITAAC) 14.2 12.2Preoperationaltestingprogram 12.3Initialstartuptesting/operations

program 14 Ch.12-InitialStartupPrograms

TICAPwillhaveamajorimpacton

technicalspecifications.NRCandINL

haveidentifiedtheneedforTICAPto

considertechspecdevelopmentaspartof

TICAP.Unclearatthispointhowmuch

TICAPguidancewillbeprovidedinthis

area.TechSpecsguidancewillalsobe

15 TechnicalSpecification influencedbythefinaltextofSubpartBof

thefinalPart53rule.

Thiswarrantssomediscussion.Industry

seestheconnectionviaspecialtreatments

AdditionalContentsofApplication forSSCs.However,underPart50/52,

Industrydoesnotseeaneedtorevisit

existingguidanceforTechSpecs.

Existingguidanceinthisareaneedstobe

16 TechnicalRequirementsManual

adjustedtoreflectLMPterminology TICAPoutcomesexpectedtoheavily

influencequalityassuranceplanforthe

design.AppendixBexpectedtoapplyto

safetyrelatedSSCs.Unclearatthispoint

howTICAPwilladdressQAforNonsafety

17 QualityAssurancePlan(design)

relatedspecialtreatmentSSCs IndustrydoesnotplanspecialTICAP

guidance.NSRSTSSCswillhavetheoption

toinvokeelementsoftheQAprogramas

specialtreatments.

RG1.189 ResultsofTICAPdevelopedaffirmative

safetycaseexpectedtoinfluencefire

18 FireProtectionProgram(design) protectionprogram Nospecialguidanceisplannedaspartof

IndustrydevelopedTICAP.

RG1.200 RGendorsingnonLWR SeeFSARChapter12

PRAStandard IndustryconsideringTICAPguidanceon

19 ProbabilisticRiskAssessment PRAsummarytobeprovidedinChapter2.

RG1.28 QAPlanforsodium TICAPoutcomesexpectedtoheavily

RG1.30 cooledFASTMetallicFuel influencequalityassuranceplanforthe

RG1.33 DataQualification design.AppendixBexpectedtoapplyto

QualityAssurancePlan(Construction

20 RG1.164 safetyrelatedSSCs.Unclearatthispoint

andOperations)

howTICAPwilladdressQAforNonsafety

relatedspecialtreatmentSSCs SeeabovecommentonQA(Design).

Page 5

NUREG0396 SECY180103relatedto EPrulemakingexpectedtodevelop

21 EmergencyPlan NUREG0654 EPforSMRsandother guidanceinthisarea RG1.101 technologies SECY180075relatedto Physicalsecurityrulemakingexpectedto

22 PhysicalSecurityPlan ConsequenceBased developguidanceinthisarea Security 23 SNMphysicalprotectionplan

MC&Aisanissuethathasidentifiedas

needingtohaveguidancedevelopedfor

someofthenonlwrs.Apebblebed

SNMmaterialcontrolandaccounting MC&Aapplicationstandardandreview

24 plan standardhasbeendevelopedbyORNL.

MC&Aforliquidfueledmoltensalt

reactorswillbeaparticularchallenge.

DoesNUREG2159apply?

RG1.189 ResultsofTICAPdevelopedaffirmative

safetycaseexpectedtoinfluencefire

25 FireProtectionProgram(Operational) protectionprogram Nospecialguidanceisplannedaspartof

IndustrydevelopedTICAP.

RelatestoARCAPChapter9abovemore

26 RadiationProtectionProgram specificguidancebeingconsidered.

RelatestoARCAPChapter9abovemore

27 OffsiteDoseCalculationManual specificguidancebeingconsidered.

RG1.17 TICAPoutcomesexpectedtoheavily

RG1.178 influenceISI/IST.InadditionASME

SectionXISection2guidanceidentifiedas

needingtobedeveloped.

InserviceInspection/Inservicetesting

28 ThelinktoindustrydevelopedTICAPis

(ISI/IST)

throughspecialtreatmentrequirements.

IndustrydevelopedTICAPdoesnotintend

todevelopguidanceondocumenting

ISI/ISTprograms.

RG4.2 EnvironmentalISGfor

EnvironmentalReportandSiteRedress NUREG1555 MicroReactors 29 Plan COL/ESPISG026 DraftGEISforAdv.Rxs COL/ESPISG027 FinancialQualificationandInsurance Reportunderdevelopmenttoaddress

30 andLiability issues RG5.71 UnclearatthispointhowmuchTICAP

guidancewillbeprovidedinthisarea 31 CyberSecurityPlan IndustryledTICAPisnotplanningto

developcybersecurityspecificguidance

(subsetofsecurityguidance).

Page 6

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