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LLC Revision 1 to Standard Design Approval Application, Part 2, Chapter 16, TR-101310-NP, US460 Standard Design Approval Technical Specifications Development
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US460 Standard Design Approval Technical Specifications Development

TR-101310-NP Revision 0 Licensing Technical Report

US460 Standard Design Approval Technical Specifications Development

December 2022 Revision 0 Docket: 52-050

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

US460 Standard Design Approval Technical Specifications Development

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COPYRIGHT NOTICE

This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copi es of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a licens e, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations.

Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.

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Department of Energy Acknowledgement and Disclaimer

This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights.

Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not ne cessarily state or reflect those of the United States Government or any agency thereof.

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Table of Contents

Abstract................................................................... 1 1.0 Introduction.......................................................... 2 1.1 Purpose.............................................................. 2 1.2 Scope................................................................ 2 1.3 Abbreviations.......................................................... 3 2.0 Background.......................................................... 5 2.1 Approach............................................................. 5 2.2 Regulatory Requirements................................................ 6 2.3 Design Specific Review Standard.......................................... 6 3.0 Changes to the Content of Standard NuScale Technical Specifications......... 7 3.1 Modifications to Chapter 1, Use and Application............................... 7 3.1.1 1.1 Definitions.................................................... 7 3.1.2 Sections 1.2 through 1.4........................................... 8 3.2 Modifications to Chapter 2, Safety Limits..................................... 9 3.3 Changes to Chapter 3, Limiting Conditions for Operation and Surveillance Requirements.......................................................... 9 3.3.1 Modification of Limiting Condition of Operation 3.0.3...................... 9 3.3.2 Addition of Surveillance Requirement 3.1.9.5 - Verification of Isolation of Module Heatup System from Other Modules............................ 9 3.3.3 Modification of Limiting Condition of Operation 3.2.1, Enthalpy Rise Hot Channel Factor (FH), and 3.2.2, Axial Offset........................... 9 3.3.4 Changes to Limiting Condition of Operation 3.3.1, Module Protection System......................................................... 9 3.3.5 Addition of Limiting Condition of Operation 3.3.3 Condition for Pressurizer Line Isolation Inoperability......................................... 11 3.3.6 Addition of Surveillance Requirement 3.3.3.3, Monitoring Emergency Core Cooling System Actuation Time Delay................................ 11 3.3.7 Removal of Limiting Condition of Operation 3.3.5, Remote Shutdown Station........................................................ 11 3.3.8 Modification of Limiting Condition of Operation 3.4.2, Minimum Temperature for Criticality.................................................... 12 3.3.9 Modification of Surveillance Requirement 3.4.4.1 - Reactor Safety Valve Setpoints...................................................... 12 3.3.10 Editorial Clarification of Condition D of Limiting Condition of Operation 3.4.3, Reactor Coolant System Pressure / Temperature Limits............. 12

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Table of Contents

3.3.11 Modification of Limiting Condition of Operation 3.4.5, Reactor Coolant System Operational Leakage....................................... 12 3.3.12 Modification of Limiting Condition of Operation 3.4.8, Reactor Coolant System Specific Activity........................................... 13 3.3.13 Modification of Limiting Condition of Operation 3.4.10, Low Temperature Overpressure Protection Valves..................................... 13 3.3.14 Modification of Limiting Condition of Operation 3.5.1, Emergency Core Cooling System................................................. 13 3.3.15 Modification of Limiting Condition of Operation 3.5.3, Ultimate Heat Sink..... 13 3.3.16 Addition of Limiting Condition of Operation 3.5.4, Emergency Core Cooling System Supplemental Boron....................................... 14 3.3.17 Addition of Limiting Condition of Operation 3.6.3, Containment Closure...... 14 3.3.18 Removal of Limiting Condition of Operation 3.7.3, In-Containment Secondary Piping Leakage........................................ 14 3.3.19 Other Bases Changes............................................ 15 3.4 Chapter 4, Design Features.............................................. 15 3.5 Chapter 5, Administrative Controls........................................ 15 4.0 Conformance with Industry Standard Technical Specifications and Standard Technical Specification Writers Guide................................... 17 5.0 References.......................................................... 19 5.1 Referenced Documents................................................. 19

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List of Tables

Table 1-1 Acronyms....................................................... 3 Table 1-2 Definitions....................................................... 3 Table 4-1 Standard Technical Specifications Traveler Adaptation................... 18

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List of Figures

Figure 3-1 MODES 2 and 3 Illustration......................................... 8

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Abstract

This report describes the development process of the NuScale Power Plant US460 Standard Design Approval (SDA) technical specifications (T S) to conform with regulatory requirements and expectations regarding scope, content, and forma t. This report also provides the basis for including the requirements chosen for the NuScale TS.

This report describes the development process with emphasis on the differences from the NuScale US600 Design Certification Applicatio n (DCA) Technical Specifications, Revision 5 (Reference 5.1.1 and Reference 5.1.2). The development process for the NuScale DCA Technical Specifications is described in TR-1116-52011-NP, Revision 4 (Reference 5.1.3).

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1.0 Introduction

1.1 Purpose

The purpose of this report is to outline the development process of the NuScale Power Plant technical specifications (TS) to conform with the applicable regulatory requirements and expectations regarding scope, content, and format. This report also provides the basis for including the specifications chosen for the NuScale TS. The report focuses on three aspects:

Conformance with Title 10 Code of Federal Regulations Section (10 CFR) 50.36 (Reference 5.1.4) and 10 CFR 50.36a (Referenc e 5.1.5), including considerations in 58 FR 39132, Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Reference 5.1.6).

Conformance with regulatory expectations as expressed by the precedent established in the NuScale US600 Design Certification Application (DCA) TS (Reference 5.1.1 and Reference 5.1.2). Consideration of Nuclear Regulatory Commission (NRC)-published standard technical specifications (STS), approved generic technical specifications (GTS), and recent changes as delineated in industry change travelers is also described where applicable.

Conformance with the technical specification format and content guidance established by TSTF-GG-05-01, Revision 1, Writers Guide for Plant Specific Improved Technical Specifications, August 2010 (Reference 5.1.7).

The report supplements the descriptions in TR-1116-52011, Rev. 4 (Reference 5.1.3) describing the development of the NuScale US600 DCA technical specifications and Bases. It identifies significant changes from the US600 DCA technical specifications requirements, and identifies NRC/industry Impr oved Standard Technical Specifications (ISTS) travelers considered in the development of the US460 SDA technical specifications.

1.2 Scope

This report addresses the development of the TS applicable to an individual module installed in a US460 NuScale power plant. The NuScale US460 SDA technical specifications are drafted in the context of the US460 SDA application for a NuScale facility containing up to six modules, however the content is applicable to an individually licensed module. The content of this report and the NuScale GTS are generally applicable to any NuScale facility containi ng any number of individually licensed modules with variations to address facility specific design details.

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1.3 Abbreviations

Table 1-1 Acronyms Term Definition ADAMS (NRC) Agencywide Document s Access and Management System BWR boiling water reactor CFR Code of Federal Regulations COLR core operating limits report CRA control rod assembly CRDS control rod drive system CVCS chemical and volume control system DCA Design Certification Application fo r the NuScale US600, 12-module plant design DHRS decay heat removal system ECCS emergency core cooling system ESB ECCS supplementary boron FSAR Final Safety Analysis Report GTS generic technical specifications IAB inadvertent actuation block ISTS improved standard te chnical specifications LCO limiting condition of operation LTOP low temperature overpressure protection LWR light water reactor MCR main control room PLI pressurized line isolation PWR pressurized water reactor RCS reactor coolant system RRV reactor recirculation valve RTP rated thermal power RVV reactor vent valve SDA Standard Design Approval SR surveillance requirement STS standard technical specifications TS technical specifications T-traveler technical traveler UHS ultimate heat sink

Table 1-2 Definitions Term Definition Decay heat removal Decay heat removal system actuation means actuation of the DHRS and system (DHRS) actuation includes isolation of the steam and fe edwater flow paths outside of the decay heat removal interfaces with the steam generators in accordance with the descriptions provided in t he US460 SDA application. This is accomplished by a combination of th e module protection system DHRS actuation signal and the second ary system isolation signal.

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Table 1-2 Definitions (Continued)

Term Definition Emergency core cooling Emergency core cooling system actuation describes the signal that system (ECCS) actuation permits the ECCS valves (reactor vent valves (RVVs) and reactor recirculation valves (RRVs)) to op en. The RVVs open immediately upon receipt of an actuation signal. The RRVs may not immediately open in response to actuation depending on th e function of the pressure interlock feature that compares reactor coolant pressure with the pressure in the containment in accordance with the descriptions provided in the US460 SDA.

keff effective neutron mult iplication factor, keff = 1 is the critical configuration

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2.0 Background

2.1 Approach

The determinations required to define the cont ent of the TS are primarily based on the requirements of 10 CFR 50.36 (Reference 5.1.4), 10 CFR 50.36a (Reference 5.1.5), and the discussion in the associated NRC policy (Reference 5.1.6) as described in TR-1116-52011, Technical Specifications Regulatory Conformance and Development (Reference 5.1.3).

Chapters 1, 2, 4, and 5 of the NuScale US600 DCA technical specifications (Reference 5.1.1) and the NuScale US460 SDA TS are generally aligned with the corresponding sections of the legacy plant GTS. This has the advantage of generally aligning the NuScale TS with the NRC requir ements and expectations in these areas, and addressing the requirements of 10 CFR 50.36a. It also assists future internal and external communications and interpretations by generally conforming with the expectations and knowledge experience base of plant, industry, and regulatory staff.

Chapter 3 of the NuScale US600 DCA technical specifications presented a significant set of issues related to application of the crit eria for inclusion. To perform the review and identify appropriate limiting condition of operation (LCO) contents, a TS structure that generally parallels the contents in NUREG-1431 and the other pressurized water reactor (PWR) designs is adopted for the proposed NuScale TS, albeit with some significant changes as described in TR-1116-52011 (Reference 5.1.3).

The US460 SDA technical specifications utilize the organization and groupings of LCO requirements adopted in the NuScale US600 DC A TS that have shown to provide clear information to the operating staff. This organization also permits some level of comparison of the NuScale US460 SDA technical specifications to existing large light water reactor (LWR) standard technical specifications when appropriate.

Inclusion of individual Chapter 3 specifications in the NuScale GTS is based on application of the four criteria in 10 CFR 50.36(c)(2)(ii):

1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

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Section 3.0 of TR-1116-52011 (Reference 5.1.3) describes the assessment of each TS chapter and its incorporation into the NuScale GTS. Section 3.0 of this report describes substantive changes to the content of the TS from the DCA generic technical specifications to the US460 SDA technical specifications, including criterion for inclusion where relevant.

2.2 Regulatory Requirements

Regulation 10 CFR 50.36 describes requirement for and the content to be included in the TS.

Regulation 10 CFR 50.36a requires applicants for a design certification to include technical specifications that address applicable provisions of 10 CFR 20.1301, and procedures related to the control of effluents and radioactive waste systems.

Subpart E of 10 CFR 52, Standard Design Approvals, does not require submittal of Technical Specifications for consideration. However, NuScale determined the TS and Bases support the US460 SDA application by providing a basis for the evaluation of how the design and its analyses will be implemented consistent with their descriptions in the US460 Final Safety Analysis Report (FSAR) and elsewhere. Additionally, the US460 SDA technical specifications provide a basis for development of generic standard TS for the NuScale design.

2.3 Design Specific Review Standard

A design specific review standard (DSRS) was issued for the NuScale small module reactor design (ML15355A312), primarily fo r use in evaluating the US600 design application and any subsequent combined license applications. No specific mention of use of the DSRS during review of an SDA was included, however the guidance was generally applicable to, and used in, the development of the US460 SDA technical specifications.

The DSRS provided a review path that is based on the evolution of existing operating PWR into the NuScale TS. The preparation, review, and subsequent approval of the DCA technical specifications was based on that evolution as described in TR-1116-52011 (Reference 5.1.3). However with the approv ed US600 technical specifications baseline available, the US460 SDA development and t he description in this report focuses on the changes needed from the US600 DCA technical specifications content to the US460 SDA-specific content. The majority of the DSRS content remains applicable with this approach.

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3.0 Changes to the Content of Standard NuScale Technical Specifications

This discussion describes the changes from the NuScale DCA technical specifications that are incorporated into the NuScale US460 SDA technical specifications. The reason for the changes is described in general terms, and includes removals, relocations, and new requirements. Details of design changes and safety analyses are described in the relevant and referenced US460 FSAR sections.

3.1 Modifications to Chapter 1, Use and Application

3.1.1 1.1 Definitions

LEAKAGE

The definition of Pressure Boundary LEAKAGE is modified consistent with the applicable portions of the LEAKAGE definition provided in NUREG-1431, Revision 5 as appropriate for the NuScale design. The change includes modifications to the punctuation used in this definition, consistent with the Writers Guide for Plant-Specific Improved Technical Specifications, TSTF-GG-05-01, Reference 5.1.7.

This change also affects LCO 3.4.5, reac tor coolant system (RCS) Operational LEAKAGE and associated Bases and additional details regarding this change are provided in the description of changes to LCO 3.4.5.

MODE

The MODE definition used in the TS is c hanged to better align with the plant response behavior. Specifically, the upper temperature limit on MODE 3, Safe Shutdown, is removed and the operational region expanded to include temperatures above the minimum temperature for criticality. This is accomplished by including and and or requirements so that above the minimum temperature for criticality the plant is in MODE 3 if it is PASSIVELY COOLED, and in MODE 2 if it is not PASSIVELY COOLED. Below the minimum temperature for criticality, the plant is in MODE 3.

Figure 3-1 illustrates this.

This MODE definition clarifies that the plant is in a passively safe configuration once PASSIVE COOLING is established, regardless of the reactor coolant temperature relative to the minimum temperature for criticality.

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Figure 3-1 MODES 2 and 3 Illustration

MODE 2 MODE 3 Hot Shutdown Safe Shutdown

Not PASSIVELY PASSIVELY COOLED COOLED

Minimum Temperature for Criticality

MODE 3 Safe Shutdown

PASSIVELY COOLED - PASSIVE COOLING

This definition is revised to reflect the new design of the ECCS that only requires one or more RVVs and one or more RRVs to be open to perform its safety function. A related change to the OPERABILITY requirements for the system is provided in LCO 3.5.1, Emergency Core Cooling System.

RATED THERMAL POWER

The rated thermal power (RTP) for the US460 design is increased from 160 MWt to 250 MWt, consistent with the plant design and safety analyses.

SHUTDOWN MARGIN

The reference temperature used to establish the shutdown margin is reduced from 420 degrees F to 345 degrees F to maintain alignment with LCO 3.4.2, RCS Minimum Temperature for Criticality and used in the safety analyses.

3.1.2 Sections 1.2 through 1.4

No changes to these sections from that provided as NuScale US600 DCA content.

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3.2 Modifications to Chapter 2, Safety Limits

The reactor core critical heat flux correlations and limits, and the RCS pressure safety limits are revised to reflect the increased reactor power and changes to the plant design as described in the FSAR. Surveillance Requirement 3.4.4.1 also modified to reflect new limits.

3.3 Changes to Chapter 3, Limiting Conditions for Operation and Surveillance Requirements 3.3.1 Modification of Limiting Condition of Operation 3.0.3

The legacy nuclear plant owners have proposed changes to the time provided to initiate a shutdown when LCO 3.0.3 applies. The changes are described in a proposed NRC/industry traveler that is applicable to legacy plant STS. NuScale monitored these efforts in public meetin gs and believes that a corresponding change is appropriate for incorporation into the NuScale specifications.

Similarly, the Bases for LCO 3.0.3 are being revised to align to the appropriate extent with the proposed change to the legacy plant STS.

3.3.2 Addition of Surveillance Requirement 3.1.9.5 - Verification of Isolation of Module Heatup System from Other Modules

This surveillance requirement is added to verify inter-module alignment exists to prevent interactions that could affect boration of the RCS.

3.3.3 Modification of Limiting Condition of Operation 3.2.1, Enthalpy Rise Hot Channel Factor (FH), and 3.2.2, Axial Offset

This change expanded Applicability of LCO 3.2.1 and 3.2.2 to require the LCOs limits be met at or above 20 percent RTP, rather than 25 percent RTP, expanding the applicability of the requirements to a larger region of the operating power levels.

3.3.4 Changes to Limiting Condition of Operation 3.3.1, Module Protection System

Modifications are made to actuation logic to align with safety analyses and design changes consistent with the increased reactor rated thermal power, safety analyses, refinements in operational intentions, and lessons learned since the submittal of the DCA technical specifications. Changes include editorial renumbering and arrangement of functions.

Modification of Emergency Core Cooling System Actuation Signals

The design of the ECCS and the associated actuation signals have been changed.

These changes to the ECCS components are described in FSAR Chapter 7, Section 6.3, and further addressed in LCOs 3.4.10 and 3.5.1. The analysis of ECCS actuation, including consideration of boron mixing features and the ECCS

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supplementary boron (ESB) system required by the new LCO 3.5.4, resulted in use of reactor pressure vessel riser level signals to initiate ECCS response. The revised actuation logic uses two level setpoints - Low, and Low-Low, to ensure appropriate response depending on RCS conditions as indicated by the RCS T cold signal (T-5).

The new design and actuation logic are consistent with the safety analyses in the FSAR. The new actuation signals are include d because they satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Addition of Pressurizer Line Isolation Actuation

This change adds a Pressurizer Line Isolation actuation that isolates the pressurizer spray and high point vent line containment isolation valves in response to a low pressurizer level signal. This isolation provides improved plant response to postulated small line breaks occurring on the pressurizer spray and high point vent lines. The new actuation signal is included because it satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Addition of Reactor Trip on High Reactor Coolant System Average Temperature

This change adds a High Tavg actuation at lower powers. The actuation provides earlier response to conditions similar to the high RCS temperature trip when the event begins at a temperature below the full power T hot temperature. The new trip is included because it satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Addition of Emergency Core Cooling System Actuation Timer on Reactor Trip

A delayed actuation is added to ensure ECCS actuation and subsequent ECCS supplemental boron (ESB) function after reactor trip for non-loss-of-coolant accident events. Requirements for, and a description of the ESB function, is provided in new LCO 3.5.4 and the associated Bases. The new actuation signal is included because it satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Bypass Low-Low Pressurizer Pressure Reactor Trip and DWSI when Reactor Coolant System Temperature Below T-3

Reactor trip and demineralized water system isolation actuation is modified so that the signal is not active when RCS temperature is below the T-3 interlock temperature.

The RCS temperature T-3 bypass is active when RCS temperature is less than approximately 340 degrees F. This change supports startup of the reactor, while maintaining the credited safety function of the trip and actuation.

Other Table 3.3.1-1, Module Protection System Instrumentation Changes

Footnotes referring to capability to withdraw a control rod assembly (CRA), are modified to indicate the requirement applies when capable of withdrawing more than a single CRA. This change is necessary to allow energization of a portion of the control rod drive system (CRDS) in MODE 3 when preparing for module disassembly.

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This capability is required to verify the CRA is disconnected from its extension rod.

The CRDS design and administrative control ensure that no more than one CRA may be manipulated, and the definition of, and li mits on SHUTDOWN MARGIN specified in LCO 3.1.1 ensure the plant remains safely shutdown. A description of the functional design of the CRDS is provided in FSAR Section 4.6. This change results in changes to the Bases discussions for the associated Functions.

Footnotes limiting the Applicability of t he Demineralized Water System Isolation are added to only require OPERABILITY when RCS temperature is above the T-3 interlock. The RCS T-3 bypass is active when RCS temperature is less than approximately 340 degrees F. This change also resulted in addition of a footnote to Table 3.3.3-1, Engineered Safety Features Actuation System Logic and Actuation Functions.

Some other footnotes required modification because of the combination of allowances described above.

3.3.5 Addition of Limiting Condition of Operation 3.3.3 Condition for Pressurizer Line Isolation Inoperability

The description of Condition F in LCO 3.3.3 is modified to address circumstances when both divisions of the pressurizer line isolation (PLI) function are inoperable. The PLI signal closes a subset of the chemical and volume control system (CVCS) isolation valves so the revised Condition and existing Required Actions ensure the safety function is met. The revised Required Action requires closure of affected valves. If the PLI actuation logic is inoperable then the pressurizer spray and high point vent lines must be isolated. If the CVCS actuation logic is inoperable then all four CVCS lines must be isolated.

3.3.6 Addition of Surveillance Requirement 3.3.3.3, Monitoring Emergency Core Cooling System Actuation Time Delay

The delayed ECCS actuation signal added to LCO 3.3.1 is implemented by time delays established in the module protection system logic. Surveillance Requirement 3.3.3.3 is added to ensure the time delays is within limits. The time delay limits are core cycle-specific depending on fuel makeup so the time delay limits are specified in the core operating limits report (COLR). The baseline delay is approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The surveillance frequency is in accordance with the Surveillance Frequency Control Program, with an initial interval of 24 months.

3.3.7 Removal of Limiting Condition of Operation 3.3.5, Remote Shutdown Station

The DCA technical specifications LCO 3.3.5 is removed from the US460 SDA technical specifications. The Design Certification Application LCO was included in accordance with Criterion 4 of 10 CFR 50. 36(c)(2)(ii); however further consideration during the development of the US460 SDA design resulted in concluding that it is no longer appropriate for inclusion. This conclusion is based on the system design and details provided in US460 FSAR Chapters 7 and 18.

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Chapter 7 of the FSAR describes the capability of the plant design to respond in the event of a fire in the main control room (MCR). As described there, in the event of a fire in the MCR the operators trip the reactors, initiate decay heat removal and initiate containment isolation before evacuating the MCR. These actions result in passive cooling that achieves and maintains the modules in a safe shutdown condition.

Operators can also place the reactors in safe shutdown from outside the MCR in the module protection system equipment rooms within the Reactor Building.

The operators then use alternate operator workstations to monitor plant conditions.

Following shutdown and initiation of passive cooling, the design does not rely on operator action, instrumentation, or controls outside of the MCR to maintain a safe stable shutdown condition.

3.3.8 Modification of Limiting Condition of Operation 3.4.2, Minimum Temperature for Criticality

The minimum temperature for criticality is reduced to 345 degrees F to improve the ability of the reactor to startup in a timely manner after an outage by using nuclear heat to increase temperatures to the normal operating range. Safety analyses include consideration of this new limit.

3.3.9 Modification of Surveillance Requirement 3.4.4.1 - Reactor Safety Valve Setpoints

Reactor safety valve setpoints are changed to reflect increase reactor power, reactor vessel design pressure, and associated safety analyses.

3.3.10 Editorial Clarification of Condition D of Limiting Condition of Operation 3.4.3, Reactor Coolant System Pressure / Temperature Limits

Condition D is clarified to specifically addr ess initiation of containment flooding and to modify the Required Action to immediately initiate action to be in MODE 2. This change more accurately reflects plant operati ons and more closely aligns with similar Required Actions used in similar circumstances requiring immediate actions be taken.

3.3.11 Modification of Limiting Condition of Operation 3.4.5, Reactor Coolant System Operational Leakage

The definition of LEAKAGE and LCO 3.4.5 are revised based on changes to large legacy PWR standard technical specifications issued as Rev 5; however modified to reflect the design and operation of a NuScale plant. The change clarifies the requirements for pressure boundary leakage such as could be postulated to exist on RCS piping outside the containment before the outermost containment isolation valves. The FSAR Section 5.2, Integrity of Reactor Coolant Boundary describes these lines up to the outermost containment isolation valves as part of the reactor coolant pressure boundary.

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The change adds a new Condition requiring action to isolate the affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and deactivated automatic valve, blind flange, or check valve within four hours. Subsequent Conditions and Required Actions are renumbered.

The Bases for LCO 3.4.5 are also revised based on changes to large legacy PWR standard technical specifications with changes to reflect the NuScale design and operations.

Maintaining alignment with large LWR technical specifications to the extent appropriate for the design promotes understanding and interpretation of TS requirements during internal and external communications between plant and regulatory staff.

3.3.12 Modification of Limiting Condition of Operation 3.4.8, Reactor Coolant System Specific Activity

Limits on I-131 and Xe-133 are modified to re flect safety analyses for the new reactor design, including the increased reactor power level compared to the US600 design.

The US460 FSAR Section 11.1.3 describes the realistic source term used to develop the source terms used as described in F SAR Section 15.0.3 to describe consequence analyses of design basis events, including the concentration limits in US460 Standard Design Approval LCO 3.4.8. The FSAR Table 11.1-2 describes parameters used to calculate coolant source terms, including the reactor core thermal power.

3.3.13 Modification of Limiting Condition of Operation 3.4.10, Low Temperature Overpressure Protection Valves

This LCO is modified to reflect the revised ECCS design that uses two RVVs. One RVV is adequate to provide low temperature overpressure protection as described in the FSAR.

3.3.14 Modification of Limiting Condition of Operation 3.5.1, Emergency Core Cooling System

This LCO is modified to reflect the revised ECCS design that only uses two RVVs, and removes the inadvertent actuation bl ock function from the RVVs. Surveillance Requirement 3.5.1.3 is also modified to reflect the removal of the inadvertent actuation block function from the RVVs.

3.3.15 Modification of Limiting Condition of Operation 3.5.3, Ultimate Heat Sink

The ultimate heat sink (UHS) is redesigned consistent with the other changes to the plant design and analyses, primarily the increased RTP and a reduction in the number of reactors in the design to a maximum of six modules. The redesign resulted in reanalysis and redefinition of the UHS and caused changes in the credited functions of the UHS in LCO 3.5.3.

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The UHS water level requirements are specified to a new band defined by upper and lower limits that improve containment heat removal behavior. The redesigned UHS and its functions are described in FSAR Section 9.2.5. The new limits are consistent with the safety analyses in the FSAR that credit the UHS function. Similarly, the maximum bulk average pool temperature is increased to align with the safety analyses assumptions. The structure of the Actions in LCO 3.5.3 are changed to reflect the removal of distinct limits that the DCA credited for separate safety functions. This change removed the need for Condition B of the DCA technical specifications, which is now addressed by Condition A. Completion Times remain consistent with the credited functions of the UHS. Subsequent Conditions are renumbered. Corresponding changes are made to the Bases.

3.3.16 Addition of Limiting Condition of Operation 3.5.4, Emergency Core Cooling System Supplemental Boron

The US460 design adds a passive system that provides soluble boron in dissolvers mounted inside the containment. The dissolvers provide a reservoir of boron that mixes with condensate from the upper inner surfaces of the containment vessel when the ECCS is actuated. Limiting Condition of Operation 3.5.4 is added to ensure that the quantity of boron available for dissolution when the ECCS actuates conforms to the assumptions in the safety analyses. The boron ensures the reactor remains subcritical after certain events in combinat ion with limiting conditions, and subsequent cooldown of the reactor system. The quantity of boron required is specified in the COLR. The ESB satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

3.3.17 Addition of Limiting Condition of Operation 3.6.3, Containment Closure

Limiting Condition of Operation 3.6.3 is added to ensure that module inventory is preserved during movement of the module between the operating location and the containment closure tool. The LCO requires a module that is in MODE 4, with the upper module assembly seated on the lower containment vessel flange, be maintained closed. The LCO and allowances are patterned on portions of NUREG-1431, LCO 3.9.4 with extensive modifications to align with the NuScale application. Maintaining containment closure ensures that the decay heat removal mechanism required to assure core cool ing is maintained during periods when the module is isolated from other systems such as CVCS, or when the containment is disassembled from the UHS via the de-ener gized ECCS valves. Limiting Condition of Operation 3.6.3 satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

3.3.18 Removal of Limiting Condition of Operation 3.7.3, In-Containment Secondary Piping Leakage

Limiting Condition of Operation 3.7.3 is deleted as no longer necessary because the break exclusion design criteria is applied to the secondary system piping within the containment. The DCA design for secondary system piping met the leak-before-break design criteria of General Design Criteria 4.

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US460 Standard Design Approval FSAR Section 3.6 describes the application of design measures to prevent or mitigate postulated dynamic effects associated with postulated rupture of US460 piping. The US460 SDA design of secondary piping inside the containment meets the criteria for exclusion from postulated breaks and cracks provided in NRC Branch Technical Position (BTP) 3-4. Based on this change the US600 Design Certification Application LCO is no longer needed because the piping is excluded from consideration of postulated breaks and cracks.

3.3.19 Other Bases Changes

In addition to the specific changes described above, Applicable Safety Analyses sections are modified to reflect changes to th e safety analyses, primarily as a result of the increased reactor power. Other changes are made in response to operational analysis feedback to clarify and ease understanding of the requirements.

3.4 Chapter 4, Design Features

Section 4.3 Fuel Storage

The fuel storage design description is modified to reflect changes to the design and analyses. Key variables are bracketed to al low replacement with actual plant-specific values when design details are finalized by a future applicant that references the NuScale power plant US460 standard design. NuScale is monitoring industry efforts to relocate fuel storage detailed requirements to a COLR -like document and anticipates adopting this practice when the concept matures.

3.5 Chapter 5, Administrative Controls

Section 5.2.2, Facility Staff

This section is modified to reflect approved topical report TR-0420-69456, NuScale Control Room Staffing Plan, TR-0420-69456, Revision 1-A.

Section 5.5.9, Containment Leakage Rate Testing Program

The description of the Containment Leakage Ra te Testing Program is revised to provide alternatives to adopt Option A or Option B of 10 CFR 50, Appendix J.

Section 5.6.3, Core Operating Limits Report

This section is modified to align with the safety analyses, referencing technical specification limits, and topical reports that describe the limits that will be included in the COLR.

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Section 5.6.4, Reactor Coolant System PRESSURE AND TEMPERATURE LIMITS REPORT

Modified to align with the safety analyses and topical report that describes the limits that will be included in the Pressure and Temperature Limits Report.

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4.0 Conformance with Industry Standard Technical Specifications and Standard Technical Specification Writers Guide

The US460 Standard design, analyses, and therefore the technical specifications are significantly different from those provided in the STS for large LWRs. However, to the extent appropriate, conventions and content have been adopted and incorporated to parallel the NUREG standard technical specificat ions. For example, the writers guide is generally used with regard to format and content of individual specifications and bases discussions.

For example, Chapter 1, Use and Applicat ion, and Sections 3.0, LCO and Surveillance Requirement Applicability contain content generally consistent with legacy industry content. This approach supports internal and external understanding and application of the balance of the contents of the technical specifications and bases.

Industry travelers that were publicly av ailable on the NRC's Agencywide Documents Access and Management System since issuance of the DCA technical specifications and NUREG standard technical specifications revisions, have been monitored for potential applicability to the NuScale TS and bases. Where conceptually appropriate NuScale-specific content is developed and incorporated. A summary of actions taken to incorporate changes similar to those publicly av ailable industry travelers is provided in the table below.

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Table 4-1 Standard Technical Specifications Traveler Adaptation Traveler Addressed Comments No.

571 N/A T-traveler or otherwise not available 572 N/A T-traveler or otherwise not available 573 No Boiling water reactor (BWR)-specific 574 N/A T-traveler or otherwise not available 575 N/A T-traveler or otherwise not available 576 No BWR-specific 577 Yes Addressed in Section 5.5 to extent appropriate 578 No Not applicable to NuScale design 579 No Not applicable to NuScale TS 580 No Not applicable to NuScale design 581 N/A T-traveler or otherwise not available 582 No Not applicable to NuScale design 583 N/A T-traveler or otherwise not available 584 No BWR-specific NuScale incorporated revision 0 like cont ent in anticipation of industry and 585 Yesregulatory adoption of this cha nge. Adjustments will be considered for incorporation as industry and regulatory issues are resolved.

586 N/A T-traveler or otherwise not available 587 N/A T-traveler or otherwise not available 588 Pending NuScale is monitoring this proposal for future consideration and adoption.

589 No Not applicable to NuScale design 590 N/A T-traveler or otherwise not available 591 No Not applicable to NuScale TS 592 No Not applicable to NuScale design 593 N/A T-traveler or otherwise not available 594 N/A T-traveler or otherwise not available 595 N/A T-traveler or otherwise not available 596 Pending NuScale is monitoring this proposal for future consideration and adoption.

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5.0 References

5.1 Referenced Documents

5.1.1 NuScale Power, LLC, Generic Technical Specifications, NuScale Nuclear Power Plants, DCA Part 4, Volume 1: Specifications, Revision 5.

5.1.2 NuScale Power, LLC, Generic Technical Specifications, NuScale Nuclear Power Plants, DCA Part 4, Volume 2: Bases, Revision 5.

5.1.3 NuScale Power, LLC, Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4, May 2020.

5.1.4 U.S. Code of Federal Regulations, Technical Specifications, Section 50.36, Part 50, Chapter I, Title 10, Energy, (10 CFR 50.36).

5.1.5 U.S. Code of Federal Regulations, Technical Specifications on Effluents from Nuclear Power Reactors, Section 50.36a, Part 50, Chapter I, Title 10, Energy, (10 CFR 50.36a).

5.1.6 U.S. Nuclear Regulatory Commission, Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, Federal Register, Vol. 58 FR 39132, July 22, 1993.

5.1.7 Technical Specification Task Force, Writers Guide for Plant-Specific Improved Technical Specifications, TSTF-GG-05-01, Revision 1, Rockville, MD, August 2010.

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2.0 Background

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