ML23234A085

8/23/2023 EI-NI Independence ACRS Presentation Slides (99902100)

ML23234A085
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Site:	99902100
Issue date:	08/22/2023
From:	Mallecia Sutton NRC/NRR/DANU/UAL1
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Download: ML23234A085 (28)

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Staff Review of NATD-LIC-RPRT-0001, Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces Mallecia Sutton, Senior Project Manager, NRR/DANU Reed Anzalone, Senior Nuclear Engineer, NRR/DANU Jesse Seymour, Senior Reactor Engineer (Examiner), NRR/DRO

Agenda

• Review Chronology

• TR Overview and Review Strategy

• Safety Evaluation Overview

• Plant design and transients

• Safety classification using NEI 18-04

• Regulatory evaluations

• Limitations and conditions

• Conclusions 2

Review Chronology

• Pre-Application Public Meeting

• November 17, 2021

• Submittals on Decoupling Strategy

• White Paper: Energy Island Decoupling Strategy February 4, 2022 (voluntarily withdrawn)

• Topical Report: Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces October 4, 2022

• TR Accepted

• November 16, 2022

• Audit Conducted

• January 23, 2023 - March 10, 2023

• Final Draft Safety Evaluation Issued

• August 10, 2023 3

TR Overview and Review Strategy

• Purpose of TR:

• The independence of operation between the systems contained within the NI and the plant systems composing the EI is a key aspect of the Natrium design philosophy.

The NI boundary conditions have been intentionally designed so the interrelationship with the EI does not impact the NI safety case.

• [E]valuate regulatory impacts of the Natrium design interfaces with respect to the interaction of NI and EI systems

• Review Strategy

• Examine key aspects of Natrium design and analysis presented in the TR

• Assess regulatory evaluations in the context of these key aspects

• Propose appropriate limitations and conditions necessary for evaluations to be acceptable 4

Safety Evaluation Overview

• Sections 1 Introduction and Background

• Section 5 - Staff Evaluation

• Section 5.1 - Natrium Plant Design and Transients

• Section 5.2 - Natrium Safety Classification of SSCs

• Section 5.3 - Regulatory Analyses

• Section 6 - Limitations and Conditions

• Section 7 - Conclusions 5

Plant Design and Response to Transients SHX NI EI PHT IHT IAC RAC TSS SGS Images sourced from NATD-LIC-RPRT-0001, Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces (ML22277A824) 6 6

Safety Classification Using NEI 18-04*

• NEI 18-04 endorsed in RG 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors

• Risk-informed, performance-based safety classification is integrated with other aspects of NEI 18-04 process, including selection and analysis of licensing basis events and evaluation of defense-in-depth adequacy

• Different definition for safety-related SSCs than 10 CFR 50.2

• NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1 (ML19241A472) 7

Safety Classification Using NEI 18-04 From NEI 18-04, Rev. 1 (ML19241A472) 8

Role of F-C target in Safety Classification DBEs High Consequence BDBEs Only SR SSCs available to keep DBAs below 10 CFR 50.34 dose limit From NEI 18-04, Rev. 1 (ML19241A472) 9 9

Overview of Regulations Covered in TR

• 10 CFR 50.10, License required; limited work authorization

• 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants

• 10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

• 10 CFR 55, Operators Licenses 10

10 CFR 50.10 (LWA rule) Overview

• 10 CFR 50.10(c): [n]o person may begin the construction of a production or utilization facility on a site on which the facility is to be operated until that person has been issued either a [construction permit] or a limited work authorization [LWA].

• 10 CFR 50.10(a)(1) provides scoping criteria defining construction.

• Rule issuance FRN (72 FR 57415) notes that scoping criteria were chosen to encompass those SSCs that have a reasonable nexus to radiological health and safety or common defense and security.

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10 CFR 50.10(a)(1)

Activities constituting construction are the driving of piles, subsurface preparation, placement of backfill, concrete, or permanent retaining walls within an excavation, installation of foundations, or in-place assembly, erection, fabrication, or testing, which are for:

(i) Safety-related structures, systems, or components (SSCs) of a facility, as defined in 10 CFR 50.2; (ii) SSCs relied upon to mitigate accidents or transients or used in plant emergency operating procedures; (iii) SSCs whose failure could prevent safety-related SSCs from fulfilling their safety-related function; (iv) SSCs whose failure could cause a reactor scram or actuation of a safety-related system; (v) SSCs necessary to comply with 10 CFR part 73; (vi) SSCs necessary to comply with 10 CFR 50.48 and criterion 3 of 10 CFR part 50, appendix A; and (vii) Onsite emergency facilities, that is, technical support and operations support centers, necessary to comply with 10 CFR 50.47 and 10 CFR part 50, appendix E.

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10 CFR 50.10(a)(1) Evaluation Criterion TerraPower Staff (i)

Not applicable because EI SSCs are NST.

Reasonable, consistent with NEI 18-04 safety classification definition. See L&C 3 regarding definition of safety-related.

(ii)

Not applicable because EI SSCs are not used to mitigate accidents or transients or used in EOPs.

Consistent with plant design. Would not expect NST SSCs to participate in mitigation or prevention.

Not enough information on EOPs. See L&C 5.

(iii)

Not applicable because NST SSCs would not be capable of preventing SR SSCs from fulfilling safety functions.

Consistent with NEI 18-04 safety classification definition.

(iv)

Applicable because failure of EI SSCs could eventually cause a reactor trip.

Plan to seek exemption. Exemption basis would be the same for 10 CFR 50.10(a)(1)(iv) and 10 CFR 50.65(b)(2)(iii).

Agree with determination that criterion is applicable.

Reasonable to use same exemption basis for 10 CFR 50.10(a)(1)(iv) and 10 CFR 50.65(b)(2)(iii). Not taking a position on prospective exemptions. See L&C 10.

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10 CFR 50.10(a)(1) Evaluation Criterion TerraPower Staff (v)

Not applicable because no physical security program SSCs are on EI; CDAs will not be installed on EI prior to CP.

Appears consistent with design, but not enough information to support. See L&C 6.

(vi)

Not applicable because fires on EI will not prevent ability to maintain and achieve safe shutdown.

Adequate because EI SSCs are NST and safe shutdown can be achieved and maintained solely with NI systems.

(vii)

Not applicable because onsite emergency facilities will not be located on EI.

Consistent with plant design, but not enough information to support. See L&C 7.

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10 CFR 50.65 (Maintenance Rule)

• Requires licensees to have a program that monitors the performance or condition of certain SSCs or demonstrates the performance or condition of these SSCs through appropriate preventative maintenance, to provide reasonable assurance that they are capable of fulfilling their intended functions.

• Scope of SSCs considered under the Maintenance Rule are in 10 CFR 50.65(b)

• 10 CFR 50.65(b)(1) is similar to 10 CFR 50.10(a)(1)(i)

• 10 CFR 50.65(b)(2)(i) through (iii) are identical to 10 CFR 50.10(a)(1)(ii) through (iv) 15

10 CFR 50.65 Evaluation

• TerraPower did not evaluate 10 CFR 50.65(b)(1).

• TerraPower determined criteria (i) and (ii) are not applicable. Staff discussions and conclusions are the same as for 10 CFR 50.10(a)(1)(ii) and (iii).

• TerraPower determined criterion (iii) is applicable and plans to seek an exemption. Staff discussion and conclusions are the same as for 10 CFR 50.10(a)(1)(iv).

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10 CFR 50, Appendix B

• Appendix B provides QA requirements for the design, manufacture, and construction of certain SSCs.

• Appendix B applies to all activities affecting the safety related functions of SSCs that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public.

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10 CFR 50, Appendix B Evaluation

• TerraPower asserted that all EI SSCs will be NST, and SSCs classified as NST under the NEI 18-04 process are not capable of affecting the SR functions of SSCs used for prevention or mitigation.

• Staff considered the role of NST SSCs and determined TerraPowers evaluation was acceptable

• RG 1.233 Staff Position C.2 states, in part: The staff expects that SSCs that provide essential support (including required human actions) for SR or NSRST SSCs will be classified in a manner consistent with the higher-level function, even if the supporting SSC is not explicitly modeled in the PRA.

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10 CFR 55

• TerraPower states that the Natrium design removes direct interaction between the reactor and the turbine generator

• Turbine operations would not be an apparatus or mechanism whose manipulation directly affects the reactivity or power level of the reactor

• Natrium design should allow for a non-licensed individual to fully operate the turbine generator

• Staff evaluated that the nature of the TSS is consistent with Natrium not treating turbine generator operations as a control as defined under 10 CFR 50.2 and 10 CFR 55.4

• A key consideration is the term direct and its specific meaning within the definition of controls 19

10 CFR 55

• The AEA Section 11 defines operators as individuals who manipulate the controls of utilization facilities; definition of controls is left to the NRC

• Section 107 further mandates that individuals who operate utilization facility controls must be licensed by the NRC.

• Manipulation of the controls is restricted to licensed operators under 10 CFR 50.54(i)

• Definition of controls in Parts 50 & 55 dates back to 1963 (28 FR 3197)

• Narrowed from an earlier, broader definition; FRN notes that this [current] narrower interpretation is in accord with the original Commission intent.

• The direct aspect of a control means they cause reactor power level or reactivity changes without needing something intermediate to make that happen

• Natrium TSS is a significant intermediary from a reactivity standpoint 20

L&C 1 - Key Aspects of Natrium Design

• Applicants referencing TR must use a plant design that is substantially similar to what was discussed in TR. Any deviations from plant design discussed in TR that could affect SE conclusions must be justified when the TR is referenced.

• Staff underscores the importance of the design features that enable the independence of the NI and EI. While the regulatory evaluations may be relatively high-level, the staffs determinations were made in the context of the Natrium design and its capabilities.

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L&Cs 2&3 - Safety Classification; Definition of Safety-Related

• L&C 2

• Staffs conclusions rely on the use of the NEI 18-04 process for safety classification and a determination that all EI SSCs are classified as NST.

• Design process is iterative; so is NEI 18-04. As design matures, TerraPower will need to ensure EI SSCs continue to be NST for conclusions to be applicable.

• L&C 3

• Definition of SR in NEI 18-04 is different from that in 10 CFR 50.2.

• Some regulations explicitly reference 10 CFR 50.2 definition rather than simply saying safety-related.

• Use of NEI 18-04 may require exemptions from 10 CFR 50.2 and regulations that refer to it.

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L&Cs 4 Evaluation of 10 CFR 50.10 and 10 CFR 50.65

• L&C 4 - TerraPower did not address definitions of construction outside of 10 CFR 50.10; this limitation makes the scope of the TR clear.

• L&C 5 - Staff has not reviewed EOPs for Natrium and TerraPower did not discuss in sufficient detail to ensure they do not rely on EI SSCs.

• L&Cs 6&7 - TerraPower did not discuss physical and cyber security programs or onsite emergency facilities in sufficient detail for staff to ensure the EI does not include SSCs that fall under the scope of 10 CFR 50.10(a)(1)(v) and (a)(1)(vii).

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L&Cs 8 Evaluation of 10 CFR 55

• L&C 8 - 10 CFR 50.54(j) not addressed in TR.

• L&C 9 - Since manipulating the EI does not directly change reactivity or power, operator licensing experience requirements under 10 CFR 55.31(a)(5) are not satisfied by operating the EI.

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L&C 10 - Prospective Exemption Requests

• TR discussed basis for proposed exemptions.

• Staff reviewed for information but did not take a position on prospective exemptions in the SE.

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Conclusions TR is acceptable for referencing in future licensing submittals, subject to limitations and conditions.

Plant design and EI SSC safety classification [assumptions] are integral parts of staffs evaluation.

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Abbreviations BDBE - Beyond design basis event CFR - Code of Federal Regulations CP - Construction permit CDA - Critical digital assets DBA - Design basis accident DBE - Design basis event EI - Energy Island EOP - Emergency operating procedure F-C - Frequency-consequence FR - Federal Register FRN - Federal Register notice IAC - Intermediate air cooling system IHT - Intermediate heat transport system L&C - Limitation and/or condition LWA - Limited work authorization NEI - Nuclear Energy Institute NI - Nuclear island NSRST - Non-safety related with special treatment NST - Non-safety related with no special treatment PHT - Primary heat transport system QA - Quality assurance RAC - Reactor air cooling system RG - Regulatory guide SGS - Steam generation system SHX - Sodium/salt heat exchanger SSC - Structure, system, or component SE - Safety evaluation SR - Safety related TR - Topical report TSS - Thermal salt storage system 27

From NEI 18-04, Rev. 1 (ML19241A472) 28