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LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4
	ML20141L804
Person / Time
Site:	NuScale
Issue date:	05/20/2020
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LO-0520-70089
	Download: ML20141L804 (74)
	v • d • e

LO-0520-70089 May 20, 2020 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4

REFERENCE:

Letter from NuScale Power to NRC, NuScale Power, LLC Submittal of Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 3, dated November 15, 2019 (ML19319C787)

NuScale Power, LLC (NuScale) hereby submits Revision 4 of Technical Specifications Regulatory Conformance and Development (TR-1116-52011).

The enclosure to this letter contains the nonproprietary report entitled Technical Specifications Regulatory Conformance and Development.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions, please contact Rebecca Norris at 541-602-1260 or at RNorris@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC Prosanta Chowdhury, NRC Michael Dudek, NRC Michael Snodderly, NRC Christiana Liu, NRC Christopher Brown, NRC Marieliz Johnson, NRC Getachew Tesfaye, NRC

Enclosure:

Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0520-70089

Enclosure:

Technical Specifications Regulatory Conformance and Development, TR-1116-52011, Revision 4 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Licensing Technical Report Technical Specifications Regulatory Conformance and Development May 2020 Revision 4 Docket: 52-048 NuScale Nonproprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Licensing Technical Report 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 copies 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 license, 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.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Licensing Technical Report 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 necessarily state or reflect those of the United States Government or any agency thereof.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Licensing Technical Report 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 ............................................................................................................... 6 2.2 Regulatory Requirements ...................................................................................... 7 3.0 Content of NuScale Generic Technical Specifications ................................................ 8 3.1 Chapter 1, Use and Application ........................................................................... 10 3.2 Chapter 2, Safety Limits ...................................................................................... 13 3.3 Chapter 3, Limiting Conditions for Operation and Surveillance Requirements ...................................................................................................... 14 3.4 Chapter 4, Design Features ................................................................................ 22 3.5 Chapter 5, Administrative Controls ...................................................................... 22 3.6 Mapping of 10 CFR 50.36 Selected Limits to Proposed Technical Specifications ...................................................................................................... 24 4.0 Comparison with Standard Technical Specifications ................................................ 25 4.1 Specification Comparisons .................................................................................. 25 4.2 Industry/NRC STS Traveler Consideration .......................................................... 26 5.0 Conformance with Standard Technical Specification Writers Guide ...................... 27 6.0 References ..................................................................................................................... 28 6.1 Source Documents .............................................................................................. 28 6.2 Referenced Documents ....................................................................................... 28 7.0 Appendices .................................................................................................................... 30 Appendix A Criteria for Inclusion of Technical Specifications .......................................... 31 Appendix B Summary Comparison of Standard Technical Specifications with NuScale Generic Technical Specifications Contents ..................................... 45 Appendix C Industry / NRC STS Traveler Consideration.................................................... 59 NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Licensing Technical Report TABLES Table 1-1 Acronyms .............................................................................................................. 3 Table 1-2 Definitions .............................................................................................................. 4 Table 3-1 Comparison of standard technical specifications and the proposed NuScale generic technical specifications ............................................................................. 9 Table 3-2 NuScale technical specification MODES ............................................................. 11 Table 3-3 Parameters and operating restrictions that require an LCO ................................ 16 Table 3-4 Core design limits that are initial conditions used in design basis event evaluation specified in the Core Operating Limits Report ................................... 17 Table 3-5 Module protection system signals used in the Chapter 15 analyses ................... 18 Table 3-6 NuScale structures, systems, and components credited to actuate or function in design basis accident and transient analyses ................................................. 20 Table 3-7 Comparison of NuScale generic technical specifications with NUREG-1431 Section 5.5 contents ............................................................................................ 23 Table A-1 Technical specifications inclusion criteria ............................................................ 32 Table B-1 Comparison of standard technical specifications with NuScale generic technical specifications ........................................................................................ 46 Table C-1 Industry / NRC STS traveler evaluation............................................................... 60 FIGURES Figure 3-1 MODES Temperature and Reactivity Conditions................................................. 12 NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Abstract This report describes the development process of the NuScale Power Plant technical specifications (TS) to conform with regulatory requirements and expectations regarding scope, content, and format. This report also provides the basis for including the requirements chosen for the NuScale TS.

Revisions 1, 2, and 3 were previously prepared to clarify the development process and changes made to the technical specifications resulting from NRC requests for additional information.

Additional changes are reflected that occurred as the result of changes to the FSAR and plant design that are now reflected in the technical specifications.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 1.0 Introduction 1.1 Purpose The purpose of this report is to describe 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. Revision 2 of this report includes consideration of changes made after revision 1 of this report and reflects changes in contents of the technical specifications through July 15, 2019. This report also provides the basis for including the specifications chosen for the NuScale TS. The report focuses on three aspects:

Conformance with 10 CFR 50.36 (Reference 6.2.1) and 10 CFR 50.36a (Reference 6.2.2), including considerations in 58 FR 39132, Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Reference 6.2.11).

Conformance with regulatory expectations as expressed by industry standards and precedent as defined in NRC-published standard technical specifications (STS),

approved generic technical specifications (GTS), and subsequent changes as delineated in industry / NRC STS travelers.

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 6.2.4).

The report also describes additional changes to the technical specifications and bases that resulted from

NRC requests for additional information (RAI),

Industry/NRC traveler status changes since December 2016, and

design or regulatory changes implemented in the FSAR that result in changes that need to be reflected in the technical specifications.

1.2 Scope This report addresses the development of the GTS applicable to an individual NuScale module. The NuScale GTS are drafted in the context of the design certification application for a 12-module NuScale facility, however the content is applicable to an individual module.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 1.3 Abbreviations Table 1-1 Acronyms Term Definition AP1000 Westinghouse AP1000 (as described in 10 CFR 52, Appendix D)

B&W Babcock and Wilcox BWR boiling water reactor CE Combustion Engineering CFDS containment flooding and drain system CHF critical heat flux CNV containment vessel COL combined license COLR core operating limits report CRA control rod assembly CVCS chemical and volume control system DCA Design Certification Application DHRS decay heat removal system ECCS emergency core cooling system ESBWR GE Hitachi Economic Simplified Boiling Water Reactor (as described in 10 CFR 52, Appendix E)

ESF engineered safety feature ESFAS engineered safety features actuation system FSAR Final Safety Analysis Report GDC General Design Criterion GE General Electric GTS generic technical specifications HELB high-energy line break HFP hot full power HZP hot zero power ISTS Improved Standard Technical Specifications LCO limiting condition of operation LOCA loss-of-coolant accident LTOP low temperature overpressure protection MPS module protection system MSIV main steam isolation valve (typically includes associated drain and bypass valves)

NPM NuScale Power Module ODCM Offsite Dose Calculation Manual PTLR Pressure Temperature Limits Report PWR pressurized water reactor NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 RAI request for additional information (from NRC)

RCPB reactor coolant pressure boundary RCS reactor coolant system RTP rated thermal power RTS reactor trip system SDM shutdown margin SG steam generator SL safety limit SP Setpoint Program SR surveillance requirement SSC structures, systems, and components SSI Secondary System Isolation STS standard technical specifications TS technical specifications TSTF Technical Specification Task Force UHS ultimate heat sink W Westinghouse Table 1-2 Definitions Term Definition Decay heat removal DHRS Actuation means actuation of the DHRS and includes system (DHRS) actuation isolation of the steam and feedwater flow paths outside of the decay heat removal interfaces with the steam generators (SGs) in accordance with the descriptions provided in the Design Certification Application (DCA). This is accomplished by a combination of the module protection system DHRS actuation signal and the secondary system isolation (SSI) signal.

Emergency core cooling ECCS actuation describes the signal which permits the ECCS system (ECCS) actuation valves (reactor vent valves and reactor recirculation valves) to open. The valves may not immediately open in response to actuation depending on the 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 DCA.

keff effective neutron multiplication factor, keff=1 is a critical configuration NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 2.0 Background Historically, the NRC approved TS evolved from plant-specific custom TS, which evolved into the old standard TS, and more recently to the existing improved standard TS.

The current standard technical specifications (STS) are published by the NRC as NUREGs 1430 - 1434 to address Combustion Engineering (CE), Westinghouse (W),

Babcock and Wilcox (B&W), and General Electric (GE) reactor designs (References 6.2.5 through 6.2.9). Additionally, the NRC issued NUREG-2194 as the STS for the AP1000 certified design (Reference 6.2.10). The format and content of the TS has evolved as facility operations and designs were refined and operating experience was gained.

The NRC initially developed the STS based on criteria in an interim NRC policy published in 1987 (Reference 6.2.3) that was later refined and published as 58 FR 39132, Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, on July 22, 1993 (Reference 6.2.11). The policy resulted in changes to ensure availability of safety systems and functions assumed or that mitigate design basis accidents, while minimizing technical specification content that reduced the focus of the plant operating staff on safety.

The transition to the published STS was preceded by the development and issuance of a report (commonly referred to as the Split Report) by the NRC on May 9, 1988 (Reference 6.2.12) that applied the NRC policy and defined the appropriate content for TS. The Split Report considered nuclear steam supply system vendor and owners group submittals that applied the interim policy inclusion criteria to the vendor-specific old standard TS. The Split Report explicitly identified TS content that should be included in the new STS and identified TS content that could be relocated to other licensee-controlled documents.

Being based on previously existing, old STS, the contents of the new STS that were retained evolved from the content and consideration of the historical TS.

The NuScale Power Plant is a unique integral pressurized water reactor design for which existing STS are not applicable and for which representative TS have not been previously issued. The plant design, safety functions, structures, systems, and components (SSC),

and behavior are significantly different from those in previously licensed designs and facilities. Integrated and simplified behavior reduces the scope of safety systems and combines safety functions into a smaller set of SSC whose operability are assumed as initial conditions or credited to respond in the safety analyses.

Because of this, NuScale does not have a historical basis of existing TS, nor design-specific operating experience to inform the content of the TS. Rather, the NuScale TS are developed directly from the design and planned operations, and are informed by industry operating experience and TS content for which similar or parallel functions and features exist in other designs. These factors are then compared with the criteria for inclusion as TS.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 This report describes the consideration of the NuScale design and operations, and applies the TS inclusion criteria in 10 CFR 50.36 and 10 CFR 50.36a, consistent with

the final NRC policy,

considerations detailed in the Split Report,

the Writer's Guide for Plant-Specific Improved Technical Specifications,

the content of the current versions of the STS, and

the refinements to the STS developed by the industry / NRC travelers through June 30, 2019 to the extent they are available and applicable.

This report also describes the changes made to the technical specifications as a result of NuScale responses to RAIs. In some cases, NuScale included additional content that does not strictly align with the above requirements or industry precedent regarding technical specification content.

2.1 Approach The determinations required to define the content of the TS are primarily based on the requirements of 10 CFR 50.36, 10 CFR 50.36a, and the discussion in the associated NRC policy and statements of consideration.

Chapters 1, 2, 4, and 5 of the STS generally provide parallels that are applied to define corresponding NuScale GTS content. This has the advantage of generally aligning the NuScale GTS with the NRC requirements and expectations in these areas, and addressing the requirements of 10 CFR 50.36a.

Chapter 3 of the TS presented a significant set of issues related to application of the criteria 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 was adopted for the proposed NuScale TS, albeit with some significant changes. The existing organization and groupings of requirements in the STS have been refined since their initial issuance, and have been demonstrated to provide clear information to the operating staff. This organization also permits some level of comparison of the NuScale TS to existing STS 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.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4

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.

Section 3 of this report describes the assessment of each TS chapter and its incorporation into the NuScale GTS.

2.2 Regulatory Requirements 10 CFR 52.47 Paragraph (11) requires applicants for standard design certifications to provide proposed TS prepared in accordance with the requirements of 10 CFR 50.36 and 10 CFR 50.36a.

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

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.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 3.0 Content of NuScale Generic Technical Specifications The NuScale design includes up to 12 NuScale Power Modules (NPMs) in a single large reactor pool as described in the Final Safety Analysis Report (FSAR). Each NPM is an independently operated, nominal 160 MWt pressurized water nuclear reactor that supplies steam to an associated turbine generator and condenser to produce electrical power. The reactor pool provides an essential shared function as the ultimate heat sink (UHS) for the passive core cooling systems of each of the 12 individual NPMs.

Significantly, the plant and NPM design precludes the need for electrical power to safely shut down or remove residual heat from the NPMs if a design basis loss of power occurs.

The 12 NPMs share a single main control room that incorporates individual digital instrumentation and control systems for each NPM.

A major difference from existing plant designs is that individual NPMs are disconnected from their operating systems, instrumentation, and controls and moved to a separate area of the Reactor Building for refueling operations and maintenance.

More detailed information regarding the design and operation of the NuScale plant is provided in the DCA.

Based on the unique nature of the NuScale design, the proposed NuScale GTS are structured similar to the legacy large light water reactors and the AP1000 STS, but with significant differences. Table 3-1 provides a comparison of the structure of the legacy large light water reactor STS and AP1000 STS with the NuScale GTS at the chapter and section levels.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table 3-1 Comparison of standard technical specifications and the proposed NuScale generic technical specifications Chapter/ Rev. 4 STS NUREG-Section AP1000 STS 1430 1431 1432 1433 and 1434 NUREG-2194 NuScale GTS 1.0 USE AND APPLICATION 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.2 SAFETY LIMIT VIOLATIONS 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 REACTIVITY CONTROL SYSTEMS 3.2 POWER DISTRIBUTION LIMITS 3.3 INSTRUMENTATION 3.4 REACTOR COOLANT SYSTEM (RCS) 3.5 EMERGENCY CORE COOLING ECCS AND PASSIVE CORE COOLING SYSTEMS SYSTEM (ECCS) REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) 3.6 CONTAINMENT SYSTEMS 3.7 PLANT SYSTEMS 3.8 ELECTRICAL POWER SYSTEMS REFUELING OPERATIONS 3.9 REFUELING OPERATIONS Relocated to 3.8 3.10 (not used) SPECIAL (not used)

OPERATIONS 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Procedures 5.5 Programs and Manuals 5.6 Reporting Requirements 5.7 High Radiation Area NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 The NuScale GTS adopted the structure and relevant content from Chapters 1, 2, 4, and 5 in recognition of the commonality of this content. Changes proposed to the Use and Application, and Applicability sections were made to reflect the contents of the NuScale GTS. Where the STS Chapter 1 describes a situation that is not consistent with the NuScale plant design and specifications, the NuScale GTS are modified to reflect the NuScale use or excluded, as appropriate.

NuScale GTS Sections 3.1 through 3.8 were drafted based largely on the NUREG-2194 and NUREG-1431 specifications; however other STS and the Economic Simplified Boiling Water Reactor GTS (Reference 6.2.13) were also used during development. Individual specifications that do not have a parallel SSC or function in the NuScale Power Plant design were omitted. Similarly, NuScale Power Plant SSC or functions that initially screened to meet one or more of the criteria in 10 CFR 50.36(c)(ii), but that are not represented by similar SSC or functions in the NUREGs, were prepared and included. The integrated PWR nature of the NuScale Power Plant, SSC, and control systems resulted in consolidation of some specifications that were formerly defined individually.

The 12-NPM NuScale design and operating paradigm resulted in changes to conform with the STS-like document construction, specifically those related to the relocation of individual NPMs during refueling.

Sections 3.1 through 3.5 of this report outline the NuScale GTS chapter-level structure and discuss the associated content.

3.1 Chapter 1, Use and Application With the exception of the MODE definitions and certain definitions related to the design of individual SSC, the NuScale GTS generally adopt the structure and relevant content from Chapters 1 of the STS. A significant change was the replacement of definitions for response times due to variation from the industry standard meaning and significant differences in the way safety system response times are measured. The digital safety system response time is continuously monitored while in service and evaluated for consideration by conservative calculations as described in TR-1015-18653-P-A, Design of the Highly Integrated Protection System Platform." Sensor response times are measured from the sensor through input to the digital systems. Actuation response times are measured from the output of the digital system until the associated system or component has performed its safety function. Overall function response times are evaluated for acceptability based on the sum of the sensor, digital, and actuation response times.

Changes proposed to the Use and Application sections are limited to those necessary to remain consistent with the balance of the NuScale GTS. For example Section 1.3 refers to a credited pump in the STS, however the NuScale design does not credit any pumps and the example was re-drafted to use a credited valve to illustrate the condition.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 MODES The MODE definitions applicable to PWRs were determined to be inconsistent with NuScale design and operation. Individual NPMs use a comparatively small reactor that depends on natural circulation for flow in the reactor, NPMs are passively cooled in postulated accident conditions, and the design includes relocation of NPMs to perform refueling operations. Therefore, NuScale developed a new MODE structure that more appropriately addresses the NuScale operations paradigm. The NuScale GTS MODES are described in Table 3-2.

Table 3-2 NuScale technical specification MODES INDICATED REACTOR REACTIVITY COOLANT MODE TITLE CONDITION (keff) TEMPERATURES (°F) 1 Operations 0.99 All 420 2 Hot Shutdown < 0.99 Any 420 3 Safe Shutdown(a) < 0.99 All < 420 4 Transition(b)(c) < 0.95 N/A 5 Refueling(d) N/A N/A (a) Any CRA capable of withdrawal, any CVCS or CFDS connection to the module not isolated.

(b) All CRAs incapable of withdrawal, CVCS and CFDS connections to the module isolated, and all reactor vent valves (RVV) de-energized.

(c) All reactor vessel flange bolts fully tensioned.

(d) One or more reactor vessel flange bolts less than fully tensioned.

Descriptions and Rationale for NuScale MODE Structure MODE 1 - Operations This MODE replaces both MODE 1 Power Operation and MODE 2 Startup used in pressurized water reactor STS. This MODE is defined by the reactivity condition (keff) being greater than or equal to 0.99, i.e., approaching criticality or in critical operation. The NuScale design uses an external heat source to raise temperatures above the minimum required for criticality. Once critical, the design initially functions somewhat like a boiling water reactor design, using nuclear heat to increase temperatures during power ascension from initial criticality to full power temperatures at approximately 15 percent of rated thermal power (RTP). The minimum temperature for criticality is included with a requirement that all indicated reactor coolant temperatures be 420 degrees F. The temperature requirements are specified in this manner because dependence on natural circulation makes identification of highest and lowest coolant temperatures difficult during NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 low-flow conditions. This definition ensures consistency with design and accident analyses.

The NuScale reactor operates at comparatively lower power levels and power densities, yet with a fuel design similar to a large PWR. Individual specifications that require a power restriction are explicitly described in their Applicability rather than by using a distinct Startup MODE as found in the PWR STS. Therefore, there is no need for a distinct Startup MODE.

MODE 2 - Hot Shutdown MODE 2 is defined by reactivity condition (keff) less than 0.99 and any indicated reactor coolant temperature at or above 420 degrees F. This MODE corresponds to the transition from safe shutdown toward critical operations, and to conditions immediately after a reactor trip and during intial cooldown. While in this MODE during startup, the NPM has been reconfigured from passive cooling to permit heatup toward operations. The containment vessel (CNV) is drained, and the ECCS and decay heat removal system (DHRS) are aligned to operate if needed.

MODE 3 - Safe Shutdown This MODE is defined by reactivity condition (keff) less than 0.99 and all indicated reactor coolant temperatures less than 420 degrees F. In this MODE, the NPM is in a safe, stable state. Passive decay heat removal is by conduction and convective heat transfer through the CNV walls to the reactor pool. The containment is or will be flooded to a level that ensures passive cooling of the reactor. During unit startup, re-alignment and configuration changes to increase temperatures and enter Hot Shutdown occur in MODE 3.

Graphically, MODES 1, 2, and 3 interact as shown below.

Figure 3-1 MODES Temperature and Reactivity Conditions NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 MODE 4 - Transition The relocation of an NPM from its operating position to the refueling area results in definition of a MODE that addresses this transitional configuration. The module remains partially submerged in the reactor pool throughout movement in this MODE. The MODE governs the period from isolation and disconnection of reactivity controls in the operating position, until the first reactor vessel flange bolt is detensioned in preparation for further disassembly and fuel movement. This MODE includes relocation of the NPM from the operating position in the reactor pool to the containment vessel flange tool and the reactor vessel flange tool for reactor vessel disassembly. The MODE also governs the process from re-tensioning the reactor vessel bolts through reassembly of the CNV and return to the operating position.

Entry into MODE 4 from MODE 3 requires verification of reactivity condition (keff) less than 0.95 and compliance with footnote (b) of Table 3-2, which prevents reactivity changes during movement and ensures low temperature overpressure events cannot occur.

Overpressure protection in MODE 4 is provided by disconnected, de-energized and therefore open, ECCS reactor vessel vent valves.

The subcriticality and safety of the NPM in this MODE is ensured by the inability to alter the reactivity condition of the core, prevention of overpressure conditions, and the passive cooling of the NPM.

MODE 5 - Refueling This MODE is similar to Refueling in the STS and addresses the conditions during which the reactor pressure vessel (RPV) is not intact, including during movement of the reactor fuel in and around the reactor core in the refueling tool. The upper portion of the CNV and upper portion of the reactor vessel are removed and typically located away from the reactor core during this MODE. The reactor core and lower reactor vessel remain fully submerged in the reactor pool during operations in this MODE. Decay heat removal and core reactivity are ensured by submersion in the reactor pool in this MODE.

3.2 Chapter 2, Safety Limits The structure and content of Chapter 2 of the NuScale GTS closely align with Chapter 2 content of existing reactors. Reactor core safety limits are established to protect the integrity of the reactor coolant system (RCS) and the reactor fuel cladding, which are the two principal physical barriers that guard against the uncontrolled release of radioactivity.

The combination of THERMAL POWER, Reactor Coolant System (RCS) hot temperature (Thot), pressurizer pressure specified in the Core Operating Limits Report (COLR), and safety limits (SLs) on the critical heat flux (CHF) ratio and peak fuel centerline temperature prevent overheating of the fuel and clad, as well as possible cladding perforation that NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 would result in the release of fission products to the reactor coolant. These variables were chosen to best address the conditions and variables in the NuScale design that uses natural circulation to maintain flow through the reactor core.

The SL on RCS pressure protects the integrity of the RCS from overpressurization that could result in a breach of the reactor coolant pressure boundary (RCPB). If such a breach occured in conjunction with a fuel cladding failure, fission products could enter the containment, raising concerns about radioactive releases.

3.3 Chapter 3, Limiting Conditions for Operation and Surveillance Requirements This chapter of the NuScale GTS reflects and addresses the NuScale design and operating paradigm. The integrated, passive, modular NuScale design is reflected in the chapter. As stated in 10 CFR 50.36(c)(2)(i) the limiting conditions for operation (LCO) are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Additionally 10 CFR 50.36(c)(2)(ii) provides four criteria to be applied to specify LCO.

The criteria of 10 CFR 50.36(c)(2)(i) and 10 CFR 50.36(c)(2)(ii) were applied as follows.

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

Consistent with PWR designs, instrumentation is provided to detect significant abnormal leakage from the RCPB. Specification 3.4.7 specifies operability and surveillance requirements for instrumentation to detect leakage and provide indication in the control room of a significant abnormal degradation of the RCPB.

Criterion 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.

As a new design, the NuScale LCO were developed applying the criteria of 10 CFR 50.36(c)(2)(i) and 10 CFR 50.36(c)(2)(ii), consistent with the standard technical specifications and existing plant-specific technical specifications. Safety analyses are performed as described in Chapter 15 and elsewhere in the FSAR.

In general, normal operating conditions are described in relevant portions of the FSAR analyses or system operational descriptions. LCO are not appropriate for many of these conditions as they do not satisfy the criteria of 10 CFR 50.36(c)(2)(i) in that they are not NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 the lowest functional capability or performance level of equipment required for safe operation. Consistent with the Commission final policy on technical specification improvements (58 FR 39132) control over those normal operating conditions are established by the facility operating license by reference to the FSAR. 10 CFR 50.71 maintains the description of the facility consistent with the FSAR. Any proposed changes thereto are subject to 10 CFR 50.59 review and require NRC approval before implementation of the change if an unreviewed safety question is identified.

In other cases, the operating limit is an initial condition of a design basis accident or transient analysis that satisfies the criteria of both 10 CFR 50.36(c)(2)(i) and 10 CFR 50.36(c)(2)(ii). This results in the need for an LCO to ensure the initial condition is met.

As an example, the pressurizer pressure normal operating condition range is established at about 1850 psia, with a normal range from 1780 to 1920 psia (values are illustrative only - not for use.) Operating a NuScale plant with the pressurizer pressure slightly above or below this normal operating range does not exceed a lowest functional capability or performance level required for safe operation. Operation outside the normal operating range would be inconsistent with the description in the FSAR. Operating outside these limits would require corrective actions to be taken. If the normal operating range was to be changed then a 10 CFR 50.59 evaluation and modification of the FSAR for continued operation outside these previously described limits would be required. Based on this, the normal operating range of pressurizer pressure does not require an LCO, although operation within that range should generally be assumed to exist (with appropriate conservative bias) at the beginning of a transient analysis.

Pressurizer pressure limits do define the lowest functional capability or performance level required for safe operation at the limits assumed in the safety analyses for actuations as described in LCO 3.3.1, Module Protection System, and analyses as set by the Core Operating Limits Report (COLR) and referenced by LCO 3.4.1. Safety analyses and the topical reports used to generate the COLR limits may, however are not expected to coincide with the normal operating range.

The unique NuScale design provides extensive conservatism in many parameters which provide flexibility in choosing both normal operating ranges and analysis limits that are demonstrated to be the lowest functional capability or performance level required for safe operation.

The FSAR describes normal operating limits and operating limits that are credited as initial conditions or operating restrictions. The tables below identify the parameters and the limits that satisfy 10 CFR 50.36(c)(2)(i) and 10 CFR 50.36(c)(2)(ii). Where identified, normal operating limits are identified by reference to the FSAR location if they are different from the LCO limits that define the lowest functional capability or performance level required for safe operation.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table 3-3 Parameters and operating restrictions that require an LCO 10 CFR 50.36(c)(2)(i) and 10 CFR Normal Operating Range 50.36(c)(2)(ii) FSAR Reference Parameters and operating restrictions Design core power Yes (Operating License)

Minimum RCS temperature for criticality Yes Same as LCO RCS pressure for criticality Yes 4.4 RCS temperatures Yes 5.1.4 Containment pressure Yes 6.2.1, 6.2.2 RCS flow Yes 5.1.4, 4.4 Shutdown margin (SDM) Yes 4.3.2 Enthalpy rise hot channel factor (FH) Yes Same as LCO Axial offset (AO) Yes Same as LCO CRA positions Yes 4.3.2 RCS specific activity Yes 11.1 Primary-to-secondary leakage Yes Same as LCO Containment leak rate Yes Same as LCO Decay time before fuel handling Yes Same as LCO Design features Fuel design

enrichment No 4.2.2

cladding No 4.2.2

geometry No 4.2.2 Reactor pool

Depth Yes 9.2.5

Temperature Yes 9.2.5

Boron concentration Yes 9.2.5, 9.1.3 The variables listed above are mapped to the GTS in Table A-1.

Some of the LCO limits are established by reference to the COLR; additional parameters in Table 3-4 below are managed and established in the COLR required by TS 5.6.3.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table 3-4 Core design limits that are initial conditions used in design basis event evaluation specified in the Core Operating Limits Report Maximum assembly radial peaking Peak rod exposure Maximum hot zero power (HZP) critical boron concentration Maximum hot full power (HFP) critical boron concentration Minimum refueling boron concentration Most positive moderator temperature coefficient (MTC) at power 25% RTP Most positive MTC at power >25% RTP Most negative MTC at HZP Most negative MTC at HFP Minimum HFP SDM (critical keff)

Minimum HZP SDM (critical keff)

Cold shutdown criterion Maximum fuel pellet enrichment (wt. %)

CRA position uncertainty Boric Acid supply boron concentration Maximum CVCS makeup pump water flow path flowrate Criterion 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.

The FSAR describes SSC and functions that operate or actuate to mitigate design basis accident or transients in the safety analyses of the NuScale design.

The module protection system (MPS) and neutron monitoring system provide instrumentation and actuation functions credited to detect and mitigate design basis NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 events in the analyses. Table 3-5 lists the signals of the MPS that are used in the design basis analyses.

Table 3-5 Module protection system signals used in the Chapter 15 analyses Signal Basis and Protected Limit or Parameter High power Exceeding CHF limits for reactivity and overcooling events Startup and Exceeding CHF and energy deposition limits during intermediate startup power excursions range log power rate High power rate Exceeding CHF limits for reactivity and over-cooling events High startup Exceeding CHF and energy deposition limits during rapid range countrate startup power excursions High subcritical Detect and mitigate inadvertent subcritical boron multiplication dilutions in operating MODES 2 and 3 High RCS hot Exceeding CHF limits for reactivity and heatup events temperature High containment Mitigate RCS or secondary leaks above the allowable pressure limits to protect RCS inventory and ECCS function during these events High pressurizer Protect against exceeding RPV pressure limits for pressure reactivity and heatup events High pressurizer Mitigate CVCS malfunctions to protect against overfilling level the pressurizer Low pressurizer Detect and mitigate primary high-energy line break pressure (HELB) outside the CNV and protect RCS subcooled margin against instability events Low-low Detect and mitigate primary HELB outside the CNV and pressurizer protect RCS subcooled margin against instability events pressure Low pressurizer Mitigate primary HELB outside the CNV and CVCS level malfunctions to protect the pressurizer heaters from uncovering and overheating Low-low Mitigate loss-of-coolant accidents (LOCAs) to protect pressurizer level RCS inventory and ECCS functionality during LOCA and primary HELB outside CNV events Low steam Mitigate secondary HELB outside the CNV to protect SG pressure inventory and DHRS functionality NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Signal Basis and Protected Limit or Parameter Low-low steam Mitigate secondary HELB outside the CNV to protect SG pressure inventory and DHRS functionality High steam Mitigate loss of steam demand to protect primary and pressure secondary pressure limits during heatup events High steam Mitigate SG boil-off to protect DHRS functionality during superheat at-power and post-trip conditions Low steam Mitigate SG overfilling to protect DHRS functionality superheat during at- power and post-trip conditions Low RCS flow Ensure boron dilution cannot be performed at low RCS flowrates because the loop transit time is too long to be able to detect the reactivity change in the core within sufficient time to mitigate the event Low-low RCS Ensure flow remains measureable and positive during flow low power startup conditions High CNV water Protect water level above the core in LOCA events level Low RCS Actuate ECCS during small loss of coolant events to Pressure prevent boron distribution gradients in the RCS Low AC voltage Ensure appropriate load shedding occurs in the highly reliable DC power system in the event of extended loss of AC power to the battery chargers High under-the- Detect high energy leaks or breaks at the top of the NPM bioshield under-the-bioshield to reduce the consequences of HELB temperature on the safety-related equipment located on top of the NPM The NuScale engineered safety feature (ESF) systems consist primarily of the CNV, the ECCS, and the DHRS. Additionally, the reactor pool provides the UHS, and aspects of other design features and safety functions are credited in the accident and transient analyses. Table 3-6 identifies the SSC, features, and safety functions that satisfy Criterion 3 as they may actuate to mitigate a design basis accident or transient.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table 3-6 NuScale structures, systems, and components credited to actuate or function in design basis accident and transient analyses Nuclear fuel, fuel cladding, and fuel assemblies Control rod drives and assemblies MPS, including

neutron monitoring system

reactor trip system (RTS)

engineered safety features actuation system (ESFAS)

Reactor vessel Reactor safety valves Emergency core cooling valves Steam generators DHRS heat exchanger, piping, and valves including main steam and feedwater isolation CNV and containment isolation valves CVCS demineralized water isolation valves Reactor pool The variables listed above are mapped to the GTS in Table A-1.

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

The NuScale Power Plant is designed to a fundamentally different safety paradigm and has no design-specific operating experience. Probabilistic risk assessment results show that while core damage frequencies (CDFs) for the previously operating nuclear power plants typically fall between 10-5 to 10-6 per reactor year, the NuScale CDF is approximately 10-8 per reactor year.

The NRC guidance on risk-informed applications was developed in the context of risk results calculated for large PWRs. Regulatory Guide 1.174 (Reference 6.2.14) describes NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 changes in CDF that the NRC has considered acceptable when making permanent changes to a plants licensing basis. This provides a relative scale for assessing whether the function of an SSC is significant to the public health and safety, i.e., in the region approaching or greater than 10-6 per reactor year.

NuScale used risk information in many decision-making processes employed to develop a safe, economical, and efficient design. These include risk-informed SSC categorization, risk-informed inservice inspections and inservice testing, design reliability assurance program, as well as assessing specific design issues as risk-significant or not risk-significant.

Based on the combination of orders-of-magnitude difference between the NuScale CDF and the NRC-accepted CDF, and programs used to ensure that the CDF remains at extremely low levels, no non-safety related SSC or functions approached the 10-6 per reactor year criteria for inclusion in the TS in accordance with Criterion 4.

Based on criterion 4 operating experience, LCOs are proposed for:

manual actuation functions of the MPS

the remote shutdown station

nonsafety-related main steam line and feedwater isolation functions

refueling neutron monitoring instrumentation The MPS allows manual actuation of protective functions, however this capability is not credited in the design; it is included based on industry inclusion as a Criterion 4 condition.

The remote shutdown station does not have an active function in the NuScale design other than monitoring conditions; it is included based on industry inclusion as a Criterion 4 condition. Similarly, the refueling neutron monitoring instrumentation does not provide an actuation; however, it does provide information about the reactor core condition during conduct of refueling activities and is included based on Criterion 4.

The backup nonsafety-related main steam line and feedwater line isolation valves and their automatic actuation were identified for inclusion by application of the design reliability assurance program and are also included in the TS in accordance with Criterion 4.

The functions listed above are mapped to the GTS in Table A-1.

Other Limiting Conditions of Operation Other requirements or limitations in the Limiting Conditions of Operation that are included but that do not satisfy any of the criteria in 10 CFR 50.36(c)(2).

After receiving input from the staff, NuScale set RCS specific activity limits in LCO 3.4.8 at a level consistent with the dose rates that may exist in adjoining areas based on calculated design basis source terms. This limit is not consistent with 10 CFR 50.36, NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 however is conservative compared with RCS specific activity limits established at traditional plants based on safety analysis assumptions, and therefore commonly established in accordance with criterion 2 above.

Similarly, after staff provided input to NuScale to establish a limit on in-containment secondary side piping leakage. The LCO was included because if a seismic event occurs when the in-containment secondary leakage is greater than the LCO limit, a main steam or feedwater pipe break could occur. This could result in an adverse interaction between the affected in-containment secondary system piping and other safety related equipment located inside the containment.

3.4 Chapter 4, Design Features The structure, content, and level of detail of Chapter 4 of the NuScale GTS directly align with Chapter 4 content typical of PWRs. A description of the site location, the site and exclusion area boundaries, and the low population zone around the reactor are to be provided by combined license (COL) applicants, consistent with the descriptions provided in the FSAR.

A description of the reactor core is provided, including the number of fuel assemblies and the materials used in their construction. A description of the CRA makeup and arrangement is provided.

Additionally, a description of the storage of new and irradiated fuel assemblies, including measures to prevent inadvertent criticality, limit exposures associated with storage, and the overall capacity of the storage area is provided.

This content directly aligns with the requirements of 10 CFR 50.36(c)(4). Other features of the facility that could have a significant effect on safety are described in Chapter 2 or 3 of the TS.

3.5 Chapter 5, Administrative Controls The structure and content of Chapter 5 of the NuScale GTS closely aligns with Chapter 5 content typical of STS for PWRs. Deviations occur primarily in Section 5.5, Programs and Manuals, because the NuScale design does not include the features that require the identified program or manual. A comparison of the contents of Section 5.5 and explanation of omissions are provided in Table 3-7 below.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table 3-7 Comparison of NuScale generic technical specifications with NUREG-1431 Section 5.5 contents NuScale GTS Section 5.5 NUREG-1431 Section 5.5 5.5.1 Offsite Dose Calculation Manual 5.5.1 Offsite Dose Calculation Manual (ODCM) (ODCM)

The NuScale design does not include safety- 5.5.2 Primary Coolant Sources Outside related RCS flow loops outside the Containment containment. See AP1000 SER Chapter 20, TMI Item III.D.1 The NuScale design does not include a post- 5.5.3 [Post Accident Sampling]

accident sampling system that contributes significantly to plant safety and accident recovery. See travelers 366, 413, 442, and NUREG-2194.

5.5.2 Radioactive Effluent Controls Program 5.5.4 Radioactive Effluent Controls Program 5.5.3 Component Cyclic or Transient Limit 5.5.5 Component Cyclic or Transient Limit The NuScale design does not include a pre- 5.5.6 [Pre-Stressed Concrete Containment stressed concrete containment. Tendon Surveillance Program]

The NuScale design does not include reactor 5.5.7 Reactor Coolant Pump Flywheel coolant pumps. Inspection Program Relocated consistent with the incorporation of 5.5.8 Inservice Testing Program traveler 545.

5.5.4 Steam Generator (SG) Program 5.5.9 Steam Generator (SG) Program 5.5.5 Secondary Water Chemistry Program 5.5.10 Secondary Water Chemistry Program The NuScale design does not credit 5.5.11 Ventilation Filter Testing Program ventilation filtration; therefore, the GTS do not include ventilation filtration.

5.5.6 Explosive Gas and Storage Tank 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program Radioactivity Monitoring Program The NuScale design does not credit 5.5.13 Diesel Fuel Oil Testing Program emergency diesel generators.

5.5.7 Technical Specifications (TS) Bases 5.5.14 Technical Specifications (TS) Bases Control Program Control Program 5.5.8 Safety Function Determination 5.5.15 Safety Function Determination Program (SFDP) Program (SFDP) 5.5.9 Containment Leakage Rate Testing 5.5.16 Containment Leakage Rate Testing Program Program The NuScale design does not include safety- 5.5.17 Battery Monitoring and Maintenance related batteries. Program The NuScale design does not include 5.5.18 Control Room Envelope (CRE) specifications that result in a need for a Habitability Program system level OPERABILITY program.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 NuScale GTS Section 5.5 NUREG-1431 Section 5.5 5.5.10 Setpoint Program (SP) 5.5.19 [Setpoint Control Program]

5.5.11 Surveillance Frequency Control 5.5.20 [Surveillance Frequency Control Program Program]

5.5.12 Spent Fuel Storage Rack Neutron 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring Program Absorber Monitoring Program Programs were omitted either because the associated SSC do not exist as credited features in the NuScale design, or the program was removed or relocated in accordance with associated industry / NRC STS travelers.

3.6 Mapping of 10 CFR 50.36 Selected Limits to Proposed Technical Specifications The parameters, SSC, and functions identified in Sections 3.1 through 3.5 above are correlated with the proposed TS in Appendix A.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 4.0 Comparison with Standard Technical Specifications The NuScale power plant design is different from previously licensed nuclear power plants. Plant operations are also different from previously operating nuclear power plants.

Experience and lessons learned from the improved technical specifications were extensively considered during development of the proposed GTS. However, in many safety-related functional areas the NuScale GTS do not directly correspond with similarly named systems or components. As described in Section 2.0 above, the GTS were developed without direct reliance on previously existing generic technical specifications.

Development required consideration of the applicable regulations, accompanying Commission policies and statements of consideration, evolution of the STS, and industry experience including the industry/NRC STS change travelers.

Consideration of the contents of the STS and travelers does not imply direct correspondence or functional equivalent unless described as such. The NuScale design is not addressed in the traveler process, so none of the travelers are explicitly or directly applicable to the NuScale GTS. Rather the intent of the traveler was considered based on available information related to the changes made or proposed to the STS. The following sections provide additional details regarding the extent of consideration and where appropriate, the adoption of similar controls and limitations in the proposed NuScale GTS.

4.1 Specification Comparisons As described in Section 3, the NuScale GTS were developed to be consistent with the NuScale-specific safety analyses and the design-specific probabilistic risk analyses.

Additionally, the STS were used as a basis for the content and format of the proposed GTS.

The STS are published by the NRC as six NUREGs tailored to various reactor designs.

The versions of the STS that were considered during GTS development included:

Manuscript Current Completion Publication NUREG Design Revision Date Date NUREG-1430 Babcock and Wilcox 4 October 2011 April 2012 NUREG-1431 Westinghouse 4 October 2011 April 2012 NUREG-1432 Combustion Engineering 4 October 2011 April 2012 NUREG-1433 General Electric BWR4 4 October 2011 April 2012 NUREG-1434 General Electric BWR6 4 October 2011 April 2012 NUREG-2194 Westinghouse AP1000 0 December 2015 April 2016 NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Additionally, the NRC has issued the design certification for the ESBWR which included GTS. The specifications and bases provided in the GTS for the ESBWR (Reference 6.2.13) were also consulted as a basis of comparison during preparation of some parts of the NuScale proposed TS.

The focus of the development and comparisons were generally against the Westinghouse, CE, and AP1000 STS. Appendix B provides a comparison of the contents of those STS with the contents of the proposed NuScale GTS.

4.2 Industry/NRC STS Traveler Consideration The previously existing nuclear electric power generation industry works with the NRC staff in the STS maintenance and change process through the Technical Specification Task Force (TSTF), a joint activity of the PWR, boiling water reactor (BWR), and AP1000 Owners Groups. The TSTF coordinates with the NRC to implement the change process (referred to as the traveler process) that maintains the STS and has periodically resulted in publication of revisions to the STS. Although no travelers were developed with direct consideration of the NuScale design or proposed GTS, information regarding travelers available to NuScale through the June 30, 2018 was considered during preparation of the NuScale GTS.

Appendix C of this report describes the consideration and extent of incorporation of traveler contents and intentions. The scope of the Appendix is generally limited to travelers issued since the last published revision of the NUREG STS listed in Section 4.1 above.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 5.0 Conformance with Standard Technical Specification Writers Guide The guidance provided in the Writers Guide (Reference 6.2.4) for Plant Specific Improved Technical Specifications, TSTF-GG-05-01, Revision 1, August 2010 was used to prepare the proposed NuScale GTS. Portions of the guide are specific to BWRs or PWRs and in some cases conformance was not possible or was otherwise inappropriate.

Chapter 1 and specifications LCO 3.0 and SR 3.0 of the proposed specifications are minimally modified to conform to the facility design and specifications. This results in formatting generally consistent with the Writers Guide. Additional emphasis was placed on conformance with the expectations described in section 4 of the guide related to content of each portion of the TS and Bases.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 6.0 References 6.1 Source Documents 6.1.1 10 CFR 20, Standards for Protection Against Radiation.

6.1.2 10 CFR 50, Domestic Licensing of Production and Utilization Facilities.

6.1.3 10 CFR 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

6.2 Referenced Documents 6.2.1 U.S. Code of Federal Regulations, Technical Specifications, Section 50.36, Part 50, Chapter I, Title 10, Energy, (10 CFR 50.36).

6.2.2 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).

6.2.3 U.S. Nuclear Regulatory Commission, Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, Federal Register, Vol.

52 FR 3788, February 6, 1987.

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

6.2.5 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Revision 4.0, Volumes 1 and 2, April 2012.

6.2.6 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Revision 4.0, Volumes 1 and 2, April 2012.

6.2.7 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Revision 4.0, Volumes 1 and 2, April 2012.

6.2.8 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/4 Plants, NUREG-1433, Revision 4.0, Volumes 1 and 2, April 2012.

6.2.9 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/6 Plants, NUREG-1434, Revision 4.0, Volumes 1 and 2, April 2012.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 6.2.10 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Advanced Passive 1000 (AP1000) Plants, NUREG-2194, Volumes 1 and 2, April 2016.

6.2.11 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.

6.2.12 Murley, T.E., U.S. Nuclear Regulatory Commission, letter to Walter S. Wilgus, B&W Owners Group, May 9, 1988, Agencywide Document Access and Management System (ADAMS) Accession No. ML11264A057.

6.2.13 GE-Hitachi Nuclear Energy, ESBWR Design Control Document, Tier 2 and Generic Technical Specifications, Chapter 16, Technical Specifications, and Chapter 16B, Bases, Revision 10, April 2014.

6.2.14 Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, May 2011.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 7.0 Appendices Appendix A Criteria for Inclusion of Technical Specifications Appendix B Summary Comparison of Standard Technical Specifications with NuScale Generic Technical Specifications Contents Appendix C Technical Specification Task Force Traveler Evaluations NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Appendix A Criteria for Inclusion of Technical Specifications The accident analyses provided in the DCA with emphasis on Final Safety Analysis Report (FSAR) Chapter 15 and supporting calculations were reviewed to identify SSC and parameters that satisfy regulatory requirements and are discussed in Sections 3.1 through 3.5. This appendix correlates the identified parameters, SSC, and functions with the proposed TS. The results provided in tables below.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Table A-1 Technical specifications inclusion criteria Parameter, SSC, or Function Relevant Associated NuScale Discussion 10 CFR 50.36 Technical Specifications Criteria or Reference Reactor core safety limits (c)( 1)(i)(A) 2.1.1 Reactor Core SLs Consistent with and similar to large plant safety limits with variables selected based on consideration of natural circulation flow through the reactor core.

critical heat flux ratio

peak fuel centerline temperature Safety limit on RCS pressure (c)(1)(i)(A) 2.1.2 RCS Pressure SL Consistent with and similar to large PWR SLs.

Safety limit violations (c)(1)(i)(A) 2.2 Safety Limit Violations Consistent with and similar to large PWR SLs.

Design core power (c)(2)(ii)(B) Facility Operating License This is an assumed input parameter of safety analyses calculations. It is maintained by manual and automatic controls in accordance with operating procedures.

Reactor coolant system (RCS) (c)(2)(ii)(B) 1.1 Definition of MODES as provided in Table 1.1-1 These are assumed input parameters of safety analyses calculations. They are maintained by manual temperatures and automatic controls in accordance with operating procedures.

3.4.1 RCS Pressure and Temperature Critical Heat Flux (CHF) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 5.6.4 Pressure Temperature Limits Report (PTLR)

Pressurizer pressure (c)(2)(ii)(B) 3.4.1 RCS Pressure and Temperature Critical Heat This is an assumed input parameters of safety analyses calculations. It is maintained by manual and Flux (CHF) Limits automatic controls in accordance with operating procedures.

3.4.3 RCS Pressure and Temperature (P/T) Limits 5.6.4 Pressure Temperature Limits Report (PTLR)

Pressurizer level (c)(2)(ii)(B) 3.3.1 Module Protection System Instrumentation Operation within conservative normal operating limits is ensured by manual and automatic controls in accordance with operating procedures. This is also an actuation parameter credited in safety analyses.

5.5.10 Setpoint Program (SP)

Protective actuation setpoints are established in accordance with the Setpoint Program such that protective features will initiate at established limits.

Containment pressure (c)(2)(ii)(A) 3.3.1 Module Protection System Instrumentation Operation within conservative normal operating limits is ensured by manual and automatic controls in accordance with operating procedures. Operation within limits is also ensured by the OPERABILITY (c)(2)(ii)(B) 5.5.10 Setpoint Program (SP) requirements of LCO 3.4.7, RCS Leakage Detection Instrumentation. This is also an actuation 3.4.7 RCS Leakage Detection parameter credited in safety analyses. Protective actuation setpoints are established in accordance with the Setpoint Program such that protective features will initiate at established limits.

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Parameter, SSC, or Function Relevant Associated NuScale Discussion 10 CFR 50.36 Technical Specifications Criteria or Reference Steam pressure (c)(2)(ii)(B) 3.3.1 Module Protection System Instrumentation This is an actuation parameter credited in safety analyses. Operation within the assumed limits is ensured by manual and automatic controls in accordance with operating procedures. Protective 5.5.10 Setpoint Program (SP) actuation setpoints are established in accordance with the Setpoint Program such that protective features will initiate at established limits.

Feedwater temperature (c)(2)(ii)(B) 3.3.1 Module Protection System Instrumentation This is an actuation parameter credited in safety analyses. Operation within the assumed limits is ensured by manual and automatic controls in accordance with operating procedures. Protective 5.5.10 Setpoint Program (SP) actuation setpoints are established in accordance with the Setpoint Program such that protective features will initiate at established limits.

RCS flow (c)(2)(ii)(B) Natural phenomenon This is an actuation parameter credited in safety analyses. It is ensured by physical phenomena caused by density differences in the natural circulation in the RCS. Protective actuation setpoints are 3.3.1 Module Protection System Instrumentation established in accordance with the Setpoint Program such that protective features will initiate at 5.5.10 Setpoint Program (SP) established limits.

Shutdown margin (c)(2)(ii)(B) 3.1.1 SHUTDOWN MARGIN This is an assumed input parameter of the safety analyses calculations. It is maintained by manual and automatic controls in accordance with operating procedures.

3.1.8 PHYSICS TESTS Exceptions Control rod assembly positions (c)(2)(ii)(B) 3.1.4 Rod Group Alignment Limits These are assumed input parameters of the safety analyses calculations. It is maintained by manual and automatic controls in accordance with operating procedures. LCO 3.1.7 supports the parameter 3.1.5 Shutdown Group Insertion Limits determination by ensuring OPERABLE indication is used to verify compliance with those LCO.

3.1.6 Regulating Group Insertion Limits 3.1.7 Rod Position Indication RCS specific activity (c)(2)(ii)(B) 3.4.8 RCS Specific Activity This is an assumed input parameter of certain safety analyses calculations. However, the limits were restricted further and established in response to NRC staff direction. The RCS specific activity limits in Input from LCO 3.4.8 are set at a level consistent with the dose rates that may exist in adjoining areas based on NRC staff calculated design basis source terms. These limits are not consistent with the safety analyses; however, other than they are conservative compared with RCS specific activity limits established at traditional plants based criteria of on safety analysis assumptions. RCS Specific Activity is monitored and maintained in accordance with 10 CFR 50.36.

plant procedures.

Primary to secondary leakage (c)(2)(ii)(B) 3.4.5 RCS Operational LEAKAGE This is an assumed input parameter of certain safety analyses calculations. It is monitored and maintained in accordance with plant procedures.

Containment leak rate (c)(2)(ii)(B) 3.6.1 Containment - Operating This is an assumed input parameter of certain safety analyses calculations. It is monitored and maintained in accordance with plant programs and procedures.

3.6.2 Containment Isolation Valves 5.5.9 Containment Leakage Rate Testing Program NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Parameter, SSC, or Function Relevant Associated NuScale Discussion 10 CFR 50.36 Technical Specifications Criteria or Reference In-containment secondary piping Input from 3.7.3 In-Containment Secondary Piping Leakage This establishes a limit implementing controls and actions to be taken if secondary piping leakage leakage NRC staff exceeds those used to satisfy the leak-before-break criteria as applied to piping inside the containment.

other than criteria of 10 CFR 50.36 Decay time before fuel handling (c)(2)(ii)(B) 3.8.2 Decay Time This is an assumed input parameter of certain safety analyses calculations. It is met in accordance with plant procedures.

Fuel design (c)(2)(ii)(B) 2.1.1 Reactor Core SLs These are assumed input parameters of certain safety analyses calculations. It includes design features of the fuel and assemblies, and it is maintained in accordance with plant procedures.

enrichment 3.1.2 Core Reactivity

cladding 3.1.8 PHYSICS TESTS Exceptions

geometry 4.2 Reactor Core Core design (c)(2)(ii)(B) 2.1.1 Reactor Core SLs These are assumed input parameters of safety analyses calculations. It includes design features of the fuel and assemblies, and it are maintained in accordance with plant design, programs, and procedures.

reactivity 3.1.1 SHUTDOWN MARGIN (SDM)

temperature coefficients of 3.1.2 Core Reactivity reactivity 3.1.3 Moderator Temperature Coefficient (MTC)

core power distribution 3.1.4 Rod Group Alignment Limits

fuel burnup 3.1.5 Shutdown Group Insertion Limits 3.1.6 Regulating Group Insertion Limits 3.1.7 Rod Position Indication 3.1.8 PHYSICS TESTS Exceptions 3.2.1 Enthalpy Rise Hot Channel Factor (FH) 3.2.2 AXIAL OFFSET (AO) 3.4.1 RCS Pressure, Temperature, and Flow Resistance Critical Heat Flux (CHF) Limits 3.4.2 RCS Minimum Temperature for Criticality 4.2 Reactor Core 5.6.3 Core Operating Limits Report (COLR)

NuScale Nonproprietary

Technical Specifications Regulatory Conformance and Development NuScale Nonproprietary TR-1116-52011-NP Rev. 4 Parameter, SSC, or Function Relevant Associated NuScale Discussion 10 CFR 50.36 Technical Specifications Criteria or Reference Fuel storage and handling system (c)(2)(ii)(B) 3.5.3 Ultimate Heat Sink These are assumed input parameters of safety analyses calculations. They consist of design features of the fuel storage location and racks.

geometry 4.3 Fuel Storage

location 5.5.12 Spent Fuel Storage Rack Neutron Absorber Monitoring Program Reactor pool (c)(2)(ii)(B) 3.5.3 Ultimate Heat Sink These are assumed input parameters of safety analyses calculations. They are maintained in accordance with plant procedures.

depth

temperature

boron concentration High range linear power (c)(1)(ii)(A) 3.3.1 Module Protection System Instrumentation This function is credited with detecting when the parameter is out of limits, initiating a reactor trip, and actuating the associated engineered safety features. The limit(s) constitute a limiting safety setpoint as (c)(2)(ii)(C) 3.3.2 Reactor Trip System (RTS) Logic and Actuation described in 5.5.10, the Setpoint Program. The limit(s) are established in accordance with the Setpoint 3.3.3 Engineered Safety Features Actuation System Program.