ML23355A271

Sdaa Chapter 13, Conduct of Operations, SE with No Open Items

ML23355A271
Person / Time
Issue date:	03/11/2024
From:	Vivanco R NRC/NRR/DNRL/NRLB
To:	NuScale
Vivanco, R., NRR/DNRL, 301-415-0021
Shared Package
ML23355A269	List: ML23355A271
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Download: ML23355A271 (45)

Text

CONDUCT OF OPERATIONS

This chapter of the safety evaluation report (SER) documents th e U.S. Nuclear Regulatory Commission (NRC) staffs (the staff) review of Chapter 13, Con duct of Operations, of the NuScale Power, LLC (NuScale or the applicant), Standard Design Approval Application (SDAA),

Part 2, Final Safety Analysis Report (FSAR). The staffs regu latory findings documented in this report are based on Revision 1 of the SDAA, dated October 31, 2 023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23306A033). The precise parameter values, as reviewed by the staff in this SER, are pro vided by the applicant in the SDAA using the English system of measure. Where appropriate, th e NRC staff converted these values for presentation in this SER to the International System (SI) units of measure based on the NRCs standard convention. In these cases, the SI converted value is approximate and is presented first, followed by the applicant provided parameter v alue in English units within parentheses. If only one value appears in either SI or English units, it is directly quoted from the SDAA and not converted.

Organizational Structure

Introduction

A combined license (COL) applicants organizational structure i ncludes the corporate-level management and technical support organization and the onsite op erating organization. The management and technical support organization includes the corp orate or home office offsite organization; associated functions, activities, and responsibil ities; and the approximate number and qualifications of offsite personnel necessary to ensure tha t sufficient technical resources have been, are being, and will continue to be provided to accom plish the safe design, construction, testing, and operation of the nuclear plant. The onsite operating organization includes the structure, functions, activities, responsibilities, and approximate number and qualifications of onsite personnel necessary to safely operate and maintain the facility.

The staff reviewed the FSAR to evaluate the COL information ite ms that pertain to (1) COL applicant descriptions of the corporate-level management and te chnical support organization and (2) COL applicant descriptions of the onsite operating orga nization.

Summary of Application

The plans for a corporate-level, technical, and onsite organiza tional structure to support, design, construct, test, operate, and maintain the nuclear plant are no t within the scope of the NuScale SDAA. This responsibility resides with a COL applicant referenc ing the NuScale Power Plant US460 standard design. Chapter 13 of the application specifies combined license information items for the COL applicant to describe the corporate-level man agement and technical support organization and the onsite operating organization.

ITAAC: NuScale has not proposed any inspections, tests, analyses, and acceptance criteria (ITAAC) related to this area of review.

Technical Specifications: There are no technical specifications (TS) for this area of rev iew.

Technical Reports: No technical reports (TRs) are associated with this area of rev iew.

13-1 Regulatory Basis

Section 13.1.1, Management and Technical Support Organization, and Section 13.1.2-13.1.3, Operating Organization, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), identify, in part, the relevant NRC regulatory requirements for organizational structure and th e associated acceptance criteria.

The following regulatory requirements apply to the organization al structure:

• Title 10 of the Code of Federal Regulations (10 CFR) 50.34(f)(3)(vii), as it pertains to requirements related to lessons learned from the accident at Th ree Mile Island (TMI) for the standard design approval applicant, in part, to describe th e management plan for design and construction activities of the proposed plant

• 10 CFR 50.40(b), which requires the COL applicant to be techni cally qualified to engage in activities associated with the design, construction, and ope ration of a nuclear power plant

• 10 CFR 50.48(a)(1)(ii), as it pertains to information that mus t be included in the fire protection plan of the holder of a COL under 10 CFR Part 52, L icenses, Certifications, and Approvals for Nuclear Power Plants,; specifically, the ide ntification of the various positions within the licensees organization that are responsib le for the program

• 10 CFR 50.54(i), (j), (k), (l), and (m), as they pertain to th e organizational staffing requirements for, and responsibilities of, operators and senior operators licensed under 10 CFR Part 55, Operators Licenses

• 10 CFR Part 50, Domestic Licensing of Production and Utilizat ion Facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plant s and Fuel Reprocessing Plants, as it pertains to organizational responsi bilities for the establishment and execution of the quality assurance program

• 10 CFR 52.79(a)(26)-(28) and (29)(i), as they pertain to infor mation that must be included in the FSAR that is submitted as part of the applicati on for a COL, specifically, the following:

(1) the applicants organizational structure, allocations or re sponsibilities and authorities, and personnel qualifications requirements for oper ation

(2) managerial and administrative controls to be used to ensure safe operation as established in 10 CFR Part 50, Appendix B

(3) plans for preoperational testing and initial operations

(4) plans for the conduct of normal operations

13-2 The related acceptance criteria are as follows:

• Section III, Acceptance Criteria, of SRP section 13.1.1, Rev ision 6, issued August 2016

• Section III, Acceptance Criteria, of SRP section 13.1.2-13.1.3, Revision 7, issued August 2016

• Review Criterion 6.4(2) in Section 6, Staffing and Qualificat ions, of NUREG-0711, Human Factors Engineering Program Review Model, Revision 3, i ssued November 2012 (ML12324A013)

• NUREG-1791, Guidance for Assessing Exemption Requests from th e Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 C FR 50.54(m), issued July 2005 (ML052080125)

• NUREG/CR-6838, Technical Basis for Regulatory Guidance for As sessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffin g Requirements Specified in 10 CFR 50.54(m), issued February 2004 (ML04058028 9)

Technical Evaluation

The COL applicant is responsible for describing the corporate-l evel management and technical support organization and the onsite operating organization. Thi s section presents the staffs evaluation of the COL information items that pertain to the COL applicants organizational structures.

13.1.4.1 Combined License Information Items

13.1.4.1.1 Management and Technical Support Organization

SRP section 13.1.1 states that for the management and technical support organization, the COL applicants safety analysis report (SAR) should do the followin g:

• Describe the qualification requirements for each identified po sition or class of positions that provide technical support to the onsite operating organiza tion.

• Specify the qualification requirements for individuals holding management and supervisory positions in organizational units that provide supp ort to the onsite operating organization.

In Section 13.1.1, Management and Technical Support Organizati on, of the FSAR, NuScale specified COL Item 13.1-1, which directs the COL applicant to d escribe the corporate-level or home office management and technical support organization and s pecify the necessary qualification requirements for positions within the management and technical support organization that provide technical support to the onsite opera ting organization. The staff finds that COL Item 13.1-1 appropriately addresses the information th at the COL applicant should provide for corporate-level management and technical support or ganizations.

13-3 13.1.4.1.2 Operating Organization

SRP section 13.1.2-13.1.3 states that the COL applicants SAR s hould describe (1) the structure, functions, and responsibilities of the onsite operat ing organization established to operate and maintain the plant and (2) any alternatives to the requirements involving the number of licensed personnel, as specified in 10 CFR 50.54(m). Consistent with the SRP, in chapter 13 of the SDAA, Section 13.1.2, Operating Organization, COL Item 13.1-2 directs the COL applicant to describe the onsite operating organization, in cluding the organizations structure, functions, and responsibilities. In addition, COL It em 13.1-2 specifies that the proposed operating staff shall be consistent with the minimum l icensed operator staffing requirements in section 18.5, Staffing and Qualification, of the SDAA.

In SDAA section 18.5, the applicant described a staffing level and qualifications analysis that is an alternative to the requirements of 10 CFR 50.54(m). Within t he context of its review of SDAA chapter 13, the staff concludes that it is acceptable for the C OL item to reference the discussion in section 18.5. Chapter 18 of the SER describes the staffs ev aluation of the staffing and qualification element of the NuScale Human Factors Engineering (HFE) Program. Accordingly, the staff determined that COL Item 13.1-2 appropriately identif ies information the COL applicant needs to provide about the onsite operating organization.

SRP section 13.1.2-13.1.3 states that the COL applicants SAR s hould describe the education, training, and experience requirements (qualification requiremen ts) that the COL applicant established to fill each management, operating, technical, and maintenance position category in the operating organization. In the SDAA, Section 13.1.3, Quali fications of Nuclear Plant Personnel, COL Item 13.1-3 directs the COL applicant to descri be the qualification requirements for each of the identified position categories for the operating organization.

Accordingly, the staff determined that COL Item 13.1-3 appropri ately identifies information the COL applicant needs to provide regarding the qualification requ irements for the operating organization.

Combined License Information Items

Table 13.1-1 lists COL information related to the organizationa l structure from FSAR Table 1.8-1, Combined License Information Items.

Table 13.1-1 NuScale COL Information Items Related to FSAR sec tion 13.1

SDAA Item No. Description Part 2 Chapter 13

COL Item 13.1.1 13.1-1 An applicant that references the NuScale Power Plant US460 standard design will provide a description of the corporate or home office management and technical support organization, including a description of the qualification requirements for (1) each identified position or class of positions that provide technical support to the on-site operating organization, and (2) individuals holding management and supervisory positions in organizational units providing technical support to the on-s ite operating organization.

13-4 SDAA Item No. Description Part 2 Chapter 13

COL Item 13.1.2 13.1-2 An applicant that references the NuScale Power Plant US460 standard design will provide a description of the proposed structure, functions, and responsibilities of the on-site organization necessary to operate and maintain the plant. The proposed operating staff shall be consistent with the minimum licensed operator staffing requirements in Section 18.5.

COL Item 13.1.3 13.1-3 An applicant that references the NuScale Power Plant US460 standard design will provide a description of the qualification requirements for each management, operating, technical, and maintenance position described in the operating organization.

Conclusion

For the reasons given above, the staff concludes that the COL i nformation items specified in table 13.1-1 and included in the FSAR are sufficient to identif y information that the COL applicant needs to provide to meet the applicable requirements of 10 CFR 50.34(f)(3)(vii);

10 CFR 50.40(b); 10 CFR 50.48(a)(1)(ii); 10 CFR 50.54(i), (j), (k), and (l); 10 CFR Part 50, Appendix B; 10 CFR 52.79 (a)(26)-(28) and (29)(i); and the mini mum licensed operator staffing requirements in lieu of 10 CFR 50.54(m) as described in FSAR Se ction 18.5.

Training

Introduction

A COL applicants training program should include (1) the initi al license training program for reactor operators and senior reactor operators, (2) the license d operator requalification program, and (3) the nonlicensed plant staff training program. The latter consists of initial training, periodic retraining, and qualifications for nonlicens ed operators, shift supervisors, shift technical advisors, instrumentation and control technicians, el ectrical maintenance personnel, mechanical maintenance personnel, radiological protection techn icians, chemistry technicians, and engineering support personnel.

The staff reviewed the SDAA to evaluate the COL information ite ms that pertain to the COL applicants description of and schedule for (1) the licensed op erator training program for reactor operators and senior reactor operators, including the licensed operator requalification program, and (2) the training program for the nonlicensed plant staff.

Summary of Application

FSAR: The development of site-specific training programs is not withi n the scope of the NuScale SDAA. This responsibility resides with the COL applican t. In FSAR Section 13.2, Training, the SDA applicant specified two COL information ite ms that direct the COL applicant to describe the initial license training program, the licensed operator requalification program, and the nonlicensed plant staff training program and to provide schedules for these programs.

ITAAC: The applicant has not proposed any ITAAC related to this area o f review.

13-5 Technical Specifications: There are no TS for this area of review.

Technical Reports: No TRs are associated with this area of review.

Regulatory Basis

SRP Section 13.2.1, Reactor Operator Requalification Program; Reactor Operator Training, and SRP Section 13.2.2, Non-Licensed Plant Staff Training, id entify, in part, the relevant NRC regulatory requirements for training and the associated accepta nce criteria.

The following regulatory requirements are applicable for traini ng:

• 10 CFR 19.12, Instruction to workers, as it pertains to inst ructions provided to workers for the protection of personnel from exposure to radiation or r adioactive material

• 10 CFR 26.29, Training, as it pertains to employee training associated with the fitness-for-duty (FFD) program

• 10 CFR 50.34(f)(2)(ii), as it pertains to the TMI-related requ irement for applicants to establish a program to begin during construction and to follow into operation for assessing and improving plant procedures applicable to operator training

• 10 CFR 50.40(a) and (b), as they pertain to the issuance of a COL under 10 CFR Part 52 based on considerations of whether the applicant (1) is technically qualified to engage in activities associated with the design, c onstruction, and operation of a nuclear power plant and (2) has established the licensed a nd nonlicensed plant staff training programs necessary to provide reasonable assurance tha t the nuclear power plant can be safely operated

• 10 CFR 50.54(i-1), as it pertains to requirements for the esta blishment of a licensed operator requalification training program within 3 months after the date that the Commission makes the finding under 10 CFR 52.103(g) that the ac ceptance criteria in the COL are met

• 10 CFR 50.120(b)(1)-(3), as they pertain to requirements for t he establishment, implementation, and maintenance of training programs derived fr om a systems approach to training as defined in 10 CFR 55.4, Definitions, for speci fic categories of nuclear power plant personnel

• 10 CFR Part 50, Appendix B, as it pertains to the training and technical qualifications of personnel who perform activities that affect the quality of str uctures, systems, and components (SSCs) that are cover ed by the quality assurance pro gram

• 10 CFR Part 50, Appendix E, Emergency Planning and Preparedne ss for Production and Utilization Facilities, as it pertains to the requirements for emergency preparedness training of employees and other persons whose assistance may be needed in the event of a radiological emergency (e.g., local emergency services and law enforcement personnel), including participation in drill and exercise scena rios to provide performance opportunities to develop, maintain, and demonstrate key skills

13-6

• 10 CFR 52.79(a)(26)-(28) and (29)(i), as they pertain to infor mation to be included in the COL FSAR, specifically, the following:

(1) the qualification requirements of licensed and nonlicensed plant personnel to engage in activities associated with operation of the nuclear p ower plant

(2) the controls associated with the training of personnel who perform activities that affect the quality of SSCs that are covered by the quality assu rance program as established in 10 CFR Part 50, Appendix B

(3) plans for licensing personnel and training nonlicensed plan t staff before criticality to support preoperational testing activities and initial operat ions

(4) plans for licensed and nonlicensed plant staff to receive t he technical and administrative training required to operate, test, and maintain the nuclear power plant during the conduct of normal operations

• 10 CFR 52.79(a)(14), (21), (33), (34), (36), (39), (40), and ( 44), as they pertain to information that must be included in the FSAR that an applicant submits as part of the application for a COL, specifically, descriptions of the follow ing:

(1) licensed operator training required by 10 CFR Part 55

(2) training required by 10 CFR 50.120, Training and qualific ation of nuclear power plant personnel, for specific categories of nuclear power plan t personnel

(3) nonlicensed plant staff training associated with security procedures, radiological emergency plans, radiation protection, fire protection, and FFD

• 10 CFR 55.4, as it pertains to Commission-approved training pr ograms that are based on a systems approach to training

• 10 CFR 55.31(a)(4)-(5), as they pertain to the documentation r equirements associated with successful completion by an applicant for an operator lice nse of a facility licensee training program, when the facility licensee requests administr ation of the licensing exam (i.e., written examination and operating test)

• 10 CFR 55.41, Written examination: Operators, as it pertains to requirements associated with the content and makeup of the NRCs written exa mination for operators

• 10 CFR 55.43, Written examination: Senior operators, as it p ertains to requirements associated with the content and makeup of the NRCs written exa mination for senior operators

• 10 CFR 55.45, Operating tests, as it pertains to requirement s associated with (1) the content and makeup of the NRCs operating test for operators an d senior operators and (2) the use of a Commission-approved simulation facility, a pla nt-referenced simulator, or the physical plant for administration of the operating test

13-7

• 10 CFR 55.46, Simulation facilities, as it pertains to requi rements for the use of simulation facilities in the administration of the NRC operatin g test

• 10 CFR 55.59, Requalification, as it pertains to requirement s associated with licensed operator requalification training programs

The related acceptance criteria are as follows:

• Section III, Acceptance Criteria, of SRP Section 13.2.1, Rev ision 4, issued August 2016

• Section III, Acceptance Criteria, of SRP Section 13.2.2, Rev ision 4, issued August 2016

• Regulatory Guide (RG) 1.8, Qualification and Training of Pers onnel for Nuclear Power Plants, Revision 4, issued June 2019 (ML19101A395)

• RG 1.149, Nuclear Power Plant Simulation Facilities for Use i n Operator Training, License Examinations, and Applicant Experience Requirements, R evision 4, issued April 2011 (ML110420119)

• NUREG-0711, Revision 3

• NUREG-1021, Operator Licensing Examination Standards for Powe r Reactors, Revision 12, issued September 2021 (ML21256A276)

• NUREG-1220, Training Review Cri teria and Procedures, Revisio n 1, issued January 1993 (ML102571869)

Technical Evaluation

The COL applicant is responsible for the development of site-sp ecific training programs. This section presents an evaluation of the COL information items tha t pertain to training programs for licensed and nonlicensed plant staff.

13.2.4.1 Combined License Information Items

The NRC regulations require the COL applicant that references t he NuScale Power Plant US460 standard design to address the site-specific information described in COL information items at the COL stage.

13.2.4.1.1 Licensed and Nonlicensed Plant Staff Training Programs

SRP section 13.2.1 states that the COL applicant should provide the description and scheduling of the licensed operator training program for reactor operators and senior reactor operators, including the licensed operator requalification program. SRP se ction 13.2.2 states that the COL applicants nonlicensed plant staff training program should inc lude the initial training, periodic retraining, and qualification that are required for nonlicensed plant staff. The staff reviewed FSAR section 13.2, COL Item 13.2-1, and found that it specifies the appropriate and necessary information for licensed plant staff. The staff reviewed FSAR s ection 13.2, COL Item 13.2-2, and

13-8 found that it specifies the appropriate and necessary informati on for nonlicensed plant staff training programs. The staff also verified that the FSAR adequa tely incorporates the COL information items presented in table 13.2-1.

Combined License Information Items

Table 13.2-1 lists COL information item numbers and description s related to training from FSAR table 1.8-1.

Table 13.2-1 NuScale COL Information Items Related to FSAR, sec tion 13.2

SDAA Item No. Description Part 2 Chapter

COL Item An applicant that references the NuScale Power Plant US460 13.2 13.2-1: standard design will provide a description and schedule of the Initial Training and Qualification as well as Requalification Programs for reactor operators and senior reactor operators.

COL Item An applicant that references the NuScale Power Plant US460 13.2 13.2-2: standard design will provide a description and schedule of the Non-Licensed Plant Staff Training Programs including initial training, periodic retraining, and qualification requirements.

Conclusion

For the reasons given above, the staff concludes that the COL i nformation items specified in table 13.2-1 of this report and included in the FSAR are suffic ient to identify information that the COL applicant needs to provide to meet the applicable requireme nts of 10 CFR 19.12; 10 CFR 26.29; 10 CFR 50.34, Contents of applications: general information; 10 CFR 50.40, Common standards; 10 CFR 50.54, Conditions of licenses; 10 CFR 50.120; 10 CFR Part 50, Appendix B; 10 CFR Part 50, Appendix E; 10 CFR 52.79, Contents of applications; technical information in final safety analysis report; and 10 CFR Part 5 5.

Emergency Planning

The NRC staff conducted its review of the NuScale Standard Desi gn Approval (SDA) application (SDAA) emergency planning (EP) in accordance with the requireme nts contained in 10 CFR 52.137, Contents of applications; technical information, and 10 CFR 52.139, Standards for Review of Applications. This report section add resses those EP design features that are technically relevant to 10 CFR, Part 52, Subpart E, S tandard Design Approvals, that are not site specific and that affect some aspect of emergency planning or the capability of a licensee to cope with plant emergencies. The applicant may choo se the extent to which the SDAA includes EP features to be reviewed as part of the SDA eva luation.

The NRC staff conducted the review of standard design informati on and COL information items (designated as COL items) related to EP and documented the resu lts in this section of Chapter

13. The EP COL items are listed in Table 13.3-1, NuScale US460 SDAA Emergency Planning COL Items, of this report.

13-9 Introduction

The following subsections present the staffs technical evaluat ion of the EP design for the NuScale US460 SDAA.

Summary of Application

The subsections below summarize the EP design information submi tted in the NuScale US460 SDAA FSAR.

The NuScale US460 SDAA FSAR Section 13.3, Emergency Planning, states that the NuScale Power Plant design includes design features, facilities, and eq uipment that are usable for up to six (6) NuScale Power Modules to support emergency response fun ctions. A COL applicant that references the NuScale Power Plant US460 standard design is res ponsible for providing a comprehensive emergency plan (COL Item 13.3-2) and the descript ions of the emergency response facilities (ERFs) for management of overall licensee e mergency response (COL Item 13.3-1). Thus, the COL applicant that references the NuScale US 460 standard design is responsible for the interfaces of these features with site-spec ific parameters.

Further, the NuScale US460 FSAR Section 13.3, Emergency Planni ng, states that the NuScale US460 standard plant emergency planning design ensures that personnel are protected from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources under accident conditions (i.e., radiation dose i s limited to 5 roentgen equivalent man (rem) total effective dose equivalent (TEDE) for the durati on of the accident). In the event of a Technical Support Center (TSC) loss of ventilation, or if the TSC becomes otherwise uninhabitable, TSC personnel are evacuated and the TSC function s are transferred to a location designated by the comprehensive emergency plan that will be pro vided by the COL applicant.

Although the development of a comprehensive emergency plan is t he responsibility of the COL applicant that references the NuScale US460 standard design, th e design bases for the NuScale US460 standard plant include design features, facilitie s, functions, and equipment necessary for EP. The design bases for the NuScale US460 standa rd plant include the following EP features:

• TSC

A TSC is an onsite ERF that provides plant management and techn ical support to the plant operations personnel during emergency conditions.

• Emergency Response Data System (ERDS)

ERDS is a direct near real-time electronic data link between a licensees onsite computer system and the NRC Operations Center and provides for the automated transmission of a limited data set of selected parameters from a licensees installed onsite computer system in the event of an emergency.

• TSC Engineering Workstations

The TSC engineering workstations are for information display on ly and no control functions are provided on the TSC engineering workstations. Suf ficient backup power

13-10 sources are provided to maintain continuity of TSC functions an d to immediately resume data acquisition, storage, and display of TSC data if a loss of the primary TSC power source occurs. The plant network, as described in FSAR Section 7.2.13.7, Other Information Systems, provides data recording, trending, and hi storical retention that can be retrieved on the TSC engineering workstations.

• TSC Variables Available for Accident Assessment

The appropriate Regulatory Guide 1.97, Criteria for Accident M onitoring Instrumentation for Nuclear Power Plants, criteria that are essential for perf ormance of TSC functions are available in the TSC.

FSAR: The following NuScale US460 SDAA FSAR sections describe the de sign features with an interface to EP:

FSAR Section 1.2.2.2 Control Building, states that the NuScale US460 control building contains the TSC. The TSC layout within the control building is shown in FSAR Figure 1.2-19, Control Building 125-0 Elevation, and Figure 1.2-20, Control Buildi ng North-South Section View show that the TSC and the main control room (MCR) are located o n the same floor level elevation of the control building.

FSAR Section 3.7.2.1.2.5, Control Building, states that the c ontrol building is comprised of a seismic category I (SC-I) portion and a seismic category II (SC -II) portion and that the TSC is located in the SC-II portion of the control building.

FSAR Section 7.2.13.7, Other Information Systems, describes t he TSC engineering workstations.

FSAR Section 9.4.1.2, System Description, and FSAR Section 12.3.3.5, Normal Control Room Heating, Ventilation, and Air Conditioning System, discus s heating, ventilation, and air conditioning for the TSC.

FSAR Section 9.5.2, Communication System, provides descriptio ns for the TSC communication systems.

FSAR Section 18.7.2.4.3, Technical Support Center, Emergency O perating Facility, Waste Management Control Room, and Module Maintenance Center, provid e descriptions of the TSC workstation human-system interfaces.

ITAAC: The SDA applicant has not proposed any ITAAC related to this ar ea of review.

Technical Specifications: There are no TS for this area of review.

Technical Reports: Human Factors Engineering Program Management Plan, TR-130414-NP, Revision 0, dated December 2022, states in Section 2.2.3, Appl icable Facilities, that the NuScale Human Factors Engineering Program scope, applicable fro m the start of conceptual design through turnover to the licensee, is limited to defining the TSC plant data and the TSCs human-system interfaces impact on licensed operator workload.

13-11 Regulatory Basis

The following NRC regulations cont ain the relevant requirements for this review:

• 10 CFR Part 50, Appendix E, as it relates to EP and the ERDS

• 10 CFR 52.137(a)(8), which requires that an SDAA FSAR must inc lude the information necessary to demonstrate complian ce with any technically releva nt portions of the TMI requirements in 10 CFR 50.34(f), except 10 CFR 50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v)

• 10 CFR 52.137(a)(11), which requires that an SDAA FSAR must in clude the information pertaining to design features that affect plans for coping with emergencies in the operation of the reactor facility or a major portion of the fac ility

• 10 CFR 52.139, as it relates to EP information submitted under 10 CFR Part 52, Subpart E, Standard Design Approvals, for an SDAA

The following guidance documents provide criteria relevant to t his review and were used to confirm that the above requir ements have been adequately addres sed:

• NUREG-0800 (SRP), Section 13.3, Emergency Planning, lists th e acceptance criteria that are adequate to meet the above requirements and the review interfaces with other SRP sections.

• NUREG-0696, Functional Criteria for Emergency Response Facili ties, issued February 1981 (ML051390358), describes the facilities and syste ms that nuclear power plant licensees will use to improve responses to emergencies.

• NUREG-0737, Clarification of TMI Action Plan Requirements, S upplement 1, Requirements for Emergency Response Capability, issued Januar y 1983 (ML051390367), describes post-TMI requirements for emergency re sponse capabilities that have been approved for implementation.

• NUREG-1394, Emergency Response Data System, Revision 2, Augu st 2022 (ML22244A081), provides guidance for the implementation and con tinued operation of ERDS at licensee sites.

• The NRC Office of Nuclear Security and Incident Response (NSIR ), Division of Preparedness and Response (DPR), interim staff guidance (ISG) d ocument, NSIR/DPR-ISG-01, Interim Staff GuidanceEmergency Planning for Nuclear Power Plants, issued November 2011, provides updated guidance for ad dressing EP requirements for nuclear power plants based on changes to emerg ency preparedness regulations, in part, to 10 CFR 50.47, Emergency plans, and 1 0 CFR Part 50, Appendix E, that the NRC published on November 23, 2011, in the Federal Register (76 FR 72560).

Technical Evaluation

The design information required to license and operate a site-s pecific NuScale US460 nuclear power plant is identified throughout the NuScale US460 FSAR as COL information items. These

13-12 COL information items are the responsibility of a COL applicant that desires to construct and obtain a license to operate a NuScale US460 standard design pow er plant and are to be provided in the COL application. FSAR section 13.3 and table 13.3-1 of this report identifies the NuScale US460 FSAR EP COL information items that are described or provided in the NuScale SDAA FSAR and are evaluated by the staff in this report. COL It em 13.3-2 states that an applicant that references the NuScale Power Plant US460 standar d design is responsible for providing a comprehensive emergency plan. Therefore, the follow ing subsections document the staffs review and evaluation of the EP design features informa tion contained in the NuScale US460 SDAA for conformance with the applicable standard design acceptance criteria as identified in SRP section 13.3 to demonstrate compliance with t he listed EP requirements.

13.3.4.1 Emergency Response Facilities

The guidance of SRP section 13.3,Section II, SRP Acceptance C riteria, item 25, states, in part, that an applicant subject to the 10 CFR 50.34(f)(2)(xxv) criterion shall provide a technical support center (TSC) and an operational support center (OSC). N uScale FSAR Table 1.9-3, Conformance with NUREG-0800, Standard Review Plan and Design S pecific Review Standard, and Table 1.9-5, Conformance with Three Mile Island Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933), state that the NuSc ale US460 SDAA only partially conforms to 10 CFR 50.34(f)(2)(xxv). Specifically, th e NuScale US460 SDAA design includes a TSC but does not include an OSC. The design and desc riptions of emergency response facilities are the responsibility of an applicant that references the NuScale US460 standard design in a COL application. This is stated in COL Ite m 13.3-1.

The staff used the guidance in NUREG-0696, Supplement 1 to NURE G-0737, and NUREG-0800 specific acceptance criteria to evaluate the TSC des ign for compliance with the relevant standard design emergency response facilities requirem ents.

13.3.4.1.1 Technical Support Center

FSAR Section 13.3, Emergency Planning, states that a TSC is p rovided that is compliant with the design requirements of NUREG-0696 functional criteria for E RFs.

13.3.4.1.1.1 Technical Support Center Structure

FSAR Section 1.2.2.2 Control Building, states that the NuScale US460 control building contains the TSC. The TSC layout within the control building is shown in FSAR Figure 1.2-19, Control Building 125-0 Elevation, and Figure 1.2-20, Control Buildi ng North-South Section View.

FSAR Section 3.7.2.1.2.5, Control Building, states that the c ontrol building is comprised of a seismic category I (SC-I) portion and a seismic category II (SC -II) portion and that the TSC is located in the SC-II portion of the control building. In respon se to RAI 10097, Question 13.3-1 (ML23352A361), the applicant states that the SC-II structures, systems, and components are analyzed for safe shutdown earthquake loads and the same wind p ressure as the SC-I portion, that the NuScale US460 standard design approval flood level is below the 125 ft elevation of the TSC, and that the control building that the TSC is located will be constructed to a building code that is endorsed by the International Building Code that replac ed the Uniform Building Code. In that response, the applicant concluded that the TSC is a well-e ngineered structure designed to be capable of withstanding earthquakes, high winds, and floods.

The NRC staff reviewed the above NuScale US460 SDAA TSC structu re design descriptions, and finds that the NuScale US460 TSC design is consistent with the TSC structure design

13-13 guidance in Section 2.5, Structure, of NUREG-0696 and NUREG-0 737, Supplement 1, Section 8.2.1.d.

13.3.4.1.1.2 Technical Support Center Size

The TSC is sized to accommodate staffing levels of 25 persons t hat consist of 20 licensee personnel and 5 NRC personnel at 75 square feet per person. The TSC includes a technical evaluation room and additional space for storage, offices, and conference rooms. The staff finds that this NuScale US460 TSC size design information is consiste nt with the specific space and personnel accommodation guidance contained in NUREG-0696, Secti on 2.4 Size and NUREG-0737, Supplement 1, Section 8.2.1.c.

13.3.4.1.1.3 Technical Support Center Location

FSAR Figure 1.2-19 and Figure 1.2-20, show that the TSC and the main control room (MCR) are located on the same floor level elevation of the control bu ilding. When using the shortest designed direct route, the walking time between the entrance of the MCR and the entrance of the TSC does not exceed two minutes. Based on this TSC location design information, the staff finds that the NuScale US460 TSC location design addresses the applicable standard design approval TSC location guidance in Section 2.2, Location, of N UREG-0696 and Section 8.2.1.b of NUREG-0737, Supplement 1.

13.3.4.1.1.4 Technical Support Center Communications

The TSC is equipped with voice communications systems that prov ide communications between the TSC and plant, local off-site emergency response fa cilities, the NRC, and State operations centers. FSAR Section 9.5.2.3, Safety Evaluation, states that the TSC voice communications consist of the telephony system, wide area mass notification system (WAMNS), and the plant radio system. FSAR Section 9.5.2.2, Sys tem Description, provides descriptions for these TSC communication systems:

• The telephony service includes onsite communication and an int erface with an offsite public switched telephone network and has the necessary bandwid th to support peak traffic for normal and emergency plant operations modes.

• The WAMNS sends emergency alarms and communications to plant p ersonnel by broadcast paging. The WAMNS has the capability to support opera tions during normal and emergency conditions.

• The distributed antenna system distributes frequencies for the plant radio system in buildings and outdoors across the site, as needed. The distribu ted antenna system interfaces with the telephony system to allow access to both on site telephony and the external telecommunications network. The plant radio has the ca pability to support operations during normal and emergency conditions.

These TSC communication systems provide communications between the TSC and plant, offsite emergency response facilities, the NRC, and local and s tate operations centers and are physically independent of each other such that a failure in one system will not disable another system. COL Item 9.5-1 specifies that it will be the responsibi lity of a COL applicant that

13-14 references the NuScale US460 standard design approval to provid e a description of the offsite communication system, how the offsite communication system inte rfaces with the onsite communications system, and how continuous communications capabi lity is maintained to ensure effective command and control with onsite and offsite re sources during both normal and emergency situations.

Section 9.5.2 of this report documents the staffs evaluation a nd safety findings for these TSC communication systems. Based on its review of the above NuScale US460 TSC communication system design information, the staff finds that the TSC communi cations design is consistent with the TSC communication guidance in Section 2.7, Communicat ions, of NUREG-0696 and NUREG-0737, Supplement 1, Section 8.2.1.g.

13.3.4.1.1.5 Technical Support Center Technical Data, Data Syst em, and Human Factors Engineering

FSAR Section 7.2.13.7, Other Information Systems, states that the plant monitoring data is retrieved in the TSC from the plant network via the TSC enginee ring workstations and displayed in the TSC. The plant network provides plant operation and vari able data recording, trending, and historical retention that can be retrieved by the TSC engin eering workstations. The plant network is shown in Figure 7.0-1, Overall Instrumentation and Controls System Architecture Diagram.

The guidance of SRP 13.3, states that for an applicant subject to 10 CFR 50.34(f)(2)(iv), a review is to be performed to assure that the safety parameter d isplay system information capabilities are available in the TSC. The NuScale US460 safety display and indication system is integrated into the MCR Human System Interface (HSI) design. FSAR Section 18.7.2.4.3, states that the HSI design in the TSC is a derivative of the MC R HSI and complies with the NuScale HSI Style Guide. FSAR Section 18.7.2.3.1, Concept of U se, states that the HSIs facilitate the operators' abilities to perform their plant acti vities (e.g., responding to off-normal conditions, performing emergency response duties such as off-si te notifications) by providing the controls, indications, alarms, and procedures necessary for the operators to carry out their responsibilities. The HSI design element of the Human Factors E ngineering (HFE) Program provides the design of interfaces between plant personnel and p lant systems and components.

FSAR Section 18.7.1, Objectives and Scope, states that the ob jective of the HSI design element is to translate the functional requirements analysis, f unction allocation requirements, as well as task analysis requirements, into HSI design requirement s and the detailed design of alarms, indications, controls, and other aspects of the HSIs an d that this objective is accomplished by systematically applying HFE principles and crit eria. The TSC HSIs are for information display only and no control functions are provided in the TSC. The evaluation of the HFE aspects of the NuScale US460 HSI design is provided in Chap ter 18 of this report.

FSAR Section 13.3 states that FSAR Table 7.1-7, Summary of Pos t-accident Monitoring Variables, list post-accident monitoring (PAM) variables avail able in the TSC and the MCR.

FSAR Section 7.1.1.2.2, Post-Accident Monitoring, states that PAM Type B, C, D, E, and F variables are summarized in Table 7.1-7 and that the NuScale US 460 reactor design has no Type A variables. In addition, the selection of each type of va riable follows the guidance provided in Institute of Electrical and Electronics Engineers ( IEEE) Std 497-2016, "IEEE Standard Criteria for Accident Monitoring Instrumentation for N uclear Power Generating Stations, and the guidance of Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Revision 5.

13-15 Based on its review of the above TSC HSI design and TSC PAM var iable descriptions listed above, the staff finds that the NuScale US460 TSC data set disp lay and data system design is consistent with the TSC technical data system guidance in Secti on 2.9, Technical Data and Data System, of NUREG-0696 and Supplement 1 to NUREG-0737, Sec tion 8.2.1.h and Section 8.2.1.k, and SRP 13.3. Therefore, the staff finds that the MCR HSI information capabilities are available in the TSC and that the HSI TSC desi gn meets the applicable TSC requirements of 10 CFR 50.34(f)(2)(iv).

13.3.4.1.1.6 Technical Support Center Instrumentation, Data Sys tem Equipment, and Power Supplies

The plant network receives plant data from one-way deterministi c isolation devices that provide one-way communication interfaces to the plant network as shown in FSAR Figure 7.0-1. The TSC engineering workstations are separate from the MCR operator workstations, which are located in the MCR. The TSC engineering workstations have licen sed operating systems, configuration software, and a software package for complete con figuration, trending, and diagnostics. TSC workstation operation is independent of action s in the MCR and do not degrade or interfere with MCR functions or safety-related plant functions.

FSAR Section 13.3 states that the TSC engineering workstations and communications equipment are supplied with power from the normal DC power syst em (EDNS). Section 13.3 of the FSAR further states that the EDNS includes battery backup p ower and may also be provided backup power from the backup power supply system. In response to RAI 10097, Question 13.3.2 (ML23352A361), the SDA applicant states that th ere are sufficient backup power supply sources available to maintain reliability and cont inuity of the TSC data systems and the TSC functions. If a loss to the primary TSC power sourc e occurs, the TSC backup power sources immediately resume data acquisition, storage, and display of TSC data.

Based on its review of the above NuScale US460 TSC data system backup power sources provided to maintain continuity of TSC functions and TSC operat ional independence from the MCR workstation control functions, the staff finds that the NuS cale US460 TSC data system design is consistent with the TSC data system equipment guidanc e in Section 2.8, Instrumentation, Data System Equipment, and Power Supplies of NUREG-0696.

13.3.4.1.1.7 Technical Support Center Habitability

The TSC is designed to protect TSC personnel from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources unde r accident conditions (i.e.,

maximum of 5 rem TEDE for the duration of the accident). As sta ted in FSAR Section 15.0.3.6.4, Radiological Exposures from Shine, the potential radiological exposures following radiological release events were evaluated and the 30-day cumul ative TSC doses, considering the combination of the beyond-design-basis core damage event-ba sed shine dose and the event-specific doses calculat ed, remained below the limit of 5 rem TEDE for the duration of the accident. FSAR Section 9.4.1.2, System Description, and FSAR Section 12.3.3.5, Normal Control Room Heating, Ventilation, and Air Conditioning System, state that during normal operations, the normal control room heating, ventilation, and t he air conditioning (HVAC) system (CRVS) supplies conditioned air to the control building, includ ing the TSC, with outside air that is filtered to maintain a suitable environment for personnel an d equipment. The CRVS maintains temperature and humidity control within ranges suitable for the comfort of personnel and to prevent degradation of equipment. The CRVS has two 100 percent capacity air handling units designed such that if the current operating air handling unit f ails, the standby air handling unit

13-16 will start automatically. The CRVS is designed to maintain a po sitive pressure inside the control building with respect to adjacent spaces. If a high radiation i ndication is received from an outside air intake radiation monitor, the supply air is routed through the CRVS filter unit which provides high-efficiency particulate air (HEPA) and charcoal fi ltration. Section 9.4 of this report provides the staffs evaluation of the CRVS. The staffs evalua tion of the radiological habitability of the TSC is provided in Section 6.4 and Chapter 15 of this re port.

In the event of a loss of TSC ventilation, or if the TSC become s uninhabitable, TSC personnel will be evacuated and the TSC functions will be transferred to a location designated by the COL applicants emergency plan (COL Item 13.3-2). Based on its revi ew of the above NuScale US460 SDAA accident event analysis results demonstrating protec tion of TSC personnel from radiological release hazards under accident conditions, the sta ff finds that the NuScale US460 TSC personnel protection from radiological hazards under accide nt conditions design is consistent with the TSC personnel radiological protection guida nce in Section 2.6, Habitability, of NUREG-0696 and NUREG-0737, Supplement 1, Section 8.2.1.f.

The TSC is designed with fixed area radiation monitors (ARM). F SAR Section 12.3.4.2, Fixed Area Radiation Monitoring Instrumentation, states that the TSC ARM alarm setpoints are established to alert plant personnel when radioactivity in a sp ecific location reaches levels that have been determined to be abnormal. The ARMs provide both indi cation and alarm functions to the TSC. The fixed ARMs and associated instrument and contro ls platforms provide information that can supplement radiological surveys, meet repo rting requirements, and inform workers of radiological conditions. In addition, ARMs system da ta is capable of being supplied to the NRC Operations Center through the connection to the ERDS. The ARM indication provides awareness to operators and workers of changing radiolo gical conditions. The TSC ARMs, including the type of radiation monitored and the associa ted principal isotope(s) and instrument ranges were reviewed by staff and are identified in FSAR Table 12.3-8, Fixed Area and Airborne Radiation Monitors Post-Accident Monitoring Variab les, and Table 12.3-10, Fixed Area Radiation Monitors. Based on its review of the NuScale US 460 TSC radiation monitoring design descriptions listed above, the staff finds that the NuSc ale US460 TSC radiation monitoring design is consistent with the TSC radiation monitori ng system guidance in Section 2.6, Habitability, of NUREG-0696.

13.3.4.1.1.8 Technical Support Center Evaluation Conclusion

The staff notes that EP specific guidance applicable to the TSC that is not addressed in this reports EP evaluation section will be addressed by the COL app licant that references the NuScale US460 standard design approval and provides the site-sp ecific TSC design information and emergency plan. Thus, based on the staffs evaluation of th e NuScale SDAA TSC design information, the staff concludes that for the matters reviewed in this TSC evaluation report section, the TSC EP design information provided in the NuScale US460 SDAA is consistent with the applicable TSC emergency planning guidance identified in NU REG-0696, Supplement 1 to NUREG-0737, and SRP Section 13.3. Therefore, the staff finds th at NuScale US460 SDAA TSC EP design information meets the applicable TSC general desc ription requirements for ERFs of 10 CFR 50.34(f)(2)(xxv).

13.3.4.2 Emergency Response Data System

SRP Section 13.3,Section II, SRP Acceptance Criteria, Item 1 2, states that Section VI, Emergency Response Data System, of Appendix E to 10 CFR Part 50, requires an ERDS. The ERDS guidance of NUREG-1394, Office of Nuclear Security and In cident Response

13-17 Emergency Response Data System, Section 2.1, ERDS Design Conc ept, states that the ERDS concept involves the direct electronic transmission of sel ected parameters from the electronic data systems that are currently installed at the lic ensee facility. FSAR Section 7.2.13.7, Other Information Systems, states that there is a l ink from the NuScale US460 plant network to the NRC ERDS through dedicated communication servers that connect to the plant network and provide data communication of required plant data. The NuScale US460 ERDS link provides a direct near-real-time electronic data link of select ed parameters between the on-site computer system and the NRC Operations Center in the event of a n emergency. The ERDS will be designed to be compliant with Section VI of 10 CFR Part 50, Appendix E. FSAR Table 1.9-3 states that the ERDS site-specific aspects are the responsibili ty of the applicant (COL Item 13.3-2).

Based on the ERDS design information above, the staff concludes that the NuScale US460 SDAA has provided adequate descriptions of the ERDS electronic data link design between the NuScale US460 facility and the NRC Operations Center and addres ses the ERDS acceptance criteria guidance of SRP 13.3,Section II, Item 12. Therefore, staff finds that the NuScale US460 SDAA ERDS design information meets the applicable ERDS standard design requirements in Section VI, Emergency Response Data System, of Appendix E to 10 CFR Part 50.

13.3.4.3 Decontamination Facilities

SRP Section 13.3,Section III, REVIEW PROCEDURES, review Sect ion Standard Design Certification, Items 3 and 4, states that the reviewer should examine the relevant sections of the FSAR that address decontamination facilities and systems th at support the emergency preparedness and response capabilities of the proposed reactor design. FSAR Section 12.1.2.3, Facility Layout General Design Considerations for Maintaining Radiation Exposures as Low as Reasonably Achievable, states that the radiation protection su pport facilities are located in the Radioactive Waste Building and include change rooms, counting r oom, and personnel decontamination facilities, and also serves as the access porta l to the radiologically controlled area and includes dosimetry issue and personnel contamination m onitors. In addition, FSAR Section 12.3.6.1.3, Design Considerations for Reduction of Cro ss-Contamination, Decontamination and Waste Generation - Objective 3, states tha t the NuScale US460 standard design contains an on-site decontamination facility to reduce t he potential for cross-contamination, the need for decontamination, and radioactive wa ste generation.

Decontamination facilities are provided to remove or reduce rad ioactive contaminants from plant equipment, protective clothing, and personnel.

Based on its review of the decontamination emergency preparedne ss capabilities design descriptions above, the staff concludes that the NuScale US460 SDAA has provided adequate descriptions for the standard designs emergency preparedness d econtamination facilities and address the applicable decontamination guidance of SRP 13.3, Se ction III, review Section Standard Design Certification, Items 3 and 4.

13.3.4.4 Post-accident Sampling System

SRP Section 13.3,Section II, Item 27, states that 10 CFR 50.34 (f)(2)(viii) requires that an applicant provide a capability to promptly obtain and analyze s amples from the reactor coolant system and containment that may contain accident source term ra dioactive materials, while ensuring that no individual receives radiation exposure in exce ss of 0.05 Sv (5 rem) to the whole body or 0.5 Sv (50 rem) to the extremities. FSAR Table 1.9-5 st ates that the NuScale US460 standard design does not rely on primary coolant or containment samples to assess the extent

13-18 of potential core damage and that the NuScale US460 standard de sign supports an exemption from the 10 CFR 50.34(f)(2)(viii) requirement. The exemption fr om the design criterion of CFR 50.34(f)(2)(viii) is evaluated in Section 9.3.2 of this evaluat ion report.

13.3.4.5 Containment Monitoring and Continuous Sampling of Potential Accident Release Points

SRP Section 13.3,Section II, Item 28, states that 10 CFR 50.34 (f)(2)(xvii) requires instrumentation to measure, record, and readout in the control room various containment parameters, including noble gas effluents at all potential acci dent release points. In addition, an applicant must provide for continuous sampling of radioactive i odine and particulates in gaseous effluents from all potential accident release points, and for o nsite capability to analyze and measure these samples. FSAR Table 1.9-5 and FSAR Section 9.3.2. 2.3, System Operation, state that the NuScale US460 standard design supports an exempt ion from 10 CFR 50.34(f)(2)(xvii)(c). This exemption from the requirement of 50.34(f)(2)(xvii)(c) is evaluated in Sections 6.2.5 and 9.3.2 of this report. The evaluation of the sampling system instrumentation is provided in Section 7.2.13 of this report. In addition, the eva luation for continuous sampling of radioactive iodine and particulates in gaseous effluents from a ll potential accident release points is conducted in Section 11.5.2 and Section 12.3.4 of this repor t.

13.3.4.6 Inspections, Tests, Analyses, and Acceptance Criteria

FSAR Table 1.9-3 states that NuScale US460 EP ITAAC were not in cluded as part of the SDAA. COL Item 14.3-1 specifies that an applicant that referenc es the NuScale US460 standard design will provide the site-specific selection method ology and ITAAC for emergency planning.

13.3.4.7 Combined License Information Items

Table 13.3-1, NuScale US460 SDAA Emergency Planning COL Items, of this report provide the NuScale US460 SDAA COL design information items related to EP. The NRC staff reviewed the COL items and found them to be consistent with the regulato ry standards in 10 CFR Part 52.139 and with the applicable standard design SRP guidance. Th erefore, the staff finds that the proposed COL items are sufficient in identifying informatio n a COL applicant needs to provide to address the applicable EP requirements.

Table 13.3-1 NuScale US460 SDAA Emergency Planning COL Items

COL Item No. Description FSAR Section An applicant that references the NuScale Power Plant US460 standard design will provide a description of the offsite communication system, how that system interfaces with the COL Item 9.5-1 onsite communications system, as well as how continuous 9.5.2.1 communications capability is maintained to ensure effective command and control with onsite and offsite resources during both normal and emergency situations.

An applicant that references the NuScale Power Plant US460 COL Item 13.3-1 standard design will provide a description of the Emergency 13.3 Response facilities for management of overall licensee

13-19 Emergency Response. The facility will meet the requirements of 10 CFR 52.79.

An applicant that references the NuScale Power Plant US460 COL Item 13.3-2 standard design will provide a comprehensive Emergency 13.3 Plan in accordance with 10 CFR 50 and 10 CFR 52.79(a)(21).

An applicant that references the NuScale Power Plant US460 COL Item 14.3-1 standard design will provide the site-specific selection 14.3.1 methodology and inspections, tests, analyses, and acceptance criteria (ITAAC) for emergency planning.

Conclusion

The staff concludes, on the basis of its review as described ab ove, that the applicant has adequately addressed the EP standard design related guidance de scribed in section 13.3.3 of this report for the NuScale US460 standard design approval. The refore, staff finds that the NuScale US460 SDAA EP design information meets the SDA requirem ents of 10 CFR 52.137(a)(8) and 10 CFR 52.137(a)(11).

Operational Programs

Introduction

A COL applicant is required by 10 CFR 52.79 to describe operati onal programs, but similar requirements do not exist for SDAAs. NuScale provided a COL ite m describing a future COL applicants obligation to provide operational program informati on. The staff evaluated this section using draft Revision 4 of SRP Section 13.4, Operationa l Programs, issued September 2018.

Summary of Application

In FSAR Section 13.4, Operational Programs, the applicant pro vided COL Item 13.4-1, which states that a COL applicant that references the NuScale power p lant standard design approval will provide site-specific information, including an implementa tion schedule, for the listed operational programs.

FSAR: FSAR section 13.4 provides the applicants COL information item on operational programs.

ITAAC: No ITAAC are associated with the operational programs.

Technical Specifications: No TS are associated with the operational programs.

Technical Reports: No TRs are associated with the operational programs.

Regulatory Basis

There are no regulatory requirements for operational programs f or an SDAA. An SDAA is required to include a quality assurance program meeting the cri teria of 10 CFR Part 50, Appendix B. SER chapter 17 describes how the applicant meets th at requirement. Similarly, SER section 13.6 describes how the applicant meets the informat ion security requirements of 10 CFR Part 73, Physical Protection of Plants and Materials.

13-20 Technical Evaluation

The staff compared the list of operational programs in COL Item 13.4-1 with the recommended list in SRP section 13.4. The staff finds that the applicants list includes all of the applicable programs recommended by the SRP.

Combined License Information Items

Table 13.4-1 lists a COL information item related to operationa l programs from FSAR table 1.8-1.

Table 13.4-1 NuScale COL Information Items Related to FSAR Sect ion 13.4

Item No. Description FSAR Section COL Item An applicant that references the NuScale Power Plant US460 13.4 13.4-1 standard design will provide site-specific information, includi ng implementation milestones, for Operational Programs:

• Inservice Inspection Programs (Section 5.2, Section 5.4, and Section 6.6)

• Inservice Testing Programs (Section 3.9 and Section 5.2)

• Environmental Qualification Program (Section 3.11)

• Preservice Inspection Program (Section 5.2 and Section 5.4)

• Preservice Testing Program (Section 3.9.6 and Section 5.2)

• Containment Leakage Rate Testing Program (Section 6.2)

• Fire Protection Program (Section 9.5.1)

• Process and Effluent Monitoring and Sampling Program (Section 11.5)

• Radiation Protection Program (Section 12.5)

• Non-Licensed Plant Staff Training Program (Section 13.2)

• Reactor Operator Training Program (Section 13.2)

• Reactor Operator Requalification Program (Section 13.2)

• Emergency Planning (Section 13.3)

• Process Control Program (Section 11.4)

• Security (Section 13.6)

13-21 Item No. Description FSAR Section

• Quality Assurance Program (Section 17.5)

• Maintenance Rule (Section 17.6)

• Initial Test Program (Section 14.2)

Conclusion

The staff determined that the COL item listed above is acceptab le because the SDAA appropriately directs the COL applicant to develop operational programs, consistent with the list provided and as applicable in SRP section 13.4, draft Revision 4.

Plant Procedures

Introduction

A COL holders plant procedures include (1) administrative proc edures that provide for administrative control over safety-related activities for the o peration of the facility, (2) operating procedures and emergency operating procedures (EOPs) used to en sure that routine operating, off-normal (i.e., abnormal), and emergency activities are condu cted safely, and (3) procedures for other safety-related plant operating activities, including related maintenance activities, that the operating program or EOP program does not cover.

The staff reviewed the SDAA to evaluate the COL information ite ms for plant procedures. The staff will review the technical content of the generic guidance used to develop plant-specific technical guidelines when a COL applicant submits a procedure g eneration package.

Summary of Application

FSAR: Procedure development is not within the scope of the NuScale US 460 SDAA. This responsibility resides with the COL applicant. FSAR Section 13. 5, Plant Procedures, specifies COL information items directing the COL applicant to describe t he administrative, operating, and maintenance procedures.

ITAAC: The applicant has not proposed any ITAAC related to this area of review.

Technical Specifications: There are no TS for this area of review.

Technical Reports: There are no TRs for this area of review.

Regulatory Basis

SRP Section 13.5.1.1, Administrative ProceduresGeneral, and SRP Section 13.5.2.1, Operating and Emergency Operating Procedures, identify, in pa rt, the relevant NRC regulatory requirements for plant procedures and the associated acceptance criteria.

The following regulatory requirements are applicable for plant procedures:

13-22

• 10 CFR 50.34(f)(2)(ii), as it pertains to the TMI-related requ irement for applicants to establish a program to begin during construction and to follow into operation for assessing and improving plant emergency procedures

• 10 CFR 50.34(f)(3)(i), as it pertains to the TMI-related requi rement to provide administrative procedures that evaluate and provide feedback on operating, design, and construction experience

• 10 CFR 50.40(a), as it pertains to the issuance of a COL under 10 CFR Part 52 based on considerations of whether the applicant has developed operat ing procedures that are sufficient to provide reasonable assurance that the nuclear pow er plant can be safely operated

• 10 CFR Part 50, Appendix B, as it pertains to the establishmen t of criteria for the development, approval, and control of procedures for all activi ties affecting quality

• 10 CFR 52.79(a)(27), (29)(i), and (29)(ii), as they pertain to information that must be included in the FSAR submitted as part of the application for a COL, specifically, (1) the managerial and administrative controls associated with procedur es used to perform activities that affect the quality of SSCs covered under the qu ality assurance program, as established in 10 CFR Part 50, Appendix B, and (2) plans for the development and implementation of plant procedures used for emergency operation s (other than EP) and the conduct of normal operations, including maintenance, survei llance, and periodic testing of SSCs

The related acceptance criteria are as follows:

• RG 1.33, Quality Assurance Program Requirements (Operation), Revision 3, issued June 2013 (ML13109A458)

• Appendix A, Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors, to American National Standards Institute (ANSI)/Amer ican Nuclear Society (ANS) 3.2-2012, Managerial, Administrative, and Quality Assura nce Controls for Operational Phase of Nuclear Power Plants

• Section III, Acceptance Criteria, of SRP Section 13.5.1.1, R evision 2, issued August 2016

• SRP Section 13.5.2.1, Revision 3, issued August 2014

• Section I.C.1, Guidance for the Evaluation and Development of Procedures for Transients and Accidents, of NUREG-0737 (ML051400209)

• Section 7, Upgrade Emergency Operating Procedures, of Supple ment 1 to NUREG-0737 (ML102560009)

• NUREG-0899, Guidelines for the Preparation of Emergency Opera ting Procedures Resolution of Comments on NURE G-0799, issued August 1982 (ML102560007)

13-23 Technical Evaluation

FSAR section 13.5 identifies procedure development as the COL a pplicants responsibility. This section evaluates the adequacy of the COL information items for plant procedures.

13.5.4.1 Combined License Information Items

13.5.4.1.1 Administrative Procedures

SRP section 13.5.1.1 describes administrative procedures as tho se that provide for administrative control over safety-related activities for the o peration of the facility. The staffs review of the NuScale US460 used SRP section 13.5.1.1 and focus ed on the evaluation of COL information items pertaining to administrative procedures. COL Item 13.5-1 in FSAR Section 13.5.1, Administrative Procedures, directs the COL ap plicant to describe site-specific procedures that provide administrative control for activities t hat are important for the safe operation of the facility consistent with the guidance in RG 1. 33, Revision 3, which endorses ANSI/ANS 3.2-2012. Accordingly, the staff determined that COL I tem 13.5-1 identifies information on administrative procedures that the COL applicant needs to provide.

SRP section 13.5.1.1 provides the technical rationale for apply ing SRP acceptance criteria to the establishment of a program for the development and implemen tation of administrative procedures. FSAR section 13.5.1, COL Item 13.5-2, directs the C OL applicant to provide a plan for the development, implementation, and control of administrat ive procedures, including preliminary schedules for preparation and target completion dat es. Additionally, the COL applicant will identify the group within the operating organiza tion responsible for maintaining these procedures. The staff determined that COL Item 13.5-2 is consistent with provisions in SRP section 13.5.1.1.

13.5.4.1.2 Operating and Maintenance Procedures

SRP section 13.5.2.1 states that the applicants SAR should des cribe the different classifications of procedures that the operators will use in th e control room and locally for operations in the plant. FSAR Section 13.5.2.1, Operating and Emergency Operating Procedures, COL Item 13.5-3, directs the COL applicant to desc ribe the site-specific procedures that operators use in the MCR and locally in the pla nt, including normal operating procedures, abnormal operating procedures, and EOPs. The COL ap plicant will also describe the classification system for these procedures and the general format and content of the different classifications. The staff determined that COL Item 1 3.5-3 appropriately directs the COL applicant to describe the different classifications of the site-specific procedures that licensed operators and nonlicensed operators perform.

SRP section 13.5.2.1 provides the technical rationale for apply ing SRP acceptance criteria to the establishment of programs for the development and implement ation of operating and maintenance procedures. Thus, an applicant should consider incl uding COL information items that direct the COL applicant to provide programs for developme nt and implementation of the operating and maintenance procedures. FSAR Section 13.5.2, Ope rating and Maintenance Procedures, COL Item 13.5-4, directs a COL applicant to provid e a plan for the development, implementation, and control of operating procedures, including preliminary schedules for preparation and target completion dates. Additionally, the COL applicant will identify the group within the operating organization responsible for maintaining t hese procedures.

13-24 COL Item 13.5-5 directs a COL applicant that references the sta ndard design approval for NuScale US640 to provide a plan for the development, implementa tion, and control of EOPs, including preliminary schedules for preparation and target comp letion dates. Additionally, the COL applicant will identify the group within the operating orga nization responsible for maintaining these procedures.

For the reasons stated above, the staff concludes that COL Item s 13.5-4 and 13.5-5 appropriately state that the COL applicant will provide program s for the development, implementation, and control of operating and maintenance proced ures.

FSAR section 13.5.2.1, COL Item 13.5-6, directs an applicant re ferencing the standard design for NuScale US460 to describe site-specific maintenance and oth er operating procedures. It also requires COL applicants to describe how these procedures a re classified, including the general format and content of the different classifications. Th is COL information item contains a list of the categories of procedures to be included.

COL Item 13.5-7 directs a COL applicant to provide a plan for t he development, implementation, and control of maintenance and other operating procedures, incl uding preliminary schedules for preparation and target completion dates. Additionally, the COL applicant will identify the group or groups within the operating organization that will be respon sible for maintaining and following these procedures.

For the reasons stated above, the staff determined that COL Ite ms 13.5-6 and 13.5-7 appropriately direct the COL applicant to describe the differen t classifications of procedures for developing maintenance and other operating procedures (i.e., pr ocedures for activities not covered under the operating procedures or EOPs identified in se ction I.1 of SRP section 13.5.2.1).

Combined License Information Items

Table 13.5-1 lists COL information item numbers and description s related to plant procedures from FSAR table 1.8-1.

Table 13.5-1 NuScale COL Information Items Related to FSAR Sec tion 13.5

Item No. Description FSAR Section

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-1: standard design will describe the site-specific procedures that provide administrative control for activities that are importan t for the safe operation of the facility consistent with the guidance provided in Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 3.

13-25 Item No. Description FSAR Section

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-2: standard design will provide a plan for the development, implementation, and control of administrative procedures, including preliminary schedules for preparation and target date s for completion. Additionally, the applicant will identify the g roup within the operating organization responsible for maintaining these procedures.

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-3: standard design will describe the process to manage the development, review, and approval of the site-specific procedures that operators use in the main control room and locally in the plant, including normal operating procedures, abnormal operating procedures, and emergency operating procedures. The applicant will describe the classification syst em for these procedures, and the general format and content of the different classifications.

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-4: standard design will provide a plan for the development, implementation, and control of operating procedures, including preliminary schedules for preparation and target dates for completion. Additionally, the applicant will identify the group within the operating organization responsible for maintaining these procedures.

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-5: standard design will provide a plan for the development, implementation, and control of emergency operating procedures, including preliminary schedules for preparation and target date s for completion. Additionally, the applicant will identify the g roup within the operating organization responsible for maintaining these procedures.

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-6: standard design will describe the site-specific maintenance and other operating procedures, including how these procedures are classified, and the general format and content of the different classifications. The categories of procedures listed below will be included:

• plant radiation protection procedures

• emergency preparedness procedures

• calibration and test procedures

• chemical-radiochemical control procedures

• radioactive waste management procedures

• maintenance and modification procedures

• material control procedures

• plant security procedures

13-26 Item No. Description FSAR Section

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-7: standard design will provide a plan for the development, implementation, and control of maintenance and other operating procedures, including preliminary schedules for preparation and target dates for completion. Additionally, the applicant will i dentify what group or groups within the operating organization have the responsibility for maintaining and following these procedures.

Conclusion

The COL applicant is responsible for the development of plant p rocedures. In its review of the FSAR for US460, section 13.5, the staff evaluated seven COL inf ormation items. The staff determined that the seven items are sufficient to identify info rmation the COL applicant needs to provide to address the applicable requirements for plant proced ures.

The staff concludes that the COL information items specified in table 13.5-1 are sufficient to identify information that the COL applicant needs to provide to address the applicable requirements of 10 CFR 50.34, 10 CFR 50.40, Appendix B to 10 CF R Part 50, and 10 CFR 52.79.

Physical Security

Introduction

This chapter of the SER documents the NRC staffs review of sec tion 13.6 of the FSAR, Revision 1. The FSAR and the referenced Technical Report (TR) 1 18318, NuScale Design of Physical Security Systems, Revision 1, dated August 2023, desc ribe the physical security systems, hardware, and features (PSS) that are within the scope of the standard design approval. The PSS is relied on to implement security response f unctions (i.e., detection, assessment, communications, security responsedelays, interdict ions, and neutralization).

Specifically, the SDAA provides the design descriptions for eng ineered PSS and credited design features (e.g., structural walls, floors, ceilings, and configu rations of the nuclear island and structures), descriptions of intended security functions and pe rformance requirements, design bases for the detailed design, and supporting technical bases t hat a COL applicant will incorporate by reference as part of its design and licensing ba ses to meet 10 CFR Part 73.

The COL applicant that references the standard design will addr ess the PSS designs that are not included in the scope of the standard design approval. FSAR Section 13.6.1, Physical Security, includes COL Items 13.6-1 through 13.6-3, which addr ess the establishment of a physical security program for operations and site-specific PSS designs. These COL items direct an applicant that references the standard design to establish o perational programs and to provide security plans, address requirements involving the cent ral alarm station (CAS) consistent with TR-118318 and provide a secondary alarm station that is equal and redundant to the CAS. COL Items 13.6-4 and 13.6-5 direct a COL applicant to develop an access authorization program and a cyber security program, respectively. Where practical, the programs may be implemented in phases, and the COL applicant is to inclu de the phased implementation milestones. This section also notes that ITAAC for site-specifi c SSCs are described in FSAR Section 14.3, Inspections, Tests, Analyses, and Acceptance Cri teria. Specifically, COL

13-27 Item 14.3-2, located in FSAR section 14.3, directs an applicant that references the standard design to provide ITAAC for site-specific SSCs.

Summary of Application

The FSAR sections cited below, and the referenced TR, contain t he applicants descriptions of the PSS and physical security ITAAC (SDAA Part 8) for the stand ard design approval and describe how they meet regulatory requirements.

FSAR: FSAR Section 1.2, General Plant Description, through Section 1.9, Conformance with Regulatory Criteria, describe the scope of the standard design approval. FSAR Section 1.8, Interfaces with Standard Design, addresses the interface requ irements between the standard design approval and the site-specific design. FSAR Figure 1.2-1, Conceptual Site Layout, depicts the general boundaries of structures or components betw een the standard design and site-specific design. FSAR Section 1.8.1, Combined License Inf ormation Items, identifies information that must be provided to license and operate a site -specific NuScale power plant but is not included in the standard design. FSAR table 1.8-1 lists the descriptions of COL information items that are to be addressed by the COL applicant. COL Items 13.6-1 through 13.6-5 address physical security.

FSAR section 13.6 provides design descriptions of the PSS. TR-1 18318, which is incorporated by reference, describes the design of the PSS within the nuclea r island and structures.

The applicant describes conformance with the NRC RGs in FSAR se ction 1.9. Tables 1.9-1 through 1.9-4 identify conformance to RGs, standard review plan s, design-specific review standards, and interim staff guidance. FSAR Table 1.9-2, Confo rmance with Regulatory Guides, identifies the applicants conformance or partial conf ormance with Division 5, Materials and Plant Protection, RGs that apply to security an d lists guidance that includes RG 5.7, Entry/Exit Control for Protected Areas, Vital Areas, a nd Material Access Areas, and RG 5.79, Protection of Safeguards Information, for elements o f the site-specific PSS design details that the COL applicant will address and that do not app ly to the standard design. FSAR table 1.9-3 specifically describes the applicability of standar d review plans to the applicants standard design approval.

SDAA Part 8: SDAA Part 8, Section 3.0, Shared Structures, Systems, and Comp onents and Non-Structures, Systems, and Components Based Inspections, Test s, Analyses, and Acceptance Criteria (ITAAC) Design Descriptions and ITAAC, inc ludes design descriptions and ITAAC for portions of the plant that are common or shared by mu ltiple modules in the standard design approval. SDAA Part 8, Section 3.16.1, Physical Securit y System, describes the standard design commitments for PSS that facilitate the impleme ntation of a physical protection program to protect against potential acts of radiological sabot age. SDAA Part 8, Table 3.16-1, Physical Security System Inspections, Tests, Analyses, and Acc eptance Criteria (ITAAC 03.16.xx), establishes design commitments and ITAAC to verify PSS that are within the scope of the standard design approval. The staff has documented its review of these ITAAC in chapter 14 of this SER.

Technical Specifications: There are no TS established for PSS or operations.

Technical Reports: By letter dated November 23, 2022, the applicant submitted to t he NRC, TR-118318, Revision 0, which describes the security considerati ons in the standard design approval. By letter dated August 28, 2023, the applicant submit ted to the NRC, TR-118318,

13-28 Revision 1. This TR describes the design bases for the PSS desi gns, including plant layout and building configurations, results of evaluations, and identified vital equipment and areas for the standard design approval. The scope of the PSS described in the SDAA is limited to the PSS related to the nuclear island and structures that are within th e scope of the standard design approval. TR-118318 contains safeguards information (SGI), secu rity-related information, and proprietary information; therefore, it is protected in accordan ce with 10 CFR 73.21, Protection of Safeguards Information: Performance requirements, and 10 CF R 2.390, Public inspections, exemptions, requests for withholding.

Section 4.1, Design Element No. 1, through Section 4.24, Des ign Element No. 24, of TR-118318 provide design descriptions and system performance in formation that support the SDAA Part 8 physical security ITAAC. The descriptions correlate to each of the physical security hardware ITAAC in SRP Section 14.3.12, Physical Secur ity HardwareInspections, Tests, Analyses, and Acceptance Criteria; FSAR section 14.2, Initial Plant Test Program; and SDAA Part 8, section 3.16.

TR-118318 identifies PSS that are not within the scope of the s tandard design approval (e.g., the protected area (PA) barrier systems, unattended open ings, isolation zones, vehicle barrier systems (VBSs), PA security lighting, perimeter defensi ve fighting positions, personnel and vehicle access control portals, PA penetrations). COL Item 13.6-1 directs the applicant to describe site-specific PSS designs (i.e., outside of the scope of the standard design approval) and security plans that indicate how engineered and administrat ive controls, management systems, and organization will meet the requirements of 10 CFR Part 73 that apply to an operating nuclear power reactor.

Regulatory Basis

The provisions of 52.139, Standards for review of applications, requires applications filed under this subpart will be reviewed for compliance with the sta ndards set out in 10 CFR Part 73.

The security regulations in 10 CFR Part 73 include performance and prescriptive requirements that, when adequately met and implemented, provide protection a gainst acts of radiological sabotage, prevent the theft or diversion of special nuclear mat erial, and protect SGI.

Under 10 CFR 73.55(b), the NRC requires the COL applicant to de scribe a physical protection system and security organization the objective of which will be to provide high assurance 1 that activities involving special nuclear material are not inimical to the common defense and security and do not constitute an unreasonable risk to public health and safety. Physical protection systems and features are relied on to implement security respon se functions (i.e., detection, assessment, communications, security responsedelays, interdict ions, and neutralization).

1 The general performance objective of 10 CFR 73.55(b)(1) is to provide high assurance that activities involving special nuclear material are not inimical to the common defense and security and do not constitute an unreasonable risk to the public health and safety. In SRM-SECY-16-0073, Staff Requirements SECY-16-0073 Options and Recommendations for the Force-on-Force Inspection Program in Response to SRM-SECY-14-0088, dated October 5, 2016, the Commission stated that the concept of high assurance of adequate protection found in our security regulations is equivalent to reasonable assurance when it comes to determining what level of regulation is appropriate. Throughout this publication, the term high assurance is used in alignment with Commission policy statements that high assurance is equivalent to reasonable assurance of adequate protection.

13-29 The regulations in 10 CFR 73.55(b)(2) establish the performance requirements to protect a nuclear power plant against the design-basis threat (DBT) of ra diological sabotage as described in 10 CFR 73.1(a)(1). The COL applicant must describe how it wi ll meet regulatory requirements, including achieving the objective to protect agai nst the DBT of radiological sabotage. The provisions in 10 CFR 73.54, Protection of digita l computer and communication systems and networks; 10 CFR 73. 55, Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabot age; 10 CFR 73.56, Personnel access authorization requirements for nuclear power plants; 10 CFR 73.58, Safety/security interface requirements for nuclear power reactors; and Appendi x B, General Criteria for Security Personnel, and Appendix C, Licensee Safeguards Conti ngency Plans, to 10 CFR Part 73 establish perfo rmance and prescriptive requireme nts that apply to PSS designs, operational security, management processes, and progra ms.

The applicable requirements for a standard design are limited t o PSS within the scope of the standard design approval. According to 10 CFR 52.79, the COL ap plicant addresses the operational or administrative controls, programs, procedures, a nd processes (e.g., management systems or controls), but these areas are not in the review sco pe of the standard design approval application.

An applicant may apply the latest revision of the following reg ulatory guidance documents and accepted industry codes, standards, or guidance to meet regulat ory requirements:

• The SRP, particularly Section 13.6.2, Physical SecurityRevie w of Physical Security System DesignsStandard Design Certification and Operating Reac tor Licensing Applications, Revision 2, issued June 2015, and section 14.3.1 2, Revision 1, issued May 2010.

The NRC guidance, approaches, and examples described above and in other guidance for methods of compliance are not regulatory requirements and are n ot intended to be all-inclusive.

The COL applicant may use methods or approaches for meeting NRC regulations other than those discussed in agency guidance if such measures satisfy the applicable NRC regulatory requirements.

Technical Evaluation

The staff reviewed the design descriptions of the PSS within th e scope of the standard design approval to determine whether they satisfy the requirements of 10 CFR Part 73. The staffs review consisted of determining whether the applicant provided adequate and reasonable descriptions of the design and technical bases and how the prop osed design will achieve the intended security functions. The staffs review does not includ e the security programs or integrations of engineered systems with administrative controls and management measures and organization to determine whet her they would provide reason able assurance of adequate protection and a finding of an adequate physical security progr am, as specified in 10 CFR 73.55(a) through 10 CFR 73.55(r) for a COL applicant. Th e NRC staff reviewed the identified COL information items to determine specific actions required for the design of the site-specific PSS and the establi shment of security programs th at COL applicants referencing the standard design will address.

The staffs review was limited to the adequacy of the design an d bases for the PSS that is relied on to implement security response functions (i.e., detection, a ssessment, communications, security responsedelays, inter dictions, and neutralization). T he COL applicant must

13-30 demonstrate reasonable assurance of adequate protection against the DBT of radiological sabotage and compliance with the programmatic requirements of 1 0 CFR Part 73, including administrative controls such as people and procedures.

The staffs review included TR-118318, Revision 1, submitted by letter to the NRC on August 28, 2023, and incorporated by reference in FSAR section 13.6.

13.6.4.1 Design Considerations for Physical Security

In FSAR section 13.6 and TR-118318, the applicant described how the PSS is designed to protect against potential acts of radiological sabotage.

TR-118318 descriptions of the PSS design conform to SRP section 13.6.2, Revision 2, which was in effect when the SDAA was docketed. Conforming to guidanc e, the applicants design descriptions address design elements identified in SRP section 13.6.2, Table 13.6.2.1, Design of Physical Security Systems within the Nuclear Island and Stru ctures. The applicant also considered additional PSSs identified in SRP Table 13.6.2.2, D esigns of Physical Security Systems for Plant Area Beyond the Nuclear Island and Structures, which may be included within the scope of the standard design approval or reserved fo r a COL applicant that references the standard design. Section 7, Figures, of TR-118 318 provides the plant layout diagram that identifies SSCs and design configurations of the P SS that are within the scope of the standard design approval.

TR-118318 states that the nuclear island and structures physic al security design provides features to detect, assess, impede, and delay threats up to and including the design basis threat [of] radiological sabotage in compliance with the requir ements of 10 CFR 73.55, Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors against Radiological Sabotage.

TR-118318 supplements the information in FSAR Chapter 13 with d esign and related information, results of evaluations or analyses, and design and performance requirements. The applicants descriptions of security design elements and concep ts (e.g., engineered systems, technologies, and equipment) address the following for the nucl ear island and structures within the scope of the standard design approval:

• the design of the PSS for interior detection, assessment, acce ss control, and security response

• physical barriers (e.g., control (or denial) of access, interi or security response, deterrence, and delay, securing and monitoring of openings, bul let-resistance, protection of vital equipment)

• vital equipment, vital areas, and intrusion detection and cont rol of access systems

• minimum safe standoff distances (MSSDs)

• interior detection and assessment systems

• central (security) alarm station

• illuminations

13-31

• communications

The staff found the following:

Consistent with SRP Section 13.6.2, the applicant adequately co nsidered physical security in the standard design by including design information on PSS within the nuclear island and structures to address security functions tha t meet the applicable requirements of 10 CFR 73.55. A detailed explanation of how the PSS specifically meet the applicable requirements is given below in Sections 13.6.4.2 -13.6.4.4.

TR-118318, Table 5-1, Applicant Responsibilities, states the following:

[The] applicant addresses design elements involving site-specif ic conditions unable to be addressed in the standard design approval (e.g., p rograms, personnel, plans, and procedures) and design element details ex empted in accordance with Criterion 3(a) or 3(b) [described in SRP sectio n 13.6.2]. An applicant that references the standard design is responsible fo r the items listed in Table 5-1.

The items identify information that the COL applicant provides to satisfy COL Items 13.6-1 and 13.6-3. COL Item 13.6-2 states that the COL applicant will be responsible for the requirements in table 5-1 of TR-118318. Table 5-1 includes the COL applicant s responsibilities for providing design details that address PSSs outside the scope of the stand ard design approval and program descriptions and security plans in accordance with the requirements in 10 CFR Part 73. A COL application that addresses COL Items 13.6 -1 and 13.6-3 would include site-specific PSS design details such as the following:

• location and design details for the secondary alarm station

• physical security barriers outside the nuclear island and stru ctures

• isolation zones, PA, and associated intrusion assessment syste ms

• VBS

• exterior personnel, vehicle, and material access control porta ls

• main security building

• secondary power supply for the communication system

• secondary security power system

• bounding MSSD, alarm station survivability, and protection aga inst vehicle bombs

• alarm station functions and redundant capabilities

• detection and assessment functions

• illumination of isolation zone and PA

• secondary alarm station communications

• uninterruptible power system and inline generators or other so urces of backup power

The staff finds that the applicant adequately established the C OL applicants responsibilities for providing the design of PSSs that are not located within or int egral to the construction of the nuclear island and structures and providing security programs t hat are outside of the scope of the standard design approval.

13-32 13.6.4.2 Security Evaluations and Analyses

Vital Equipment Identification Process

TR-118318, Section 4.8, Design Element No. 8, lists vital equ ipment for the standard design approval.

The applicant evaluated reactor design and safety analysis info rmation in the SDAA and supporting analyses and documentation that served as the source for the vital equipment identification process. The applicant indicated that it based i ts identification of vital equipment on the definition of vital equipment in 10 CFR 73.2, Definitions. In TR-118318, section 4.8, the applicant stated the following about identifying vital equipmen t:

An interdisciplinary design team evaluated SSC for vital equipm ent designation.

The team included members of NuScale s Physical Security, Plan t Operations, Electrical Engineering, Instrumentation and Controls Engineerin g, Civil/Structural Engineering, Nuclear Safety Engineering, and Probabilistic Risk Assessment Engineering. Using the 10 CFR 73.2 definition for vital equipme nt, the team evaluated systems and components fo r potential inclusion as vital equipment.

The applicant applied the definition of vital equipment in 10 C FR 73.2, which states that vital equipment means any equipment, systems, devices, or mate rial, that failure, destruction, or release of which could directly or indirectly e ndanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect public health and safety following such a f ailure, destruction, or release are considered to be vital.

The staff finds the following:

The applicant considered information from the design and safety analyses to identify vital equipment and determined that the applicant established a reasonable process and criteria to identify vital equipment for the standard design ap proval using the definition of vital equipment in 10 CFR 73.2.

Vital Equipment List

The details of the SSCs (e.g., frontline systems and supporting systems) that make up the vital equipment for the standard design approval are identified as SG I; therefore, they are protected in accordance with 10 CFR 73.21 and withheld from the public in accordance with provisions of 10 CFR 2.390. In its review of the applicants vital equipment list, the staff did not identify cases in which the applicant excluded frontline systems or functions or primary supporting systems that meet the definition of vital equipment.

The staff finds the following:

The applicant identified and provided a sufficiently complete a nd accurate list of vital equipment for the standard design approval based on the definit ion of vital equipment in 10 CFR 73.2.

13-33 Vital Areas

The requirements in 10 CFR 73.55(e)(9)(i) state the following:

Vital equipment must be located only within vital areas, which must be located within a protected area so that access to vital equipment requi res passage through at least two physical barriers, except as otherwise app roved by the Commission and identified in the security plans.

The applicant established vital areas within the scope of the s tandard design approval based on the safety-related systems and components identified on the vit al equipment list and other areas required by 10 CFR 73.55(e)(9) to be designated as vital (the M CR, CAS, spent fuel pool (SFP),

and secondary power supply). TR-118318, Section 4.10, Design E lement No. 10, addresses the specific areas that are designated vital, which are documen ted in TR-118318, section 7.

The applicant identified the vital areas that consist of the nu clear island and structures. {Given the diverse locations of equipment considered vital, the applic ant established certain building perimeters that enclose the vital equipment as boundaries of th e vital areas}. The applicant indicated that the designs and configurations of vital areas re strict access and limit access pathways, which facilitates the im plementation of security for unauthorized access. TR-118318, section 7 (figures 1-10), shows the specific boundaries of a st ructure that form the vital areas within the nuclear island and structures. The figures also show the exterior boundaries of the plant structures that form vital areas. The detailed boundaries of the vital areas are identified as SGI and are protected in accordance with 10 CFR 73.21. The figu res in TR-118318, section 7, also show the specific boundaries of the SFP that form the vita l area.

TR-118318, Section 4, Design Element Responses, describes the designs of the PSS for the vital area portal detection system, interior assessment and mon itoring, the vital area access control system, and the alarm system associated with the protec tion of the vital areas.

Specifically, TR-118318, section 4.1; Section 4.2, Design Elem ent No. 2; and Section 4.9, Design Element No. 9, describe the design requirements for sy stems and components that provide access control, locking, and intrusion detection; asses sment; communications; and emergency egress for the vital areas. The design descriptions i nclude the system interfaces with security alarm stations necessary for the redundant intrusion d etection alarm indications and assessment of alarms and physical barriers to address unauthori zed access. TR-118318, Section 4.6, Design Element No. 6, addresses the design of se curity systems for securing, monitoring, detecting intrusion, and controlling access of vita l area barrier system openings.

TR-118318, Section 4.16, Design Element No. 16, describes the design of system logic sequences for initiating alarm conditions and the supervision a nd monitoring of alarm signal integrity and system normal and trouble conditions, such as tam pering, loss of or degraded signals, or a short in the system signal circuits for detecting the loss of system functions or abnormal system functions.

TR-118318, section 4.1, establishes design requirements for int erfaces between the access control system and locking devic es in the event of a loss of bo th primary and secondary power and identifies the design requirements for protecting control a nd power wiring against physical tampering. TR-118318, section 7, Figure 38, Security Power One Line Diagram, and Figure 39, Simplified Security System Interconnection Diagram, show the configurations for the design of the primary and secondary power supply for perfor ming security functions, vital entry controls, and alarms with intrusion detection systems tha t annunciate at alarm stations to

13-34 comply with regulatory requirements. The vital area physical bo undaries are spatially separated from the PA boundary. TR-118318, section 7, provides the vital area boundaries.

The staff finds the following:

The applicant identified and designated vital areas to include vital equipment listed in TR-118318, Section 4.8, Design Element No. 8, and established that no vital equipment within the scope of the standard design approval is l ocated outside of areas designated as vital.

The applicant adequately described the design bases for physica l barriers for the nuclear island and structures that have been designated as vita l areas to address one of two barriers in accordance with 10 CFR 73.55(e)(9)(i). The othe r barrier is the PA barrier, which is not within the scope of the standard design a pproval and would be addressed by a COL applicant referencing the standard design ap proval.

The applicant adequately described the design of physical barri ers to control access to the vital areas within the scope of the standard design approva l and satisfied the requirements of 10 CFR 73.55(e)(1). The design provided the con trol and delay of access necessary to facilitate the implementation of security r esponses.

The applicant adequately addressed the requirements of 10 CFR 7 3.55(e)(9)(v) by designating the MCR, CAS, SFP, and secondary power supply as vi tal areas.

The COL applicant that references the standard design is respon sible for location and design details for the secondary alarm station that is equal an d redundant to the CAS.

Security Computer Design Requirements

TR-118318 provides system functional diagrams showing the desig n interfaces of security computer systems with subsystems for performing redundant intru sion detection and assessment, access controls, and the interfaces between alarm s tations. TR-118318, section 7, figure 39, shows the design diagram addressing the capabilities of the systems for data communication and interfaces with subsystems and components.

The security computer systems support the plant security functi ons by continuous access control, monitoring of doors, and the prompt reporting and perm anent recording of all alarm points and system conditions (e.g., intrusions, tampers, and tr ouble conditions). The security computers are located within vital areas, and access is control led. TR-118318, section 7, figure 39, shows the redundant security computers, which are sp atially separated and independently powered by diverse security power subsystems; eac h one is independently capable of providing the required security functions. The secur ity computer systems network is isolated and does not connect to any other plant system, comput er, or data network. The CAS workstation and monitors are used to display the area of the or iginating alarm.

The security computer systems will be capable of data communica tions using the dedicated network. The computers, graphic displays, closed-circuit televi sion system (CCTV) servers, and digital video recording systems are connected to the network. T he network configuration allows communication between devices to provide information to the ala rm station operators.

TR-118318, section 7, figure 39, shows the functional diagram f or the design of the security computer systems network. The figure shows how the network will be configured and how the

13-35 backbone and infrastructure will accommodate the security devic es. The remote field devices, such as intrusion detectors, CCTV, door card readers, and secur ity alarming devices, are connected to the network and will be supplied by the COL applic ant to complete the total integrated security systems. The security circuits are supervis ed and tamper-indicating for indication of system conditions and operability.

The computer systems are also designed such that an alarm stati on operator cannot change the status of a detection point or deactivate a locking control dev ice at a PA or vital area portal without the knowledge and concurrence of the alarm station oper ator in the other alarm station.

All wiring that connects the computer systems with remote acces s control components (e.g., card readers, controllers) and with other security subsy stems (e.g., perimeter intrusion detection) is configured as electronically supervised circuits. The primary and secondary cables between the alarm stations and controllers are separated to pre vent simultaneous damage caused by a sabotage attempt or any unintended actions.

The security computer systems also interface with the CCTV. The functions of the CCTV system include operating cameras that provide visual monitoring of the area with an alarm if the intrusion detection system actuates and allow assessment of the area with an alarm.

A computer-based automatic access control system controls perso nnel access for the standard design approval. The computer for the access control system wil l also interface with security subsystems, such as intrusion detection and CCTV images. The ac cess control system permits entry only to those persons authorized to enter specific areas at the access point into the PA, buildings, and vital areas. Access point activities (including open or close door status, alarm indications, and attempts at unauthorized entry) are recorded. For continuity of access control functions, the system provides for automatic switchover to an u ninterrupted power supply and secondary power if primary power is interrupted.

In TR-118318, section 4.1.2, the applicant indicated that the C OL applicant is responsible for providing vendor-specific design descriptions for the assessmen t system.

The COL applicant that references the standard design must esta blish and describe how it will meet the requirements of 10 CFR 73.54. RG 5.71, Cyber Security Programs for Nuclear Facilities, and Nuclear Energy Institute (NEI) report NEI 08-0 9, Revision 6, Cyber Security Plan for Nuclear Power Reactors, dated April 2010, provide acc eptable methods and approaches for developing and establishing a cybersecurity prog ram and submitting a cybersecurity plan to satisfy the requirement of 10 CFR 52.79(a )(36)(iii). COL Item 13.6-5 and TR-118318, Table 5-1, address the need for this information.

The staff finds the following:

The applicant adequately described the design of independent an d redundant security computer systems and interfaces that support redundancy for the alarm station security functions of intrusion detection, assessment, and access contro l.

The COL applicant that references the standard design is respon sible for meeting the requirements of 10 CFR 73.54 for a cybersecurity program that p rotects digital computers and communication systems and networks.

The determination and finding on whether the applicant has met the requirements of 10 CFR 73.54 for a cybersecurity program are beyond the scope o f the standard design

13-36 approval. The staff will evaluate compliance with the regulator y requirements for an adequate cybersecurity program as part of the review of a COL o r an operating license application.

13.6.4.3 Design for Physical Barriers

Vital Area and Security Delay Barriers

Figure 1.2-1 in FSAR Section 1.2.1, Principal Site Characteris tics, shows the separation from a PA boundary that a COL applicant will establish to comply wit h the requirements of 10 CFR 73.55(e)(8). The physical barriers for the PA perimeter and the vital area barriers and access controls delay an unauthorized persons access to a vita l area and allow security responders to interdict the unauthorized person before they can reach a vital area boundary and delay their access into a vital area.

The applicant described the design of the PSS that protects the access to vital areas.

Specifically, TR-118318 describes the design requirements for t he protection of unoccupied vital areas, establishment of vital area physical barriers and separa tion from the PA, protection of penetrations through vital area physical barriers, minimization of entry points, hardening of vital area portal egress, control of access to vital areas, and detec tion and assessment of unauthorized access or intrusion for security response.

Section 3, Security by Design, of TR-118318 describes the des ign and construction of vital area barriers, the vital area access control system, and alarm station design (bullet-resistant).

The configurations of vital area boundaries are described in se ctions 4.2, 4.3, 4.6, 4.7, 4.14, and 7 (figures 1-10) of the TR. Sections 4.1-4.4, 4.7, 4.9, and 7 (Figure 37, Figure 37 Central Alarm Station Layout) describe the minimum construction design requirements for walls, floors, and ceilings to establish physical barriers that enclose the de signated vital areas, the MCR, and the CAS to satisfy bullet-resisting requirements. TR-118318, se ction 4.10, describes the identification of the walls, floor, and roof that form the boun daries enclosing the SFP, which is designated as vital in accordance with 10 CFR 73.55(e)(9)(v) an d (9)(vi).

In TR-118318, Section 3, item 16, the applicant described physi cal barriers within the reactor building to delay the DBT adversary. The applicant identified p reliminary locations for such barriers in TR-118318, Section 4.4, Design Element No. 4, and Section 7 (Figure 40, HVAC Barrier Simplified Drawing, and Figure 41, {Mall Gate Simplif ied Drawing}), based on recommended typical design of physical barriers. The applicant indicated that final delay credited for physical barriers, including access and exit barri ers, will be the COL applicants responsibility. The COL applicants protective strategy must account for site-specific conditions, in accordance with 10 CFR 73.55(b)(4), for the design of a phys ical protection system that protects against the DBT of radiological sabotage.

TR-118318, Section 4.7, Design Element No. 7, describes the m inimum design requirements of the walls, floor, and ceiling needed to meet the function of bullet-resisting barriers. The design descriptions include the requirement for doors to meet Underwri ters Laboratories (UL) 752, Standard for Bullet-Resisting Equipment, which is an acceptab le standard for meeting NRC requirements as discussed in SRP section 13.6.2. The design req uirements include the protection of openings, such as for HVAC, that penetrate the vi tal area barriers. TR-118318, section 7 (figures 37 and 39), describe the barriers for protec ting the CAS and typical protection for HVAC penetrations through the vital area barriers. The desi gn for HVAC penetration openings requires the installation of barriers that allow airfl ow but do not allow the passage of a

13-37 person. The physical barriers installed for HVAC penetrations a re to restrict access and provide a security delay against forced entry.

TR-118318, sections 3.0, 4.2-4.6, 4.8, 4.9, 4.13, 4.14, 4.19, a nd 4.21, provide additional design descriptions for the protection of penetrations through the vit al area physical barriers.

Engineered systems or features that provide delay, denial, cont rol, detection, and monitoring functions for unauthorized access must protect all openings tha t exceed a standard opening that is too small for the passage of an individual. TR-118318, secti on 7, shows the typical configuration of a vital area door with locking and alarming ca pabilities and the locations for installations of bullet-resistant doors (Figures 20, Figure 20 Control Building Security Equipment 100-0, 21, Figure 21 Control Building Security Eq uipment 125-0, and 37). The penetrations of HVAC ducts, cable trays, ventilation fans, and other such features are protected to ensure that the integrity of the vital area barrier is not d ecreased and that the penetrations do not allow for the passage of a person. TR-118318, section 7 (fi gure 39), shows the design configurations of vital area access controls, locks, and alarms for PSSs that are included in the standard design approval.

The applicant indicated that barriers to protect penetrations t hrough the vital area barriers will provide for a delay like that afforded by the adjacent portion of the vital area barriers or will otherwise provide the delay needed, and these barriers will com ply with the regulatory requirements for a security barrier in 10 CFR 73.2. The securit y design features include hardened doors that delay forced entry and resist mechanical an d explosive breaching to allow for security responses. TR-118318, section 7 (figures 4-7, 10, 14-17, 20-21), shows locations and doors that will be designed to delay unauthorized entries i nto designated vital areas and to control access to vital equipment.

TR-118318, sections 4.6, 4.7, and 4.9, describe the design and construction requirements for delay to forced entry and locking mechanisms to secure vital ar ea portals for ingress and egress. The design includes locking devices that allow for rapi d egress during an emergency.

UL-listed exit devices or panic and locking hardware account fo r normal and emergency operations and functions in the event of a loss of power.

The system functional diagrams in TR-118318, section 7 (figure 39), show the design for the access control system, door control, intrusion detection compon ents, and network management systems for vital areas. The design provides redundant systems for access control functions at alarm stations. Similarly, the design details of the intrusion detection and assessment systems show and establish the designed redundancy and separation of sy stems that provide intrusion detection and assessment functions.

The staff finds the following:

The applicant adequately described the design bases for the phy sical barriers of the nuclear island and structures that are within the scope of the standard design approval to meet 10 CFR 73.55(e).

A COL applicant that references the standard design will analyz e site-specific conditions and describe the integration and design of additional physical barriers for meeting the requirements of 10 CFR 73.55(e), including sufficient delay to support the required security response time.

13-38 The applicant adequately addressed the requirements of 10 CFR 7 3.55(e)(9)(ii) by providing a standard design that protects all vital area access points and vital area emergency exits with intrusion detection equipment and locking devices that satisfy the vital area entry control requirements and meet the requirement in 10 CFR 73.55(e)(9)(iii) that unoccupied vital areas must be locked and alarmed.

The applicant adequately described the design and performance r equirements of the PSS for access control. Specifically, the applicants design ad dresses the requirements of 10 CFR 73.55(g) as they pertain to access to the nuclear isl and and structures of the standard design approval. The PSS des ign includes access control systems that meet the requirements of 10 CFR 73.55(g)(1)(i)(A) and (i)(B) at the vital area boundaries for the control of personnel, protection of openings with physical barriers with locking devices to delay access, inclusion of intrusion detection syste ms to detect unauthorized access, and provision of equipment to assess physical condition s of designated vital areas.

The applicant adequately met the prescriptive requirements in t he 10 CFR 73.2 definition for physical barrier by providing the design of PSS or by cre diting building structural systems that satisfy the require ment for using brick, cinder block, concrete, steel, or comparable material for the construction of walls, ceilings, an d floors. The openings in such structures are secured by grates, doors, or covers of cons truction and fastening with sufficient strength such that any opening will not lessen the integrity of the structures.

The staff determined that the 10 CFR 73.2 prescriptive requirem ents for physical barriers related to site specific designs for fence construction do not apply to the physical barrier systems described for the nucl ear island and structures within the scope of the standard design approval because it is not part of the nuclear island.

The COL applicant will address and satisfy the requirements for site-specific physical barriers.

Bullet-Resistant Barriers

In TR-118318, the applicant described the minimum construction standards for the walls, floors, and ceilings of the MCR and CAS and the exterior and interior b oundaries of buildings that have been designated as physical barriers that enclose vital areas. The applicant included design information for protecting openings and penetrations through vi tal area barriers, as previously discussed in this SER.

The structural design for walls, floors, and ceilings consists of varying thicknesses of reinforced concrete that exceed the minimum thickness required for structu res, walls, and locations of doors needed to meet bullet-resistance requirements. The walls, floors, and ceilings of the CAS are of a thickness beyond that chosen as a baseline minimum req uired for resisting bullets.

The building that houses the CAS is designated as a vital area and will be constructed and installed with access controls and protection of openings and p enetrations to meet vital area and bullet-resistance requirements. The areas that contain the alarm station will also be designated as vital areas and will meet the appropriate vital a rea requirements. The applicant indicated that the design of the last access control location i s outside the scope of the standard

13-39 design approval and will be specified by the COL applicant. The COL applicant will include the construction requirements for bullet-resistant physical barrier s.

The applicant indicated that the MCR and CAS walls, floors, cei lings, doors, and windows are designed and will be constructed to meet a minimum bullet-resis tance. The applicant indicated that the walls, floors, and ceilings of the MCR have a minimum thickness of reinforced concrete that is credited to meet the physical protection requirement fo r a bullet-resistant barrier. The thickness of concrete exceeds the bullet-resistance requirement s of the Underwriters Laboratories (UL) 752 standard, as described in SRP Section 13. 6.2. Any doors on the MCR boundary will be bullet-resisting to the minimum of the UL 752 standard. The windows on doors that lead into the MCR will be bullet resistant.

The staff finds the following:

The applicant adequately described the design for the MCR and C AS to meet the requirements of 10 CFR 73.55(e)(5). The design provides protect ion for the MCR and CAS with a bullet-resistant enclosure by crediting structural e lements of the standard design approval and providing hardened doors and engineered bar riers for protecting openings and penetrations of the bullet-resistant enclosure.

The design of the last access control to the PA is outside the scope of the standard design approval.

Vital Area Doors

TR-118318, section 7 (figures 11-17, 20, 21, 23, 24), establish es door schedules for the design and locations of doors with card reader access, lock, and alarm. The figures in TR-118318, section 7, show the typical vital area access control doors and the design configuration for the installation of intrusion detection, access control, locking, a nd other design features for securing vital areas. To provide delay and access control, exterior door s have a delay capability equivalent to the delay capability credited for the structure w alls. The remaining exterior doors are hardened to provide resistance to penetrations with delay c ontrol as stated in TR-118318.

The design descriptions in TR-118318, section 4.9, address requ irements to provide exit devices on vital area egress doors that require emergency egres s capability. Utility penetrations, such as HVAC ducts and other piping, will be equipped with barr iers hardened with construction material that delays unauthorized access. Section 7 (figures 25 -36 and 40) depicts HVAC barriers.

The staff finds the following:

The applicants description of the design bases for physical ba rriers, as detailed in TR-118318, adequately addresses the requirements of 10 CFR 73.5 5(e)(4) by providing the design of physical barrier systems that secure openings or penetrations into the structural boundaries of the nuclear island and structures.

Vehicle Barrier System

The COL applicant will address the construction and installatio n of the VBS. However, in TR-118318, Section 4.11, Design Element No. 11, the applicant established and showed the bounding MSSD for protecting the nuclear island and structures, including the CAS, from the

13-40 maximum DBT vehicle-borne explosive. Table 4-1, Minimum Stando ff Distances, in TR-118318 shows the required MSSDs for the construction and ins tallation of a continuous VBS, along with results for the required minimum standard of di stance for the CAS and the protection of physical security SSCs and personnel that must be met for a bounding MSSD.

TR-118318, section 4.11, indicates that the VBS must be located at least the bounding MSSD from the nearest external surface of any vital areas. The dista nce required is based on methods or approaches referenced in NUREG/CR-6190, Protection against Malevolent Use of Vehicles at Nuclear Power Plants, dated March 17, 2004. The applicant a pplied Department of Defense methods and guidance for predicting blast effects and structura l responses to assess and evaluate the various distances that would be safe for SSCs for the safety of nuclear plant operations and personnel. They included UFC 3-340-02, Structur es to Resist the Effect of Accidental Explosion, dated 2008, and U.S. Army Corps of Engin eers Protective Design Center TR-06-08, Single Degree Freedom Structural Response Limits for Antiterrorism Design, dated 2008.

TR-118318, section 4.11, provided required MSSDs for the struct ures within the scope of the standard design approval. Item 25 in table 5-1 of TR-118318 cla rifies that it is a COL applicants responsibility to verify that the door is designed appropriatel y.

In TR-118318, table 4-1, the applicant provided the minimum sta ndoff distances analyzed for the reactor building to protect against the DBT vehicle-borne e xplosive. The applicants analysis did not include the determination of minimum standoff distances for the secondary alarm station, personnel in open or in nonhardened enclosures, and blast-and bullet-resistant enclosures, which are not included in the scope of the standard design appr oval.

In TR-118318, Section 6, References, the applicant identified engineering calculations, analyses, assessments, or other references that provide the des ign and technical basis for the summary descriptions of designs, design bases, results, and con clusions.

The staff finds the following:

The applicant adequately assessed and documented the required M SSDs for the nuclear island and structures based on a maximum quantity of ex plosives associated with the adversarial characteristics of the DBT.

The applicant adequately established the design basis for a loc ation of the VBS that would be sufficient to protect safety-related SSCs or loss of SFP coolin g against the DBT vehicle-borne explosive threats.

13.6.4.4 Design Features Facilitating Security Response

The applicant did not include the design of PSS that facilitate security, such as hardened defensive fighting positions, in the scope of the standard desi gn approval. Other than the PSS described above, the design of the fighting positions (e.g., locations, blast and bullet-resistance, firing ports, material construction, fully o r partially enclosed fighting positions to protect security personnel from attack, blast prot ection, environmental controls and protection, lighting, communications) and other fe atures (e.g., delay, protection against hand-thrown explosives) for security respons es to interdict or neutralize the DBT must be provided by the COL applicant.

13-41 Combined License Information Items

The staff reviewed the applicants descriptions of COL informat ion items that a COL applicant is directed to address if referencing the NuScale US460 standard d esign. The applicant provided COL information items related to Section 13.6 in FSAR Table 1.8 -1, Combined License Information Items.

Table 13.6-1 NuScale COL Information Items Related to Section 1 3.6

Item No. Description FSAR Section COL Item An applicant that references the NuScale Power Plant US460 13.4 13.4-1 standard design will provide site-specific information, includi ng implementation milestones, for Operational Programs:

Security (refer to Section 13.6)

COL Item An applicant that references the NuScale Power Plant US460 13.5 13.5-6 standard design will describe the process to manage the development, review, and approval of the site-specific procedur es that operators use in the main control room and locally in the plant, including normal operating procedures, abnormal operating procedures, and emergency operating procedures. The applicant will describe the classification system for these procedures, a nd the general format and content of the different classifications. The categories of procedures listed below will be included: [] pla nt security procedures.

COL Item An applicant that references the NuScale Power Plant US460 13.6 13.6-1 standard design will provide the following: Security Plans (Physical Security, Security Training and Qualification, and Safeguards Contingency Plans); proposed site security provisions to be implemented during construction and as modules are completed and become operational; and elements of the physical security system not located within the nuclear island and structures.

COL Item An applicant that references the NuScale Power Plant US460 13.6 13.6-2 standard design will be responsible for the requirements described in Table 5-1 of TR-118318, NuScale Design of Physical Security Systems (Reference 13.6-1).

COL Item An applicant that references the NuScale Power Plant US460 13.6 13.6-3 standard design will provide a secondary alarm station that is equal and redundant to the central alarm station.

COL Item An applicant that references the NuScale Power Plant US460 13.6 13.6-4 standard design will provide a description of the Access Authorization Program.

COL Item An applicant that references the NuScale Power Plant US460 13.6 13.6-5 standard design will provide a Cyber Security Plan.

TR-118318, Section 5, Summary and Conclusions, states that th e COL applicant will be responsible for addressing site-specific conditions (e.g., prog rams, personnel, plans, procedures) and design element details that are not addressed i n the standard design approval, based on the guidance of SRP section 13.6.2 (i.e., Criterion 3( A) and 3(B)).

13-42 TR-118318, table 5-1, identifies 25 site-specific PSS design an d configuration details (items) that the COL applicant that references the standard design will address, including the following:

• Provide the location and design details for the secondary alar m station (item 4).

• Provide design details for physical barriers located outside t he nuclear island and structures (item 5).

• Provide design details for isolation zones, associated intrusi on detection monitoring equipment, and areas of the PA perimeter without isolation zone s (item 6).

• Provide vehicle barrier design details (item 7).

• Provide design details for the exterior personnel, vehicles, a nd material access control portals (item 8).

• Provide design details for the secondary alarm station and the main security building (item 10).

• Provide design details for, and placement of, the communicatio n system secondary power supply (item 11).

• Provide design details for, and placement of, the secondary se curity power supply (item 12).

• Ensure that the site-specific characteristics are bounded by t he calculated minimum standoff distances and ensure the survivability of the security alarm station (item 13).

• Ensure that the site-specific physical security design is boun ded by the blast analysis (item 14).

• Ensure that the CAS and secondary alarm station are designed a nd equipped in accordance with the DBT of radiological sabotage such that no s ingle act can simultaneously remove the ability of both alarm stations to (1) detect and assess alarms, (2) initiate and coordinate an adequate response to alarms, (3) summon offsite assistance, and (4) provide effective command and control (item 17).

• Design the secondary alarm station such that the CAS and the s econdary alarm station are functionally redundant (item 18).

• Ensure that the alarm system design does not allow a change in the status of a detection point, locking mechanism, or access control device without the knowledge and concurrence of the alarm station operator in the other alarm st ation (item 19).

• Provide design details for specific security illumination for the isolation zone and accessible external PAs (item 21).

• Provide design details for the communication equipment in the secondary alarm station (item 22).

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• Describe the independent security power sources that consist o f fully charged uninterrupted power supply batteries, inline generators, or oth er power sources (item 23).

• Procure, install, and test applicable surveillance, observatio n, and monitoring equipment (item 24).

• Verify the reactor building equipment door is designed to with stand the DBT (item 25).

Table 5-1 of TR-118318 identifies commitments related to the se curity operational program that a COL applicant must complete to establish elements of a physic al security program:

• Establish, maintain, and implement a standalone insider mitiga tion program (item 1).

• Establish, maintain, and implement a site-specific cybersecuri ty plan (item 2).

• Establish and implement an access authorization system or prog ram with a numbered photograph identification badge system for controlling access t o PAs and vital areas (item 3).

• Develop and implement a comprehensiv e site-specific physical s ecurity program description for PSS (item 9).

• Test intrusion detection and assessment equipment to ensure th at the requirements of 10 CFR 73.55(i)(3)(i) through 10 CFR 73.55(i)(3)(v) are met bef ore declaring that the systems are operable (item 15).

• Test intrusion detection systems to ensure that the recordkeep ing capability meets the requirements of 10 CFR 73.55(i)(4)(ii)(H) and 10 CFR 73.70(f) b efore declaring that the intrusion recording system is operable (item 16).

• Select the appropriate vendors alarm station design (item 20).

The staff finds the following:

The COL information items listed in FSAR, Section 13.6, address site-specific designs of the PSS that are outside the scope of the standard design appro val. In addition to the information reviewed in FSAR, Section 13.6, the COL applicant m ust provide information showing compliance with applicable requirements (i.e., for a se curity plan, access authorization program, and cybersecurity plan), including addre ssing the COL information items described above.

Conclusion

For the reasons discussed above, the staff concludes that the a pplicant has considered and provided the PSS in the NuScale US460 Standard Power Plant, wit hin the scope of the standard design approval, to facilitate the implementation of a physical protection program to protect against potential acts of radiological sabotage. As fur ther stated above, NuScale proposed that the standard design has adequately described the plant layout for physical protection and identified vital equipment and areas for meeting, in part, specified requirements of 10 CFR 73.55.

13-44 As explained above, NuScales proposed standard design of the P SS, including system location and configuration, is adequate with respect to the nuclear isla nd and structures within the scope of the standard design approval. This conclusion is limited to the adequacy of NuScales description of the design bases of the PSS and features within the scope of the standard design that are relied on to implement security response functions (i. e., detection, assessment, communications, security responsedelays, interdictions, and ne utralization). The demonstration of reasonable assurance of adequate protection ag ainst the DBT required by NRC regulations and compliance with the programmatic requiremen ts (including administrative controls such as people and procedures) of NRC regulations for physical protection are to be addressed by a COL applicant that is seeking a COL to construct and operate a nuclear power plant.

Fitness for Duty

In 10 CFR Part 26, Fitness for Duty Programs, the NRC prescri bes requirements and standards for the establishment, implementation, and maintenanc e of FFD programs (73 FR 17176; March 31, 2008, as amended).

FSAR Section 13.7, Fitness for Duty, states, in part, that FF D is an operational program and is not applicable to new plant designs.

Conclusion

The staff finds the statement in FSAR section 13.7 on program a pplicability acceptable, because the applicant referencing this standard design is respo nsible for providing an FFD program description and implementation as described in 10 CFR P art 26.

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