ML18054B607

Nuscale Power, LLC - Submittal of the Approved Version of Nuscale Topical Report TR-0815-16497, Safety Classification of Passive Nuclear Power Plant Electrical Systems, Revision 1 (CAC No. RQ6002)

ML18054B607
Person / Time
Site:	PROJ0769
Issue date:	02/23/2018
From:	Bergman T A NuScale
To:	Document Control Desk, Office of New Reactors
Shared Package
ML18054B606	List: ML18054B607
References
CAC RQ6002, LO-0118-58309
Download: ML18054B607 (224)

Text

LO-0118-58309 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com Docket: PROJ0769 February 2, 2018 U.S. NuclearRegulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike

Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of the Approved Version of NuScale Topical Report TR-0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1 (CAC No. RQ6002)

REFERENCE:

Letter from Frank Akstulewicz (NRC) to Thomas Bergman (NuScale), "Safety Evaluation for Topical Report: 0815-16497, Revision 1, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," dated December 13, 2017 (ML17339A533). By the referenced letter dated December 13, 2017, the NRC issued a final safety evaluation report documenting the NRC staff conclusion that the NuScale topical report TR-0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, is acceptable for referencing in licensing applications for the NuScale small modular reactor design. The referenced NRC letter requested that NuScale publish the approved version of TR-1015-18653, within three months of receipt of the letter. Accordingly, Enclosure 1 to this letter provides the approved version of TR-0815-16497-P-A, Revision 2. This enclosure includes the December 13, 2017 NRC letter and its final safety evaluation report, the NuScale response to NRC requests for additional information, and documentation of the final Topical Report submittal, Revision 1. Enclosure 1 contains proprietary information. NuScale requests that the proprietary enclosure be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 contains the nonproprietary version of the approved topical report package. This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

Please contact Jennie Wike at 541-360-0539 or at jwike@nuscalepower.com if you have any questions.

Sincerely, Thomas A. Bergman Vice President, Regulatory Affairs NuScale Power, LLC Distribution: Frank Akstulewicz, NRC, OWFN-8H4A Greg Cranston, NRC, OWFN-8G9A Omid Tabatabai, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A y, T homas A A A A A A A A A A A A A A A. Ber g man V ice Presi de de de de de de de de de de de d de e e e nt , Re g u la a a a a a a a a a a a a a ator y Affai rNuScalePo we e r r r r r r r r, r r r r r r L L L L L L L L L L L L L L L LC LC LC LC LC LC LC LC LC LC LC LC LC LC LC LC L LC LO-0118-58309 Enclo sure 1: NuScale Topical Report, TR-0815-16497-P-A, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, proprietary version Enclosure 2: NuScale Topical Report, TR-0815-16497-NP-A, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, nonproprietary version Enclosure 3: Affidavit of Thomas A. Bergman, AF-0118-58310

LO-0118-58309 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

Enclosure 1:

NuScale Topical Report, TR-0815-16497-P-A, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, proprietary version

LO-0118-58309  :

NuScale To pical Report, TR-0815-16497-NP-A, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 2, nonproprietary version Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report TR-0815-16497-NP-A Rev. 1 © Copyright 201 8 by NuScale Power, LLC Contents Section Description A B C D Letter from F rank Akstulewicz (NRC) t o Thomas Bergman (NuScale), "S afety Evaluation f or Topical Report: 0815-16 497 , Revision 1 , "S afety C lassification of P assive Nuclear Power P lant E lectrical Systems ," CA C No RQ 600 2," d ated December 1 3 , 20 17. NuScale Topical Report: S afety Cla ssification o f P assive Nuclear P ower P lant E lectr ical S ystems , TR-0 8 15-16 497-NP-A, Re vision 1 Letter f rom T homas A. B ergm an (NuScale) to NRC, "NuScale Power, LLC Submittal of R esponse t o Request for Additional I nforma tion Letter No. 8 for t he Review of NuScal e Topical R eport, T R-0 815-16 497 , "Safety C lassification of P assive Nuclear P ower P lant Electrical Systems ," R evision 0,"

dat ed December 5 , 201 6 (ML 16340D339). Letter from Thomas B ergman (NuScale) to NRC, "NuScale Power, L LC, S ubmittal o f Topical Report T R-0 815-16 497 , "S afety C lassificati on of Passive Nuclear P ower P lant E lectrical Systems, Re vision 1 (CAC N o. R Q6002)," dat ed February 17 , 201 7 Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report TR-0815-16497-NP-A Rev. 1 © Copyright 201 8 by NuScale Power, LLC

Section A December 13, 2017

Mr. Thomas Bergman Vice President, Regulatory Affairs

NuScale Power, LLC 1100 NE Circle Boulevard, Suite 200 Corvallis, OR 97330

SUBJECT:

SAFETY EVALUATION FOR TOPICAL REPORT 0815-16497, REVISION 1, "SAFETY CLASSIFICATION OF PASSIVE NUCLEAR POWER PLANT ELECTRICAL SYSTEMS" (CAC NO. RQ6002)

Dear Mr. Bergman:

By letter dated October 29, 2015, NuScale Power, LLC (NuScale), submitted Topical Report (TR) 0815-16497, Revision 0, "Safety Classification of Passive Nuclear Power Plant Electrical Systems" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15306A263). On February 7, 2017, NuScale submitted proprietary and nonproprietary versions of TR-0815-16497, Revision 1 (ADAMS Accession No. ML17048A459).

The U.S. Nuclear Regulatory Commission (NRC) staff has found that the TR 0815-16497, Revision 1, is acceptable for referencing in licensing applications for the NuScale small modular reactor design to the extent specified and under the conditions and limitations delineated in the enclosed safety evaluation report (SER). The SER defines the basis for acceptance of the TR.

The NRC's acceptance applies only to matters approved in the subject TR. We do not intend to repeat our review of the acceptable matters described in the TR. When the report appears as a reference in license applications, our review will ensure that the material presented applies to the specific plant involved. Regulatory licensing action requests that deviate from this TR will be subject to additional staff reviews in accordance with applicable review standards. In accordance with the guidance provided on the NRC's TR website (http://www.nrc.gov/about-nrc/regulatory/licensing/topical-reports.html), we request that NuScale publish an accepted version of this TR within three months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed safety evaluation between the title page and the abstract. It must be well indexed such that the information is readily located. Also, it must contain in its appendices historical review information, such as questions and accepted responses, and original report pages that were replaced. The accepted version shall include an

"-A" (designated accepted) following the report identification symbol.

T. Bergman 2 If the NRC's criteria or regulations change so that its conclusion in this letter, that the TR is acceptable, is invalidated, NuScale and/or the applicant referencing the TR will be expected to revise and resubmit its respective documentation, or submit justification for the continued applicability of the TR without revision of the respective documentation.

Sincerely, Francis M. Akstulewicz, Director /RA Anna Bradford Acting for/

Division of New Reactor Licensing Office of New Reactors Docket No. PROJ0769

Enclosure:

Safety Evaluation

cc w/encl: DC NuScale Power LLC Listserv

1 NUSCALE POWER, LLC SAFETY EVALUATION FOR TOPICAL REPORT TR-0815-16497, REVISION 1, "SAFETY CLASSIFICATION OF PASSIVE NUCLEAR POWER PLANT ELECTRICAL SYSTEMS

" (CAC. NO. RQ6002)

1.0 Introduction

By letter dated October 29, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML15306A263), NuScale Power, LLC (the applicant or NuScale), submitted Topical Report (TR)-0815-16497, Revision 0, "Safety Classification of Passive Nuclear Power Plant Electrical Systems." By letter dated February 7, 2017, NuScale submitted Revision 1 to TR-0815-16497 in proprietary (-P) and nonproprietary (-NP) versions (letter and -NP version available at ADAMS Accession No.

ML17048A459).

Section 1.1, "Purpose," of TR-0815-16497-NP, Revision 1, states the purpose of the submittal and describes the review and approval that the applicant seeks from the U.S. Nuclear Regulatory Commission (NRC or Commission) staff, as follows: The purpose of this topical report is to request Nuclear Regulatory Commission (NRC) review and approval of what are termed herein as "conditions of applicability," and the methodology and bases used in their development. The conditions of applicability comprise a set of passive reactor plant design and operational attributes that, if met in full by a reactor design or license applicant, justify the applicant's determination that none of the plant electrical systems fulfill functions that, per the regulatory definitions of "safety

-related" and "Class 1E," would warrant a Class 1E classification. The conditions of applicability are presented in Table 3-1 , "Conditions of applicability." This topical report also seeks NRC review and approval of augmented design, qualification, and quality assurance (QA) provisions that are an extension of the conditions of applicability (via Item II.1 of Table 3-1). The augmented provisions are described in Table 3-2. For reasons detailed in Section 3.2, these augmented design, qualification, and QA provisions would be applied as minimum requirements to electrical systems that have been determined to be nonsafety-related but yet are essential to the post

-accident monitoring of Type B and Type C variables. Provided the conditions of applicability are fully satisfied, the approved augmented provisions would represent an acceptable alternative to the portion of Regulatory Guide 1.97, Revision 4 (Reference 4.39), that specifies a Class 1E power source for instrumentation associated with Type B and Type C variables.

Based on its review of the TR, the NRC staff issued requests for additional information (RAIs) via letter dated October 7, 2016 (ADAMS Accession No.

ML16281A298); in particular, the RAIs addressed the direct current (dc) equipment and system , postaccident monitoring, and reactor coolant pressure boundary (RCPB) integrity and safe shutdown. In response to these RAIs , NuScale provided supplemental information in a letter dated December 5, 2016 (ADAMS Accession No.

ML16340D339).

2 2.0 Regulatory Evaluation The electric power systems for power plants include onsite electrical power systems providing alternating current (ac) power and dc power. Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations

," refers to safety-related electric al equipment as "Class 1E" equipment. As defined therein, the safety

-related or "Class 1E" classif icat ion is the safety classification of the electric al equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing a significant release of radioactive material to the environment.

As used in IEEE Std. 323-1974, Class 1E equipment includes appropriate interfaces.

If a reactor was designed so that no electrical equipment was "essential" such that it met the definition of Class 1E (i.e., the reactor plant design did not include safety-related equipment dependent on electric power

), then the design would not require Class 1E ac or dc power systems. Where no Class 1E equipment is used, the basic requirements for qualifying Class 1E equipment and interfaces, which are provided in IEEE Std. 323-1974, are inapplicable.

In TR-0815-16497, NuScale provided a method to justify that the plant electric power supplies need not be classified as Class 1E. In TR Section 3.1, "Methodology Used to Develop Conditions of Applicability," the applicant stated that "the application of augmented provisions is consistent with the process established in the NRC regulatory framework for 'special treatment' of nonsafety

-related SSCs that are determined to have risk

-significance

."

In TR Table 3-2, "Augmented Design, Qualification, and Quality Assurance Provisions," the applicant listed the regulatory requirements and guidance documents that a future passive plant applicant would need to apply or consider for the augmented design, qualification, and QA provisions of the non-Class 1E electrical systems

-termed the "highly reliable DC electrical system(s)"-for powering the postaccident monitoring instrumentation for Type B and Type C variables and for the plant emergency lighting systems.

The NRC staff evaluated the conditions of applicability in TR Table 3-1, "Conditions of Applicability,"

by first identifying the design-bas is information, as defined in Title 10 of the Code of Federal Regulations (10 CFR) 50.2, "Definitions." As defined in 10 CFR 50.2, "design basis" means that information that identifies the specific functions to be performed by an SSC of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.

The staff then ensured that Table 3-1 addressed these specific functions by the conditions of applicability. In accordance with 10 CFR 52.47(a)(3), an application for a design certification must include the design of the facility, including the following: (i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for water

-cooled nuclear power plants similar in design and location to plants for which construction permits have 3 previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units;

(ii) The design bases and the relation of the design bases to the principal design criteria; (iii) Information relative to materials of construction, general arrangement, and approximate dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with an adequate margin for safety; The staff's review considered if the design would meet the following minimum requirements in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," for principal design criteria even if no electrical equipment was classified as Class 1E: GDC 10, "Reactor Design," requires that the reactor core and associated coolant, control, and protection systems be provided with appropriate margin to assure that specified acceptable fuel design limits (SAFDL s) are not exceeded during any condition of normal operation, including the effect of anticipated operational occurrences (AOO s). GDC 13, "Instrumentation and Control," requires, in part, that the applicant provide instrumentation to monitor variables and systems over their anticipated ranges for normal operation, AOOs, and accident conditions as appropriate to assure adequate safety. GDC 15, "Reactor Coolant System Design," requires that the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs.

GDC 16, "Containment Design," requires that the reactor containment and associated systems shall be provided to establish an essentially leak

-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 19, "Control Room," requires, in part, that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including los s-of-coolant accidents (LOCAs). GDC 20, "Protection System Functions," requires, in part, that the protection system be designed to automatically initiate the operation of appropriate systems, including the reactivity control systems, to assure that SAFDLs are not exceeded as a result of AOOs.

4 GDC 26, "Reactivity Control System Redundancy and Capability," requires, in part, that the control rods be capable of reliably controlling reactivity changes to assure that SAFDLs are not exceeded under conditions of normal operation, including AOOs, and with appropriate margin for stuck rods.

GDC 27, "Combined Reactivity Control Systems Capability," requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system (ECCS), of reliably controlling reactivity changes to assure that the capability to cool the core is maintained under postulated accident conditions and with appropriate margin for stuck rods.

GDC 34, "Residual Heat Removal," requires, in part, that a residual heat removal system be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that SAFDLs and the design conditions of the RCPB are not exceeded.

GDC 35, "Emergency Core Cooling," requires, in part, that a system to provide abundant core cooling be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal

-water reaction is limited to negligible amounts.

GDC 38, "Containment Heat Removal," requires, in part, the provision of a system to remove heat from the reactor containment. The system safety function shall be to rapidly reduce, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and to maintain them at acceptably low levels. GDC 41, "Containment Atmosphere Cleanup," requires, in part, systems to control fission products, hydrogen, oxygen, and other substances that may be released into the reactor containment as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

GDC 50, "Containment Design Basis," requires, in part, that the reactor containment structure, including access openings, penetrations, and the containment heat removal system, shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. GDC 54, "Piping Systems Penetrating Containment," requires, in part, that piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities that have redundancy, reliability, and performance capabilities that reflect the importance to safety of isolating these piping systems.

5 GDC 55, "Reactor Coolant Pressure Boundary Penetrating Containment," requires, in part, that each line that is part of the RCPB and that penetrates primary reactor containment shall be provided with containment isolation valves.

GDC 56, "Primary Containment Isolation," requires, in part, that each line that connects directly to the containment atmosphere and penetrates the primary reactor containment shall be provided with containment isolation valves.

GDC 57, "Closed System Isolation Valves," requires each line that penetrate s primary reactor containment and is neither part of the RCPB nor connected directly to the containment atmosphere to have at least one containment isolation valve that shall be either automatic or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

GDC 61, "Fuel Storage and Handling and Radioactivity Control," requires, in part, that fuel storage and handling, radioactive waste, and other systems that may contain radioactivity be designed to assure adequate safety under normal and postulated accident conditions. This criterion specifies that such systems shall be designed to include appropriate containment, confinement, and filtering systems.

GDC 63, "Monitoring Fuel and Waste Storage

," requires, in part, appropriate systems in fuel storage and radioactive waste systems and handling areas to detect conditions that may cause a loss of residual heat removal capability and excessive radiation levels and to initiate appropriate safety actions.

GDC 64, "Monitoring Radioactive Releases," requires, in part, the means for monitoring the reactor containment atmosphere, spaces containing components for recirculation of LOCA fluids, effluent discharge paths, and the plant environs for radioactivity that may be released as a result of postulated accidents.

The NRC staff also determined that the following regulatory requirements and guidance documents are applicable to the review of this TR: In accordance with the requirements in 10 CFR 52.47(a)(8), an application for a design certification must include t he information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), 10 CFR 50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v). In turn, 10 CFR 50.34(f)(2) state s that to satisfy the requirements in 10 CFR 50.34(f)(2)(i)

-(xxviii), the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.

Those required actions under 10 CFR 50.34(f)(2) include the following

(viii) Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain accident source term radioactive materials without radiation exposures to any individual exceeding 5 rems to the whole body or 50 rems to the extremities.

Materials to be analyzed and 6 quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, radioiodines and cesium, and nonvolatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations.

(xvii) Provide instrumentation to measure, record and readout in the control room: (A) containment pressure, (B) containment water level, (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents at all potential, accident release points.

Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples.

(xix) Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.

(xx) Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A)

Level indicators are powered from vital buses; (B) motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety; and (C) electric power is provided from emergency power sources. (Applicable to PWR's only.) In accordance with the requirements in 10 CFR 52.47(a)(12), an application for a design certification must include an analysis and description of the equipment and systems for combustible gas control as required in 10 CFR 50.44, "Combustible Gas Control for Nuclear Power Reactors." In turn, 10 CFR 50.44 requires, in part, that an applicant must perform an analysis that demonstrates containment structural integrity. The analysis must address an accident that releases hydrogen generated from a 100-percent fuel clad-coolant reaction accompanied by the hydrogen burning. The applicant must demonst rate that systems necessary to ensure containment integrity are able to perform their function under these conditions.

In accordance with the requirements in 10 CFR 52.47(a)(4), an application for a desig n certification must include an analysis and evaluation of the design and performance of structures, systems, and components (SSCs) with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

The applicant shall perform analysis and evaluation of ECCS cooling performance and the need for high

-point vents following postulated LOCA in accordance with the requirements in 10 CFR 50.46 , "Acceptance Criteria for Emergency Core Cooling Systems for L ig ht-Water Nuclear Power Reactors," and 10 CFR 50.46a , "Acceptance Criteria for Reactor Coolant System Venting Systems." In turn, 10 CFR 50.46 sets for th 7 acceptance criteria for ECCS for light-water nuclear power reactors , and 10 CFR 50.46a sets forth acceptance criteria for reactor coolant system venting systems. In accordance with the requirements in 10 CFR 50.55a(h)(3), an application for design certification must meet the requirements for safety systems in IEEE Std. 603-1991 , "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations

," and the correction sheet dated January 30, 1995.

In accordance with the requirements in 10 CFR 52.47(a)(16), an application for a design certification must include a coping analysis, and any design features necessary to address station blackout, as required in 10 CFR 50.63 , "Loss of All Alternating Current Power." In turn, 10 CFR 50.63(a)(1) requires that each design for a light

-water-cooled nuclear power plant approved under a standard design certification must be able to withstand a station blackout for a specified duration and recover from a station blackout, as defined in 10 CFR 50.2. The specified station blackout duration shall be based on the following factors:

the redundancy of the onsite emergency ac power sources the reliability of the onsite emergency ac power sources the expected frequency of loss of offsite power the probable time needed to restore offsite power The requirements in 10 CFR 50.63(a)(2) state that the reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration.

The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis.

Applicants are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review.

In accordance with the requirements in 10 CFR 52.47(a)(2), an application for standard design certification for nuclear power reactors shall present a safety analysis of the facility design in terms of site parameters postulated for the design.

Specifically, 10 CFR 52.47(a)(2)(iv) requires that an analysis of the radiological consequences of postulated accidents include the following: The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together 8 with applicable postulated site parameters, including site meteorology, to evaluate the offsite radiological consequences. The evaluation must determine that:

(A) An individual located at any point on the boundary of the exclusion area for any 2

-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE);

(B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE. Applications for combined licenses (COLs), construction permits, and operating licenses that reference the subject TR have similar requirements to evaluate the radiological consequences of postulated accidents in accordance with 10 CFR 52.79(a)(1)(vi) and 10 CFR 50.34(a)(1). The siting requirements in 10 CFR 100.21, "Non

-Seismic Site Criteria," also reference the criteria in 10 CFR 50.34(a)(1).

In accordance with the requirements in 10 CFR 52.47(a)(2)(i ii), as part of its review of an application for a design certification, the Commission will consider the extent to which the reactor incorporates unique, unusual, or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials.

As discussed in 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities,"Section VI, "Emergency Response Data System," the Emergency Response Data System (ERDS) is a direct near real

-time electronic data link between the applicant's onsite computer system and the NRC Operations Center that provides for the automated transmission of a limited data set of selected parameters. While it is recognized that ERDS is not a safety system, it is conceivable that an applicant's ERDS interface could communicate with a safety system, and thus would require appropriate isolation devices at these interfaces.

Section VI.2.a.(i) of Appendix E requires, for pressurized

-water reactors (PWRs), that the selected plant parameters to be transmitted include those from radiation monitoring systems (i.e., reactor coolant radioactivity, containment radiation level, condenser air removal radiation level, effluent radiation monitors, and process radiation monitor levels). Regulatory Guide (RG) 1.97, Revision 4, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants," issued June 2006, describes a method that the NRC staff considers acceptable for use in complying with the agency's regulations with respect to satisfying criteria for accident monitoring instrumentation in nuclear power plants.

Specifically, the method described RG 1.97 relates to GDC 13, 19, and 64. RG 1.97 endorses (with certain clarifying regulatory positions specified in Section C of the RG) IEEE Std. 497-2002, "IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations."

9 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports: LWR Edition," Branch Technical Position 7-10, "Guidance on Application of Regulatory Guide 1.97," Revision 6, issued August 2016, provides additional guidelines for reviewing an applicant's accident monitoring instrumentation.

SECY-94-084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-safety Systems in Passive Plant Designs," dated March 28, 1994 (ADAMS Accession No.

ML003708068), presented the Commission with recommended positions pertaining to policy and technical issues affecting passive advanced light

-water reactor (ALWR) designs and requested that the Commission approve certain staff positions stated in the SECY, including the Electric Power Research Institute

's proposed alternative to the cold

-shutdown condition called for by RG 1.139, "Guidance for Residual Heat Removal," as a safe, stable condition that the passive decay heat removal systems must be capable of achieving and maintaining following non

-LOCA events.

This recommendation was predicated on an acceptable passive safety system performance and an acceptable resolution of the issue of regulatory treatment of nonsafety systems.

In its staff requirements memorandum (SRM) on SECY 084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems

," and COMSECY-94-024, "Implementation of Design Certification and Light

-Water Reactor Design Issues," dated June 30, 1994, the Commission, among other things, approved the staff's recommendation on this item.

In doing so, the Commission stated that, with respect to the 72

-hour capacity of the passive residual heat removal system water pool, the requirements for replenishing the water in the pool should be based on design-specific attributes, and the applicant

's justification of these requirements should not be based solely on the 72

-hour criterion of the utility requirement document.

Further, the Commission stated that the staff should be receptive to arguments for longer periods if technically justified.

On May 22, 1995, the staff issued SECY-95-132, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs" (ADAMS Accession No.

ML003708005), in response to SRM-SECY-94-084 and present ed a corresponding revision of SECY 084 for Commission review and approval.

On June 28, 1995, the Commission approved the staff's recommendations in SECY 132 (ADAMS Accession No.

ML003708019).

3.0 Staff

Evaluation TR Section 1.2, "Scope," gives the scope of review specific to the safety classification of plant electrical systems for which the conditions of applicability and augmented provisions apply, as follows: offsite and onsite ac electrical power systems onsite dc electrical power systems. In the TR, NuScale stated that the above scope does not include instrumentation and control equipment and circuits, which include both Class 1E and non

-Class 1E systems, that serve to monitor and control power to and operation of safety

-related and nonsafety

-related loads.

10 The TR contains four appendi c es that describe the methodology and procedures to be applied to an example power system design to ensure that a dc power system design can be "highly reliable":

(1) Appendix A , "Example Overview of Electrical Systems and Instrumentation and Control (I&C) Systems Design," gives an overall description of an onsite power system that could serve a passive plant design that meets the conditions of applicability.

In addition, Appendix A includes a set of typical one

-line diagrams to facilitate an overall understanding of the concepts as applied to a passive plant electrical system.

(2) Appendix B , "Example Safety Classification Assessment for Electrical Systems," describes how a hypothetical complete loss of all electric power (both ac and dc) would affect the various safety functions and explains how the applicant can satisfy the attributes of the conditions of applicability.

However, Appendix B does not describe how the requirements of 10 CFR Part 50, Appendix E, Section VI.2.a.(i);

10 CFR 50.34(f)(2)(viii); or 10 CFR 50.34(f)(2)(xvii) would be met.

(3) Appendix C , "Example Failure Modes and Effects Analysis-Highly Reliable DC Power System," provides an example failure modes and effects analysis of the example onsite dc power system described in Appendix A. The effects of failure modes and mechanisms for components in the example analysis establish that no single failure exists that could prevent safety

-related functions from being achieved and maintained.

(4) Appendix D , "Example Safety Analysis Results," provides example safety analysis results of a passive plant that has the design attributes described in Appendices A and B. The analysis shows that, in each postulated design

-basis event (DBE) analyzed, none of the systems credited for mitigating the event requires electric power or operator action.

TR Section 1.2 states the following:

The information provided in the appendices is provided to facilitate: (1) the NRC's review of the conditions of applicability and augmented provisions for which approval is sought; and (2) an understanding of how this topical report would be implemented by future applicants (including NuScale).

As part of the scope of this topical report, NuScale is not seeking NRC approval of the information in the appendices.

Information is provided in this report to demonstrate applicability of the methodology and to aid the reader's understanding of the application of these methodologies.

NuScale TR further stated that its design certification application (DCA) will present the final design information and will confirm that the final design meets the conditions of applicability described in TR Table 3-1 , which lists the attributes to be satisfied as conditions of applicability.

The TR Table 3-1 has two sections:

(1)Section I contains the specific conditions that, if fully met, would adequately justify that no Class 1E electrical supply systems (power sources) are required.

11 (2)Section II contains additional conditions to be applied (after meeting Section I). TR Table 3-1, S ection II, requires augmented design, qualification, and QA provisions. The provisions in Table 3-2 are the minimum requirements to be applied to non

-Class 1E electrical systems (termed as "highly reliabl e DC electrical system(s)") that will be used to power postaccident monitoring instrumentation for Type B and Type C variables and to power the plant emergency lighting system. I f a passive nuclear plant can meet all the conditions listed in Table 3-1 without the need for any electrical power, Class 1E ac or dc power supply systems may not be nec essary. This is subject to satisfying the capability T he NRC staff review of the information in the appendices does not constitute approval of the information in the appendi ces. T herefore, the NRC staff limited it s review to the main body of the TR and focused on the design criteria considered in the conditions of applicability, not an actual design.

Concept of "High ly Reliable" Non-Class 1E Direct Curren t System With regard to a fully non-Class 1E dc power system for a completely passive nuclear power plant design, the NRC staff was concerned whether the dc power system would have high reliability. More specifically, the NRC staff was concerned that the valve-regulated, lead-acid (VRLA) battery life could be seriously and suddenly reduced by exposure to prolonged periods of high temperatures, the magnitude and frequency of discharge cycles, or overcharging. Th e NRC staff devised a three-pronged review approach (i.e., performance, QA, and quantification) to determine the relative reliability of the conceptual dc power system design (presented in TR Appendix A) compared to a Class 1E dc power system.

To date, conventional large light-water nuclear power plants have not used VRLA bat teries for onsite power. Therefore, the NRC staff requested information on battery life, QA, performance, qualification, and reliability. RAI 08.03.02-01 In a letter dated December 5, 2016 (ADAMS Accession N

o. ML16340D339), NuScale acknowledge d the NRC staff's concerns with VRLA battery life and stated that these effects can be mitigated by following the recommendations in IEEE Std. 1187-2013, "IEEE Recommended Practice for Installation Design and Installation of Valve

-Regulated Lead

-Acid Batteries for Stationary Applications

," and IEEE Std.

1188-2005 (R2010), "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Valve

-Regulated Lead

-Acid (VRLA) Batteries for Stationary Applications

," as noted in TR Table 3-2. Additionally, IEEE Std.

1187-2013 refers to IEEE Std. 1491-2012, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications," and IEEE Std.

1635-2012, "IEEE/ASHRAE Guide for the Ventilation and Thermal Management of Batteries for Stationary Applications." In addition to the use of the industry standard procedures mentioned above for design, testing, and implementation of the VRLA battery

-powered dc system, the applicant stated the following:

12 The backup power supply system delivers backup power to heating, ventilation, and air conditioning systems serving the battery and associated charger rooms to avoid prolonged periods of high ambient temperature. For design consideration for magnitude and frequency of discharge cycle related monitoring, the applicant will follow the guidance in IEEE Std.

1187-2013, IEEE Std. 1188-2005, and specifically IEEE Std.

1491-2012, which provides criteria to detect and monitor a battery for degradation. Following the guidance in IEEE Std.

1187-2013, as supplemented by IEEE Std. 1491-2012, provides reasonable assurance that the VRLA batteries will not be overcharged and that instances of potential overcharging will be detected before degrading a battery to a point where it is not able to perform its intended function.

The electrical power system presented in TR Appendix A depicts an onsite power system design with no Class 1E power sources, assuming the reactor design does not require any safety-related electrical loads to support the safety analyses.

The NRC staff reviewed the RAI response and determined that the use of VRLA batteries in a nonsafety dc power system design for a passive nuclear power plant, construction and monitoring will follow the guidance in IEEE Std. 1187-2013 and IEEE Std. 1188-2005, as supplemented by IEEE Std.

1491-2012 and IEEE Std. 1635-2012. These IEEE standards provide widely established industry guidance for design, testing, and performance of VRLA batteries. The NRC staff determined that, based on the IEEE standards mentioned above, the design will give reasonable assurance that a dc power system that uses a VRLA battery will not be exposed to prolonged period s of high temperatures, will be monitored for potential overcharging, and will be monitored for magnitude and frequency of discharge cycles that may degrade the battery performance.

For the reasons discussed above, t he NRC staff concludes that, for a nonsafety dc system that uses VRLA batteries, the applicant's response gives reasonable assurance that the dc system will be monitored for degradation and the use of VRLA batteries will not adversely affect the dc system's intended function.

The NRC staff asked the applicant to include its response to RAI 08.03.02-01 in the next revision of the TR. I n Revision 1 to the TR, the applicant included the applicable year for the following IEEE standards as requested in the RAI:

IEEE Std. 1491-2012 and IEEE Std. 1635-2012. This action satisfies the NRC staff's request.

RAI 08.03.02-02 In TR Table 3-2, NuScale stated that a graded QA program will be applied to the dc electrical system that will meet or exceed the augmented QA guidance in Appendix A, "Quality Assurance Guidance for Non

-Safety Systems and Equipment," to RG 1.155, "Station Blackout." The NRC staff asked NuScale to describe the proposed QA program in sufficient detail to enable the NRC staff to verify whether it meets or exceeds the guidance in RG 1.155.

13 In its December 5, 2016, response to RAI 08.03.02-02, NuScale stated that a COL applicant that references TR

-0815-16487 will be required to follow the guidance in RG 1.155, Appendix A. The NRC staff finds NuScale's response reasonable. The NRC staff has placed Condition 4.1 in Section 4.0 of this safety evaluation to ensure that all future applicants that reference TR

-0815-16497 address the guidance in RG 1.155, Appendix A, in sufficient detail to verify whether the relevant QA program would meet or exceed the guidance in RG 1.155. RAI 08.03.02-03 In TR Table 3-2, under "Batteries," NuScale stated that the VRLA batteries have augmented design, QA, and qualification provisions. The NRC staff asked NuScale to describe the methods and processes that a passive reactor nuclear power plant will use to verify that VRLA batteries will perform their intended functions during normal operation, AOOs, and postulated DBEs.

In its response dated December 5, 2016 (ADAMS Accession No.

ML16340D339), NuScale stated that the VRLA batteries used in a passive reactor nuclear power plant design are not credited for use in mitigating the consequences of postulated DBEs. NuScale also stated that an applicant using this TR shall implement a testing and monitoring program, as described in IEEE Std. 1188-2005 and IEEE Std.

1491-2012, to ensure that VRLA batteries will perform their intended functions when called upon. These standards provide for a wide variety of operating parameters to be monitored on a continuous basis, including cell

-specific parameters. Furthermore, NuScale stated that applicants would be required to environmentally qualify their VRLA batteries in accordance with IEEE Std.

323-1974, as appropriate, and IEEE Std. 323-2003, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations," and to seismically qualify their batteries in accordance with IEEE Std. 344-2004, "IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations,"

as appropriate, to give further assurance that the batteries will perform their intended functions.

The NRC staff also asked NuScale to identify the industry standards or applicable references that will be used for verification purposes. NuScale identified the following industry standards:

IEEE Std. 323-1974 , as endorsed by RG 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants,"

for harsh environments IEEE Std. 323-2003 for mild environments IEEE Std. 344-2004 , as endorsed by RG 1.100, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants" IEEE Std. 1188-2005 IEEE Std. 1491-2012 14 The NRC staff reviewed the applicant's response to RAI 08.03.02-03 and determined that the design of the VRLA batteries used as a non

-Class 1E dc power source in a passive reactor nuclear power plant design, in accordance with the widely accepted indust ry practices IEEE Std. 1188-2005 and IEEE Std. 1491-2012 for testing and monitoring; IEEE Std. 323-1974, as appropriate, and IEEE Std. 323-2003, as appropriate, for environmental qualification; and IEEE Std. 344-2004 for seismic qualification provide reasonable assurance that the VRLA batteries will perform their intended functions.

The NRC staff concludes that NuScale's response is acceptable with regard to the methods and processes used to verify that the VRLA batteries will perform as intended.

The TR states that the VRLA batteries will be seismic Category 1; therefore, an applicant using the TR shall provide a qualification testing plan that includes an environmental and seismic qualification, and also a technical functional requirement for the VRLA batteries to provide reasonable assurance that VRLA batteries will perform their intended functions. For this reason, the NRC staff has established Condition 4.2 on the TR for the applicant to confirm that the VRLA batteries and their structures are seismic Category 1. To give reasonable assurance that the VRLA batteries will perform as intended, the applicant that references the TR must provide a COL action item to support that the VRLA batteries and their structures are seismic Category 1. RAI 08.03.02

-04 In the TR, NuScale described its dc power system as "

highly reliable" and substantially equal in reliability to that of an analogous Class 1E dc power system. However, the TR did not fully justify these statements. Therefore, to complete its review , the NRC staff asked the applicant to provide additional quantitative information. Specifically, the NRC staff asked the applicant to describe the methodology that it will use to compare the highly reliable dc system that it will describe in its DCA to a Class 1E dc power system to show that the highly reliable dc system is substantially equal in reliability to a typical Class 1E dc power system.

NuScale provided a two

-part response. The first part describes the methodology in the TR that design certification applicants would use to perform a quantitative analysis. Th is methodology comprises the following five steps needed to compare the reliability of the highly reliable dc system to that of a typical Class 1E dc power system: (1) (2) (3) (4) 15 (5) T he second part of NuScale's response gave the results of its comparative analysis using the above methodology. NuScale indicated that its results were favorable in that the augmented non-Class 1E design indicated a reliability greater than that of the Class 1E design. In its response, NuScale further concluded that amending the TR to include the methodology presented is not necessary. NuScale and the N RC staff held a conference call on January 6, 2017, to address the RAIs. First, the staff asked for clarification on whether NuScale's referenced probabilistic risk analysis (PRA) model included common

-cause failures among each of the two

-battery-in-parallel configurations. NuScale stated that the model included common-cause failure of the two

-battery configurations. The concern was that any battery operating in parallel could experience certain common-cause events. Any further questions on PRA methodology would be part of the PRA review of the referencing DCA or COL application. Second, the NRC staff requested clarification about the statement at the end of the response that the response does not require a revision to the licensing document (i.

e., TR-0815-16497). The NRC staff questioned this statement because TR-0815-16497 is a methodology document and the response to RAI 08.03.02-04 provides additional methodology necessary for use of the TR by any applicant referencing it. NuScale added this methodology to Table 3-1,Section II, of Revision 1 to the TR. This action satisfies the NRC staff's request.

Based on the review of this response, the NRC staff concludes that the five

-step process outlined in the applicant's response provides an acceptable approach for demonstrating the relative reliability of a non

-Class 1E system with that of an analogous Class 1E system.

3.1 Postaccident

Monitoring

The primary purpose of postaccident monitoring instrumentation is to display plant variables that provide information required by the control room operator during and after an accident.

GDC 13, GDC 19, GDC 64, 10 CFR 50.34(f)(2)(xix), 10 CFR 50.34(f)(2)(xx), and 10 CFR 50.55a(h) contain regulatory requirements governing postaccident monitoring instrumentation. The NRC provides the primary guidance for implementing these regulatory requirements in RG 1.97, which describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation for monitoring plant variables and systems during and after an accident. RG 1.97, which endorses IEEE Std.

497-2002, with certain clarifying regulatory positions specified in Section C of RG 1.97, specifies that a Class 1E electrical system should be provided to supply the instrumentation that monitors Type A, B, and C variables under postaccident conditions. Under 10 CFR 50.34(f)(2)(xx), the NRC requires that electric power for pressurizer level indicators must be powered by vital buses.

RG 1.97 defines Type A, B, and C variables as follows:

16 Type A variables provide the primary information required to allow main control room operators to take manual actions for which no automatic control is provided.

Type B variables provide primary information to the control room operators to assess the plant safety functions.

Type C variables provide primary information to the control room operators to indicate the potential for breach or the actual breach of fission product barriers (e.g., fuel cladding, RCPB, and containment pressure boundary).

During its review, the NRC staff considered whether the safety system design to provide accident monitoring instrumentation would require instrumentation to be powered by a Class 1E electrical system for Type B and C variables.

IEEE Std. 603-1991, Clause 5.8.1, "Displays for Manually Controlled Actions," specifies that monitoring instrumentation be part of the safety systems and meet the requirements of IEE E Std. 497-2002. For monitoring instrumentation used for these operations, IEEE Std.

603-1991 and IEEE Std.

497-2002 specify a Class 1E electrical power supply. T he NRC staff's evaluation considered the following:

Regulatory requirements in GDC 13, 19, and 64 are applicable to postulated DBEs and do not specify a Class 1E electrical power supply. Therefore, a Class 1E electrical power supply is not required to meet GDC 13, 19, and 64. In accordance with the requirements in 10 CFR 52.47(a)(8), an application for a design certification must include the information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements as stated in 10 CFR 50.34(f), except for 10 CFR 50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v).

The requirements in 10 CFR 50.34(f)(2)(xix) call for the design to provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage. This includes core damage that may be more extensive than a postulated DBE.

Finally, 10 CFR 50.34(f)(2)(xix) does not specify the quality of the electrical supply; therefore, a Class 1E electrical power supply is not required to meet 10 CFR 50.34(f)(2)(xix).

In accordance with the requirements in 10 CFR 50.34(f)(2)(xx), which are applicable to PWRs only, the design must provide power supplies for pressurizer relief valves, block valves, and level indicators such that (1) level indicators are powered from vital buses, (2) motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to

safety, and (3) electric power is provided from emergency power sources.

On its face, 17 NUREG- 0737, "Clarification of TMI Action Requirements," issued November 1980, states that the instrument channels for pressurizer level indication instrument channels shall be powered from the vital instrument buses and does not specify a Class 1E electrical power supply requirement; therefore, a Class 1E electrical power supply is not required to meet 10 CFR 50.34(f)(2)(xix). Clause 5.8.2 of IEEE Std. 603-1991 states, in part, that the display instrumentation provided for safety-system status indication need not be part of the safety systems

therefore, a Class 1E electrical power supply is not required to meet Clause 5.8.2 of IEEE Std. 603-1991. Type B and Type C accident monitoring instrumentation is required to perform its intended function under postulated accident conditions. As such, the reliability of the electrical power supply for these instruments should be substantially similar to that of a Class 1E electrical system (see Section 3.0 of this safety evaluation). In TR Section 3.2.1, the applicant provided an alternative to RG 1.97 that uses a highly reliable dc power system in lieu of a Class 1E electrical system to supply electrical power to the postaccident monitoring instrumentation. When performing this review, the NRC staff considered the electrical system reliability of the highly reliable dc electrical system. The NRC staff established a three-pronged approach to establish whether the highly reliable dc electrical system provides a substantially equal reliability to that of a Class 1E design. The three-pronged approach consisted of (1) evaluation of the augmented design, qualification, and QA provisions, (2) consideration of the rigor of the highly reliable dc power system as demonstrated by the failure modes and effect analysis, and (3) quantification via fault tree analysis to compare the NuScale design with an approved passive PWR dc system design. Section 3.0 of this safety evaluation evaluates the electrical system reliability of the highly reliable dc power system.

Based on its evaluation of the electrical system reliability, the staff concluded that the highly reliable dc electrical system provides a substantially equal reliability to that of a Class 1E design; thus, the dc electrical system provides additional assurance that postaccident monitoring capability is maintained during and following a DBE.

Based on the NRC s taff's review of the TR and the regulatory requirements governing accident monitoring instrumentation, the staff found that the augmented design, qualification, and QA provisions of the power sources for Type B and Type C variables represent an acceptable alternative to the guidance in RG 1.97. , th e staff has established Condition 4.3 (see Section 4.0 of this safety evaluation) for the applicant s referencing this safety evaluation to confirm that operator actions are not necessary to ensure safety

-related functions for any postulated DBE (i.e., the design does not include Type A variables as defined in IEEE Std.

497-2002, as modified in RG 1.97, Regulatory Position C.4). Spent Fuel Pool Considerations

The spent fuel pool (SFP) has the safety function of maintaining the spent fuel assemblies in a safe and subcritical array during all credible storage conditions. GDC 63 for spent fuel storage facilities requires monitoring systems to (1) detect conditions that may cause the loss of residual 18 heat removal capability and excessive radiation levels and (2) indicate when to take action to initiate appropriate safety actions.

In TR Appendix B, Section B.2.2, "Fuel Assembly Cooling

-Spent Fuel and Module Core Refueling," the applicant des c ribed I n TR Table 3-1, Conditions of Applicability 3 and 4 specify that for the TR to be applicable to a design, the applicant must demonstrate the following: T he NRC staff determined that Conditions of Applicability 3 and 4, as stated above, are consistent with the staff guidance in NUREG-0800, Section 19.3, "Regulatory Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors," and, therefore, if a design met these conditions, Class 1E power would not be required for monitoring SFP conditions.

3.2 Safe Shutdown, Core Cooling, and Reactor Coolant Pressure Boundary Integrity

The NRC staff used the review guidance in the NUREG-0800 to identify the Commission's regulations associated with safe shutdown, core cooling, and RCPB integrity. In accordance with 10 CFR 52.47(a)(3)(i), the staff identified, as minimum requirements, GDC 10, 15, 20, 26, 27, and 34 and 10 CFR 50.46 as associated with safety

-related SSCs (in accordance with the definition in 10 CFR 50.2) that need to be addressed by the conditions of applicability in TR Table 3-1. Condition of Applicability I.1.a, and Condition of Applicability I.1.c., r equire, in part, T he N RC staff finds these requirements to be consistent with GDC 20. Accordingly, the NRC staff finds that Conditions of Applicability I.1.a and I.1.c are necessary and sufficient for determining that no Class 1E power is required to satisfy GDC 20. Condition of Applicability I.1.b states, 19 T he NRC staff describes safe-shutdown requirements in SECY 084. In SRM-SECY-94-084, the Commission approved the staff's recommendation on safe

-shutdown requirements.

SECY-94-084 clarifies the conditions that constitute a safe

-shutdown condition as reactor subcriticality, decay heat removal, and radioactive material containment. Additionally, SECY-94-084 states that an appropriate safety analysis can be used to demonstrate passive system capabilities to bring the plant to a safe, stable condition and to maintain this condition. The staff's views on safe shutdown were not changed in SRM

-SECY-95-132 (updating the Commission on matters in SECY 084). TR Appendices B and D provide clarifying examples to illustrate how the conditions of applicability can be demonstrated. The examples did not include a quantitative safety analysis to demonstrate the ability to insert sufficient negative reactivity during and following a DBE to achieve and maintain safe shutdown. This omission caused the NRC staff to question the interpretation of safe shutdown as applied to Condition of Applicability I.1.b. Accordingly, the NRC staff issued RAI 08.03.02-05, dated October 7, 2016 (ADAMS Accession No. ML16281A298), asking the applicant to (1) specify the criteria that constitute a safe shutdown as applied to Condition of Applicability I.1.b, and (2) describe how a future applicant for a passive plant will demonstrate that electric power is not necessary to achieve and maintain a safe shutdown for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In its December 5, 2016, response (ADAMS Accession No.

ML16340D339), NuScale stated that the criteria that constitute a safe shutdown are subcriticality and decay heat removal in order to maintain fuel clad integrity (radioactive material containment).

The NRC staff finds this response acceptable because it is more restrictive than the criteria in SECY 084.

The applicant's response to RAI 08.03.02-05 further discussed the following approach to demonstrating Condition of Applicability I.1.b: -an applicant will evaluate the reactivity control systems to ensure sufficient shutdown function capability and evaluate the decay heat removal system to ensure sufficient heat removal capability. To ensure that safe shutdown capability is sufficient to address the safety issue of heat removal reliability, a probabilistic risk assessment is used to ensure that the reliability of systems used to achieve and maintain safe shutdown supports conformance to the commission's safety goal guidelines.

The applicant further explained that safety analyses of DBEs (as typically presented in Chapter 15 of a final safety analysis report (FSAR)) may not be suitable for demonstrating the ability to achieve and maintain a safe shutdown following a DBE. Specifically, the applicant's response stated the following: Conservative assumptions are applied to Chapter 15 safety analysis of DBEs appropriate for the intended purpose of ensuring appropriate margins to protect fuel integrity and core coolability.

Although these safety analyses can be used to demonstrate adequate shutdown capability per SECY 084, application of the same conservative assumptions may lead to excessive margin with respect to shutdown capability.

20 The NRC staff previously communicated positions on shutdown margin during and following DBEs in letters discussing GDC 26 and 27, dated December 5, 2016 (ADAMS Accession No. ML16292A589), and September 8, 2016 (ADAMS Accession No.

ML16116A083), respectively. These letters clarify that shutting down the reactor and maintaining a subcritical reactor are safety functions considered in GDC 26 and 27, both of which require margin for malfunctions such as stuck rods. In the letter addressing GDC 27, the NRC staff stated the following: Criterion 27 requires that the reactor be reliably controlled and that the reactor achieve and maintain a safe, stable condition, including subcriticality beyond the short term, using only safety related equipment following a postulated accident with margin for stuck rods.

Based on the shutdown margin requirements of GDC 26 and 27, the NRC staff established Condition 4.5 to require a demonstration or appropriate justification of shutdown margin. Based on the applicant's criteria for safe shutdown and pursuant to Condition 4.5, the NRC staff find s that Condition of Applicability I.1.b is necessary and sufficient for determining that no Class 1E power is required to satisfy GDC 26 and 27.

Condition of Applicability I.1.c, i s a high-level requirement associated with core cooling. GDC 10, 34, and 35 and 10 CFR 50.46 are design requirements associated with safety

-related SSCs that perform core cooling functions. In accordance with the requirements in 10 CFR 50.34, "Contents of Applications; Technical Information"; 10 CFR 52.47, "Contents of Applications; Technical Information"; and 10 CFR 52.79, "Contents of Applications; Technical Information in Final Safety Analysis Report," applicants are required to provide a description and analysis of the safety-related SSCs credited to perform core cooling functions, with emphasis upon performance requirements. The information provided by an applicant under these regulations must be sufficient to demonstrate complian ce with GDC 10, 34, and 35 and 10 CFR 50.46. Additionally, an applicant referencing the TR is required to perform these evaluations to show that safety functions will be accomplished in the absence of electrical power to demonstrate compliance with Condition of Applicability I.1.c. Accordingly, the NRC staff finds that Condition of Applicability I.1.c is necessary and sufficient for determining that Class 1E power is not required to satisfy GDC 10, 34, and 35 and 10 CFR 50.46. Condition of Applicability I.1.g states, T his statement supports Condition of Applicability I.1, which states, T R Appendices B and D provide clarifying examples to illustrate how the conditions of applicability can be demonstrated.

The example safety analysis in Appendix D shows that t he example passive plant response to an AOO includes establishing a direct coolant flowpath between the reactor core and the containment, thereby removing a fission product barrier. This caused the NRC staff to question whether Condition of Applicability I.1.g is sufficient for demonstrating RCPB integrity.

Accordingly, the NRC staff issued RAI 08.03.02-06, dated October 7, 2016 (ADAMS Accession No.

ML16281A298), asking the applicant to (1) specify the criteria that constitute RCPB integrity as applied to Condition of Applicability I.1, and (2) explain why the removal of a fission product barrier during an AOO is not considered an event escalation.

21 In its December 5, 2016, response (ADAMS Accession No.

ML16340D339), NuScale stated that a loss of RCPB integrity involves a mechanical failure in an RCPB component, but it does not include the opening of a valve. The applicant further stated that considering the RCPB to be lost when a valve opens is problematic because (1) it would preclude advanced designs that offer improvements in safety by relying on valves to depressurize the reactor coolant system for safe shutdown, (2) it is not consistent with the licensing basis for PWRs and boiling

-water reactors (BWRs), as these designs rely on safety relief valves for overpressure protection, and (3) the GDC address maintaining structural integrity of RCPB components rather than preventing the opening of valves to allow fluid to pass into or out of the RCPB.

Additionally, the applicant stated that opening a valve to depressurize the reactor coolant system and establish long

-term cooling is not considered a removal of a fission product barrier, and thus not event escalation, because the functions of the reactor coolant system barrier are not lost. The applicant further stated that events that do not result in unacceptable consequences or significantly increase the risk for radiological release do not challenge the intent of the nonescalation criterion specified in NUREG-0800, Section 15.0, "Introduction

-Transient and Accident Analyses

."

The NRC staff's evaluation of the applicant's response considered the examples from operating PWRs and BWRs. The applicant's response included examples in which valves connected to the reactor coolant system opened and allowed fluid to pass through the RCPB and included the opening of safety relief valves, shutdown cooling , and the reactor core isolation cooling system in BWRs. The NRC staff finds these examples to differ from the scenario that was the basis for RAI 08.03.02-06. In particular, the staff identifies that a rapid discharge of reactor coolant directly to the containment atmosphere, in response to an AOO, can result in significant pressurization of the containment, which is required to retain coolant and establish a return pat h to the reactor pressure vessel.

The AOO scenario in TR Appendix D appears to rely on the containment to retain the reactor coolant necessary to ensure fuel cladding integrity during an AOO. Because an AOO, by definition, is expected to occur one or more times during the life of the nuclear power plant, the NRC staff is concerned that such reliance upon the containment may not be consistent with the underlying defense

-in-depth purpose of GDC 15 , which expects the RCPB to remain available as a fission product barrier during AOOs. Accordingly, the NRC staff established Condition 4.4 on the TR to address reliability requirements for the systems necessary to retain reactor coolant within the RCPB. Condition 4.4 requi res a probabilistic determination of the expected frequency of ECCS actuation during AOO mitigation (e.g., dc power system failure th at causes ECCS actuation, ECCS pilot valve failure, spurious ECCS actuation). Opening of the ECCS valves dur ing normal, planned plant operations, including recovery from an AOO, is acceptable once a safe, stable state has been established.

Based on the overpressure protection of the RCPB and pursuant to Condition 4.4, the NRC finds that Condition of Applicability I.1.g is necessary and sufficient for determining that Class 1E power is not required to satisfy GDC 15. 3.3 Containment Isolation TR Condition of Applicability I.1.d specifies that for T h e provisions in GDC 54 , 55, 56, and 57 in part require containment isolation capabilities. Based on consideration of the relevant GDC above, the staff determined that a plant design that is able to 22 satisfy Condition I.1.d should be able to meet the minimum design requi rements in GDC 54, 55, 56, and 57. The NRC staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to achieve the containment isolation function.

3.4 Containment

Integrity TR Condition of Applicability I.1.e specifies that for T he provisions in GDC 16, 38, 41, and 50 in part require that the containment safety function can be achieved and maintained during DBEs. The provisions in 10 CFR 50.44 address the control of combustible gases in the containment. Based on consideration of the relevant GDC and 10 CFR 50.44, the staff determined that a plant design that is able to satisfy Condition of Applicability I.1.e should be able to meet the minimum design requirements in GDC 16, 38, 41, and 50 and 10 CFR 50.44. The staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to assure that containment integrity is achieved and maintained. 3.5 Fission Product Control TR Condition of Applic a bility I.1.f specifies that for T he provisions in GDC 41 in part require systems to control fission products. Based on consideration of the relevant GDC and applicable guideline exposure requirements, the staff determined that a plant design that is able to satisfy Condition of Applicability I.1.f should be able to meet the minimum design requi rements in GDC 41 and applicable guideline exposure requirements. The staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to satisfy GDC 41 and the applicable guideline exposures in 10 CFR 100.21, 10 CFR 50.34(a)(1)(ii)(D), and 10 CFR 52.47(a)(2)(iv).

3.6 Control

Room Habitability TR Condition of Applicability I.5 specifies that electrical power is not necessary T he provisions in GDC 19 in part require that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs. Based on consideration of the relevant GDC, the staff determined that a plant design that is able to satisfy Condition of Applicability I.5 should be able to meet the minimum design requirements in GDC 19. The staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to satisfy GDC 19.

3.7 Cooling

for Building Areas Containing Safety

-Related Equipment TR Condition of Applicability I.6 specifies that

23 T h e provisions in 10 CFR 50.63 in part require that the reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. Based on consideration of the 10 CFR 50.63 requirement, the staff determined that a plant design that is able to satisfy Condition of Applicability I.5 should be able to meet the requirements in 10 CFR 50.63. The staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to satisfy 10 CFR 50.63.

3.8 Building

Ventilation TR Condition of Applicability I.7 specifies that T he provisions in GDC 61 in part require that fuel storage and handling, radioactive waste, and other systems that may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. Based on consideration of the relevant GDC and the applicable guideline exposure requirements, the staff determined that a plant design that is able to satisfy Condition of Applicability I.7 should be able to meet the minimum design requi rements in GDC 61 and applicable guideline exposure requirements.

The NRC staff finds that the condition is necessary to enable the staff to determine that Class 1E electrical power is not required to satisfy GDC 61 and the applicable guideline exposures in 10 CFR 100.21, 10 CFR 50.34(a)(1)(ii)(D), and 10 CFR 52.47(a)(2)(iv).

3.9 Emergency

Lighting TR Section 3.2.2

, "Emergency Lighting,"

state s that the highly reliable dc electrical system provides power to portions of the emergency lighting system, and that the emergency lighting system is classified as non

-Class 1E. Additionally, TR Condition of Applicability II.3 (Section II of Table 3-1) s pecifies that the applicant's emergency lighting capability T he NRC staff finds that TR Condition of Applicability II.3 is consistent with the NRC staff's guidance on the classification of the emergency lighting system as non

-Class 1E and, therefore, is acceptable. 4.0 Limitations and Conditions In its letter dated July 26, 2017 (ADAMS Accession No.

ML17205A380), the Advisory Committee on Reactor Safeguards indicated that TR

-0815-16497-P, Revision 1, is acceptable for use only as a reference document for the NuScale plant electrical systems design subject to the staff limitations and conditions.

The staff responded to the committee on September 11, 2017 (ADAMS Accession No.

ML17221A058), agreeing with its recommendation.

Therefore, the NRC staff's conclusions on this TR are limited to the NuScale passive nuclear plant design.

24 If NuScale chooses to incorporate by reference TR

-0815-16497 as part of its application, it must demonstrate that the reactor design meets all the conditions of applicability in Table 3-1 and all the augmented design, qualification, and QA provisions in Table 3-2. Additionally, any applicant referencing this TR must take the following actions: 4.1 Address the guidance in RG 1.155, Appendix A, in sufficient detail to enable the NRC staff to verify that the relevant QA program would meet or exceed the guidance in RG 1.155. 4.2 Confirm that the VRLA batteries and their structures are seismic Category 1. To provide reasonable assurance that the VRLA batteries will perform as intended, an applicant that references the TR shall provide a COL action item to support that the VRLA batteries and their structures are seismic Category 1. A qualification testing plan includes environmental and seismic qualification and a technical functional requirement for VRLA batteries to show they can perform as intended. 4.3 Demonstrate that operator actions are not necessary to ensure the performance of safety-related functions for any postulated DBE (i.e., the design does not include Type A variables as defined in IEEE Std.

497-2002, as modified in RG 1.97, Regulatory Position C.4), as presented in Chapter 15 of its FSAR and the human factors analysis in Chapter 18 of its FSAR.

4.4 Evaluate

the frequency for which a combination of an AOO and an actuation of the NuScale ECCS is realistically expected to occur, and show that such a combination of events i s not expected to occur during the lifetime of the module.

4.5 Demonstrate

that the reactor can be brought to a safe shutdown using only safety-related equipment in the absence of electrical power following a DBE, with margin for stuck rods.

Alternatively, an applicant addressing this condition may provide justification, for NRC review, for a less restrictive approach. 5.0 Conclusions

The NRC staff approves the use of NuScale TR

-0815-16497 as a reference document only to the NuScale passive nuclear plant design, subject to the conditions and limitations specified in Section 4.0 of this safety evaluation report.

Specifically, based on its review of TR-0815-16497 , the NRC staff finds that if the NuScale reactor design can meet the conditions of applicability and the augmented design, qualification, and QA provisions, Class 1E power sources would not be necessary. This approval of the concepts discussed in the TR does not constitute approval of any specific design. Any applicant referencing this TR in support of a design other than the NuScale passive nuclear plant design must submit information, for NRC staff review, that justifies the applicability of this TR, or a variation of it, to the respective design.

Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report TR-0815-16497-NP-A Rev. 1 © Copyright 201 8 by NuScale Power, LLC Section B Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems

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Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems The AP1000 design as presented does not require Class 1E alternating current (ac) electrical power, except that provided by the Class 1E direct current (dc) batteries and their inverters, to accomplish the plant's safety-related

functions.

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

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Safety Classification of Passive Nuclear Power Plant Electrical Systems

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Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems

Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report TR-0815-16497-NP-A Rev. 1 © Copyright 201 8 by NuScale Power, LLC Section C LO-1116-51959 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com December 5, 2016 Docket: PROJ0769 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Response to Request for Additional Information Letter No. 8 for the review of Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 0.(CAC NO. RQ6002) dated October 7, 2016 (NRC Project No. 0769).

REFERENCES:

1.Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commissi on,"Safety Classification of Passive Nuclear Power Plant Electric al Systems,"

Revision 0, TR-0815-16497, dated October 29, 2015 (ML 15306A126).

2.NuScale Topical Report, "Safety Classification of Passive Nuclear Powe r Plant Electrical Systems," Revision 0, TR-0815-16497, dated October 29, 2015 (M L 15306A126).3.Letter from U.S. Nuclear Regulatory Commission to NuScale Power, LLC,"Request for Additional Information Letter No. 8 for the Review of Topical Report0815-16497, "Safety Classification of Passive Nuclear Power Plant Electric al Systems," Revision 0.(CAC NO. RQ6002) dated October 7, 2016 (NRC Projec t No. 0769, ML16281A103).In a letter dated October 29, 2015 (Reference 1), NuScale Power, LLC (NuScale) submitted the topical report entitled "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 0 (Reference 2). In a letter dated October 6, 2016 (Reference 3), the NRC Staff submitted Requests for Additional Information (RAI) regarding the subject topical report. The purpose of this letter is to provide NuScale's response to the NRC RAIs. Enclosure 1 is the NuScale Response to Request for Additional Information Letter No. 8 for the review of Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 0. This letter makes no regulatory commitments and no revisions to any existing regulatory commitments. Please feel free to contact Steven Unikewicz at 240-833-3015 or at sunikewicz@nuscalepower.com if you have any questions.

Sincerely, Thomas A. Bergman Vice President, regulatory Affairs NuScale Power, LLC Sincerely , Thomas A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A. Ber g man Vice Presi de de de de de de de de de d de de d de de d d d de de de d de de de de d de de de de d de de d d d de de d de de de de d de e de e e d e e e e e e e e e e e e e e e e e e e e e e e e e e e e nt n n n n n n n n n n n n n n n n n nt n n n n n n n t , re gu gu u u u u u u u u u u u u u u u u u u u u u u u g g la la la la a a la la la la a a la l la la l la l a a a l a a l l l l l l l l l l l tor y Affai rNuScalePower L L L L L L L L L L L L L L L L L LC LO-1116-51959 Page 2 of 2 11/17/16 Distribution: Frank Akstulewicz, NRC, TWFN-6C20 Greg Cranston, NRC, TWFN-6E55 Omid Tabatabai, NRC, TWFN-6E55 Mark Tonacci, NRC, TWFN-6E55 Samuel Lee, NRC TWFN-6E55

Enclosure 1: Response to NRC Letter "Request for Additional Information Letter No. 8 for the review of Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 0 LO-1116-51959 Revision 0 Page 1 of 15

© Copyright 2016 by NuScale Power, LLC

NRC RAI Number: 8 NRC RAI Date: October 29, 2016 NRC Review of: Safety Classification of Passive Nuclear Power Plant Electrical Systems, TR-

0815-16497, Revision 0.

The electrical power system presented in the Licensing Topical Report (LTR) depicts a design with no Class 1E power sources as the proposed reactor design does not require any safety-related electrical loads to support the safety analyses. However, 10 CFR 50.34(f)(2)(xx) calls for vital-bus-powered post- accident monitoring instrumentation with backup power from emergency power supplies. In order for the staff to be able to conclude that an electrical design such as the one presented in the TR provides equivalent protection to that prescribed in the regulation, the staff must be able to conclude that the proposed design is of similar (high) reliability. To that end, the staff requires the following additional information:

NRC RAI Question Number: 08.03.02-01

NRC RAI Question:

Table 3-2 of the TR states that Valve Regulated Lead Acid (VRLA) batteries will be used for the direct current (DC) power system. Based on various industry publications, including Institute of Electrical and Electronics Engineers (IEEE) Std.

1187, "Recommended Practice for Installation Design and Installation of Valve- Regulated Lead-Acid (VRLA) Batteries for Stationary Applications," the life of a VRLA battery can be seriously and suddenly reduced due to factors such as: 1) prolonged high ambient temperatures, 2) magnitude and frequency of discharge cycles, and 3) overcharging.

Please describe how these factors will be addressed in the design and operation of a passive reactor nuclear power plant that relies on VLRA battery systems to ensure high reliability DC power system.

NuScale RAI Question Response:

NuScale agrees that the life of a VRLA battery can be seriously and suddenly reduced due to prolonged high ambient temperatures. These effects are mitigated through the implementation of IEEE Std. 1187 and IEEE Std. 1188, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Valve-Regulated Lead- Acid (VRLA) Batteries for Stationary Applications" as noted in Table 3-2. Additionally, IEEE Std. 1187 refers to IEEE Std. 1491, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications," and IEEE Std. 1635, "IEEE/ASHRAE Guide for the Ventilation and Thermal Management of Batteries for Stationary Applications."

The use of IEEE Stds. 1187 and 1188 as supplemented by IEEE Stds. 1491 and 1635 provide reasonable assurance that the VLRA batteries will function as intended following exposure to prolonged periods of high ambient temperature. Further, the heating, ventilation, and air conditioning systems serving the battery and associated charger rooms are provided back-up LO-1116-51959 Revision 0 Page 2 of 15

© Copyright 2016 by NuScale Power, LLC power from the backup power supply system to avoid prolonged periods of high ambient temperature.

NuScale agrees that the life of a VRLA battery can be seriously and suddenly reduced due to the magnitude and frequency of discharge cycles. Magnitude and frequency of discharge cycles are design considerations addressed in IEEE Std. 1187 and IEEE Std. 1188, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Valve-Regulated Lead- Acid (VRLA) Batteries for Stationary Applications" as noted in Table 3-2. Additionally, IEEE Std. 1187 refers to IEEE Std. 1491, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications." IEEE Std. 1491 provides monitoring criteria that may be used to detect and monitor a battery for degradation.

The use of IEEE Stds. 1187 and 1188 as supplemented by IEEE Std. 1491 provide reasonable assurance that the VLRA batteries are designed, constructed, and monitored considering the potential for magnitude and frequency of discharge cycles to degrade battery performance.

NuScale agrees that the life of a VRLA battery can be seriously and suddenly reduced due to overcharging. These effects are mitigated through the implementation of IEEE Std. 1187.

IEEE Std. 1187 refers to IEEE Std. 1491, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications."

The use of IEEE Std. 1187 as supplemented by IEEE Std. 1491 provides reasonable assurance that the VLRA batteries will not be overcharged and that instances of potential overcharging will be detected prior to degrading a battery to a point where it is not able to perform its intended function.

Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems":

This RAI Response does not require Licensing Document revisions.

Attachments:

None

LO-1116-51959 Revision 0 Page 3 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question Number: 08.03.02-02 NRC RAI Question:

Table 3-2 of the TR provides a comparison of the "Class 1E DC Electrical system" to the "Non Safety-Related DC Electrical System(s) Relied upon to Power Type B and Type C Accident Monitoring Instrumentation." Under the provision "Quality Assurance" in the Table 3-2, it stated that a Graded QA Program will be applied to the DC Electrical System, which will meet or exceed the augmented QA provisions specified in RG 1.155, Appendix A, "Quality Assurance Guidance for Non-Safety Systems and Equipment". RG 1.155, Appendix A provides QA guidance for meeting the requirements of 10 CFR 50.63 and not already explicitly covered by existing QA requirements in 10 CFR Part 50 in Appendix B or R.

Please describe the proposed quality assurance program in sufficient detail that will allow the staff to verify it meets or exceeds the provisions of RG 1.155.

NuScale RAI Question Response:

A COL applicant that references Topical Report 0815-16497 will be required to incorporate the guidance contained in RG 1.155 Appendix A, "Quality Assurance Guidance for Non-Safety Systems and Equipment as part of their Quality Assurance Program." It is not the intention of this LTR to provide an example quality assurance program as that is COL applicant specific.

Verification of sufficient detail is considered a potential NRC COL review topic.

Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems":

This RAI Response does not require Licensing Document revisions.

Attachments:

None

LO-1116-51959 Revision 0 Page 4 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question Number: 08.03.02-03 NRC RAI Question:

Table 3-2 of the TR, under the provision "Batteries," states that the VRLA batteries have augmented design, QA, and qualification provisions.

Please describe the methods and processes that will be used by a passive reactor nuclear power plant to verify that VRLA batteries will perform their intended function(s) during normal operation, operational occurrences and postulated design basis events.

NuScale RAI Question Response:

The VRLA batteries used in a passive reactor nuclear power plant design are not credited for use in mitigating the consequences of postulated design basis events.

To provide reasonable assurance that VRLA batteries will perform their intended function(s) when called upon, an applicant utilizing this TR shall implement a testing and monitoring program as described in IEEE Std. 1188, "Recommended Practice for Maintenance, Testing, and Replacement of Valve-Regulated Lead- Acid (VRLA) Batteries for Stationary Applications" and in IEEE Std. 1491, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications." These Standards provide for a wide variety of operating parameters to be monitored on a continuous basis including cell specific parameters.

Additionally, Table 3-2 of the TR notes that applicants are required to environmentally qualify their VRLA batteries in accordance with IEEE Std. 323, "IEEE Standard for Qualifying Class 1E

Equipment for Nuclear Power Generating Stations," and seismically qualify their batteries in accordance with IEEE Std. 344, "IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." Such qualification provides further assurance that the batteries will perform their intended functions.

NRC RAI Question (Continued):

Please also provide the industry standards or applicable references that will be used for verification purposes.

NuScale Response (Continued):

The industry standards that will be used for verification purposes include:

1. IEEE Std. 323, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations" as endorsed by RG 1.89 2. IEEE Std. 344, "IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations" as endorsed by RG 1.100 3. IEEE Std. 1188, "Recommended Practice for Maintenance, Testing, and Replacement of Valve-Regulated Lead- Acid (VRLA) Batteries for Stationary Applications" 4. IEEE Std. 1491, "IEEE Guide for Selection and Use of Battery Monitoring Equipment in Stationary Applications" LO-1116-51959 Revision 0 Page 5 of 15

© Copyright 2016 by NuScale Power, LLC Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems":

This RAI Response does not require Licensing Document revisions.

Attachments:

None LO-1116-51959 Revision 0 Page 6 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question Number: 08.03.02-04 NRC RAI Question:

The TR describes the presented dc power system as "highly reliable" and substantially equal in reliability to that of an analogous Class 1E dc power system. These statements have not been described adequately in the TR. In order for the staff to be able to fully evaluate the design and ultimately conclude on its acceptability as a highly reliable power system, the staff requests that NuScale provide a description of the methodology that will be used to compare the highly reliable DC system to be described in its design certification application to a Class 1E dc power system to show that the highly reliable DC system is substantially equal in reliability to a typical Class 1E dc power system.

NuScale RAI Question Response:

The LTR seeks NRC approval of the conditions of applicability, and the methodology and bases used in their development. The LTR further seeks NRC approval of the acceptability of a set of augmented design, qualification, and QA provisions to be applied by the conditions of applicability. The augmented provisions are intended to ensure suitable reliability for a direct current (DC) power system performing the nonsafety-related functions described in the LTR, analogous to a traditional licensee's application of the augmented provisions for a 1E power system, which has been judged acceptable without a quantitative reliability acceptance criterion.

The LTR terms the subject DC power system(s) as the "highly reliable DC electrical system(s)." The LTR further states that a comparison of specified augmented design, qualification, and QA provisions to a typical Class 1E DC electrical system "supports a determination that the augmented provisions result in an electrical system reliability substantially similar to that of a Class 1E DC power system." In using these descriptive phrases, NuScale intended to reflect, qualitatively, the attributes of a DC power system meeting the specified augmented design, qualification, and QA provisions. However, these descriptive phrases were not intended to define additional conditions for use of the LTR, distinct from the specified augmented provisions that a user of the report must implement.

However, while NuScale does not intend that a user of the LTR must, as a condition of its use, explicitly and quantitatively demonstrate "substantially similar" reliability to a typical Class 1E DC power system, NuScale intends to make available such a demonstration as one method of determining the system performs at a suitable reliability to perform the important functions addressed by the LTR. The NuScale example calculation shows that the highly reliable DC electrical system has a reliability that is approximately a factor of 5 better than that of a class 1E power system. In comparing reliability to a typical design, a user of the report should consider the specific nonsafety-related functions performed by their DC system, and the safety characteristics and risk profile of the overall plant design.

The method a Design Certification Applicant may use to compare the reliability of the highly reliable DC system to that of a typical Class 1E DC power system is as follows:

• First, define the required mission(s) the power system is required to support.

LO-1116-51959 Revision 0 Page 7 of 15

© Copyright 2016 by NuScale Power, LLC

• Second, define the design and system boundaries that are needed to accomplish the required mission.

• Third, establish a measure of comparable reliability (i.e., how reliable should the system be) by reviewing a "typical" design. This typical system will be a Class 1E DC power system that supports a similar mission for a licensed facility. NuScale will build a reliability model for the typical design and assess that reliability.

• Fourth, build a reliability model for its highly reliable DC system design and assess the reliability in fulfilling the required mission.

• Finally, compare the reliability of the highly DC system to that of the typical Class 1E design. A determination that the reliability of the highly reliable DC system has reliability equal to or greater than the typical design is sufficient 1 to conclude that the reliability is acceptable for the required missions.

Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems":

This RAI Response does not require Licensing Document revisions.

Attachments:

None

1 Reliability less than, but similar to, that of the typical system is not expected, but would require further evaluation to determine if it is adequate to support the required missions.

LO-1116-51959 Revision 0 Page 8 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question Number: 08.03.02-05 NRC RAI Question:

The regulation set forth in 10 CFR 50.55a(h)(3) requires that design certification applications under part 52 meet the requirements of IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations." IEEE Std. 603-1991 provides a definition of "safety system" and states that the electrical portion of the safety systems, that perform safety functions, is classified as Class 1E. Included in the definition of safety system is a system that is relied upon to remain functional during and following a design basis event to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition.

Condition of Applicability Item I.1.b, contained in Table 3-1 of the TR, states that sufficient reactor coolant inventory and negative reactivity are assured during and following a design basis event to achieve and maintain safe shutdown. Additionally, the TR provides a clarifying example assessment to illustrate how the Conditions of Applicability would be demonstrated. This example assessment did not include a quantitative safety analysis to demonstrate the ability to insert sufficient negative reactivity during and following a design basis event to achieve and maintain safe shutdown.

SECY-94-084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs," clarifies the conditions that constitute a safe shutdown as reactor sub-criticality, decay heat removal, and radioactive material containment.

Additionally, SECY 94-084 states that an appropriate safety analysis can be used to demonstrate passive system capabilities to bring the plant to a safe stable condition and to maintain this condition. NRC staff is seeking to clarify whether Condition of Applicability Item is consistent with the description of safe shutdown provided in SECY-94-084. Additionally, NRC staff is seeking to clarify the requirements for demonstrating how Condition of

Applicability Item I.1.b is satisfied. NRC staff requests the following additional information:

1. Specify the criteria that constitute a safe-shutdown as applied to Condition of Applicability Item I.1.b NuScale RAI Question Response:

The criteria that constitute a safe shutdown are sub-criticality and decay heat removal in order to maintain fuel clad integrity (radioactive material containment).

These criteria are based on guidance for attaining safe shutdown in current generation reactors and for certified advanced reactors. NRC regulations that address safe shutdown do not include criteria for a safe shutdown condition or for the reliability of systems necessary to attain safe shutdown. What constitutes safe shutdown is addressed in SECY-94-084 for advanced reactors and in guidance such as RG 1.139 and BTP 5-4 for current generation reactors.

For current generation reactors that address RG 1.139 and BTP 5-4, safety analyses of design basis events are not typically relied on to demonstrate design capability to attain safe shutdown conditions. Rather, safety analyses of DBEs (as presented in Chapter 15 of a facility's final safety analysis report) is focused on the short term reactor response to ensure that fuel integrity LO-1116-51959 Revision 0 Page 9 of 15

© Copyright 2016 by NuScale Power, LLC is maintained for anticipated operational occurrences (AOO) and a coolable core geometry is maintained for accidents. The safety analyses thereby evaluate the capability of the reactivity control systems to perform their protection function, rather than their shutdown function.

The safety issue that underpins these NRC guidance documents (SECY-94-084, RG 1.139, and BTP 5-4) and their specification of a safe shutdown condition and the systems' capability to attain safe shutdown is relevant to GDC 34, in that systems or equipment failures resulting in insufficient heat removal capability can lead to core damage. Per RG 1.139, a risk evaluation of the heat removal capability of a typical pressurized-water reactor (PWR) and boiling water reactor (BWR) plant following a plant trip showed that

...systems or equipment failures that led to the inability to remove decay heat resulted in a higher probability of a core melt than that predicted for a large LOCA for both PWRs and BWRs. Consequently, a significant safety benefit will be gained by upgrading those systems and equipment needed to maintain the RCS at the hot-standby condition for extended periods or those needed to cool and depressurize the RCS so that the RHR system can be operated.

To address the safety issue of system limitations or equipment failures resulting in insufficient heat removal capability for advanced designs, NRC staff proposed in SECY-94-084 that passive system capabilities can be demonstrated by: 1. A safety analysis to demonstrate that the passive systems can bring the plant to a safe stable condition and maintain this condition, that no transients will result in the SAFDLs and pressure boundary design limit being violated, and that no high-energy piping failure being initiated from this condition will result in violation of 10 CFR 50.46 criteria. 2. A probabilistic reliability analysis, including events initiated from the safe shutdown conditions, to ensure conformance with the safety goal guidelines. The PRA would also determine the R/A missions of risk significant systems and components as a part of the effort for regulatory treatment of non-safety systems. Conservative assumptions are applied to Chapter 15 safety analysis of DBEs appropriate for the intended purpose of ensuring appropriate margins to protect fuel integrity or core coolability. Although these safety analyses can be used to demonstrate adequate shutdown capability per SECY-94-084, application of the same conservative assumptions may lead to excessive margin with respect to shutdown capability. Shutdown with additional margin due to conservative safety analysis assumptions may not be appropriate, considering a specific design's heat removal and

shutdown capability and reliability.

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© Copyright 2016 by NuScale Power, LLC NRC RAI Question (Continued):

2. Describe how a future passive plant applicant will demonstrate that electrical power is not necessary to achieve and maintain a safe shutdown for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

NuScale Response (Continued):

Electrical power is not necessary to achieve and maintain a safe shutdown condition for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the design includes safety-related capability to maintain a safe shutdown condition that does not depend on electrical power.

To demonstrate that capability, an applicant will evaluate the reactivity control systems to ensure sufficient shutdown function capability and evaluate the decay heat removal system to ensure sufficient heat removal capability. To ensure that safe shutdown capability is sufficient to address the safety issue of heat removal reliability, a probabilistic risk assessment is used to ensure that the reliability of systems used to achieve and maintain safe shutdown supports conformance to the commission's safety goal guidelines.

The response to RAIs 08.03.02-05, Question 1 and 2, describes an approach to meet the Conditions of Applicability. The design capability along with the approach to meet Conditions of Applicability is design specific and should be evaluated as part of an applicant's design certification or combined license application, rather than evaluating it within the scope of Topical

Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," which is intended to be design independent. Requiring a specific approach to meet Conditions of Applicability may be suitable for some designs but overly prescriptive for other designs.

Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems":

This RAI Response does not require Licensing Document revisions.

Attachments:

None LO-1116-51959 Revision 0 Page 11 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question Number: 08.03-02-06 NRC RAI Question:

GDC 15 requires the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

Condition No. I.1 of the Conditions of Applicability, contained in Table 3-1 of TR-0815- 16497, states that for a design basis event, electrical power is not necessary to maintain the reactor coolant pressure boundary (RCPB) integrity for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Additionally, TR-0815-16497 provides a clarifying example assessment to illustrate how the Conditions of Applicability would be demonstrated. This example assessment includes a safety analysis showing an example passive plant response to an anticipated operational occurrence. The safety analysis shows that the example passive plant response to the anticipated operational occurrence includes establishing a direct coolant flow path between the reactor core and the containment, thereby removing a fission product barrier. This caused NRC staff to question if the items under Conditions of Applicability I.1 are sufficient to demonstrate RCPB integrity. Additionally, RIS 2005-29, discusses the design criteria for event non-escalation. NRC staff is questioning why the removal of a fission product barrier is not considered an event escalation.

NRC staff requests the following information:

1. Specify the criteria that constitute RCPB integrity as applied to Condition No. I.1 of the Conditions of Applicability.

NuScale RAI Question Response:

RCPB integrity refers to the structural integrity of RCPB components designed to retain pressure and contain reactor coolant. A loss of RCPB integrity or loss of structural integrity involves a mechanical failure in an RCPB component, for example a pipe. For an AOO, the RCPB integrity acceptance criterion is that pressure in the reactor coolant and main steam systems should be maintained below 110 percent of the design values. For a postulated accident, the criteria for RCPB integrity is that pressure in the RCS is maintained below acceptable design limits, considering potential brittle as well as ductile failures.

Opening of a valve(s) that allows reactor coolant to pass into or out of the RCPB does not involve a mechanical failure in an RCPB component and does not constitute a loss of RCPB integrity. An interpretation that RCPB integrity is lost when opening a valve to allow fluid to pass through the RCPB is problematic in the following respects:

• It would preclude advanced designs that offer improvements in safety by relying on valves to depressurize the RCS for safe shutdown. As described in RG 1.139 and BTP 5-4, depressurization is one of the processes that support safe shutdown. The safety benefit of RCS depressurization through valves include: providing highly reliable means for depressurization; reducing the driving force for coolant out of the RCS; and reducing the driving force for fission products out of containment in the event of a loss in clad LO-1116-51959 Revision 0 Page 12 of 15

© Copyright 2016 by NuScale Power, LLC integrity. Further, the ability to depressurize and provide long term heat removal using valves is consistent with the NRC's position to minimize the potential for an intersystem Loss of Coolant Accident (LOCA) outside of containment in advanced or evolutionary light-water reactors in SECY-90-016 and SECY-93-087 and their associated staff requirements memoranda. Lastly, the capability of advanced designs, such as NuScale, to safely depressurize the RCS reduces the importance of RCPB integrity to safety.

• It is not consistent with the licensing basis for PWRs and BWRs; it would imply that these designs do not comply with GDC 15. GDC 15 is relevant to Standard Review Plan (SRP)

Section 15.0 "as it relates to the RCS and its associated auxiliaries being designed with appropriate margin to ensure that the pressure boundary will not be breached during normal operations, including AOOs." Interpreting that opening of a valve to allow fluid to pass into or out of the RCPB constitutes a breach of the RCPB would imply that licensed facilities do not meet GDC 15 in the following instances.

• PWRs and BWRs rely on safety relief valves to prevent exceeding RCPB design limits for select AOOs.

• PWRs and BWRs evaluate inadvertent opening of a pressure relief valve as an AOO in accordance with SRP 15.0 and SRP 15.6.1.

• BWRs open valves (safety relief valve, reactor core isolation cooling) to route reactor coolant out of the RCPB to containment for the purpose of heat removal and RCS depressurization prior to transitioning to heat removal using the residual heat removal (RHR) system at low pressures.

• The RHR systems for PWRs and BWRs are not part of the RCPB. Valves are opened upon RHR system actuation to cycle reactor coolant into and out of the RCPB and through the RHR system for the purpose of heat removal.

• It is not consistent with Appendix A to 10 CFR 50 which address maintaining structural integrity of RCPB components rather than preventing the opening of valves to allow fluid to pass into or out of the RCPB. Under Appendix A, GDC 14, "Protection by Multiple Fission Product Barriers," addresses RCPB integrity: "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."

Further, GDCs 15, 17, 28, 30, 31, 32, 33 and 34 include provisions to design, operate, and maintain the RCPB in order to prevent loss of structural integrity.

The opening of RCPB valves is addressed by 10 CFR 50.34(f)(1)(iv). The safety concern addressed by 10 CFR 50.34(f)(1)(iv), however, is the adverse impact on core damage frequency (CDF) due to frequent valve actuation, rather than a loss of RCPB integrity. 10 CFR 50.34(f)(1)(iv) was added after the TMI-2 accident when it was recognized that a loss of coolant from a stuck open PORV and other small-break LOCA contributors was more likely to lead to core damage that a large pipe break. The rule requires evaluation of the potential benefit from automatic PORV isolation for current generation PWR's in order to reduce CDF by reducing the demand on ECCS system. For advanced designs with a low CDF, a reduction in CDF by limiting deliberate or inadvertent RCPB valve actuation to reduce the CDF may not be warranted.

LO-1116-51959 Revision 0 Page 13 of 15

© Copyright 2016 by NuScale Power, LLC NRC RAI Question (Continued):

2.Explain why the removal of a fission product barrier during an anticipated operationa l occurrence is not considered an event escalation.NuScale RAI Question Response (Continued): Opening a valve to depressurize the RCS and establish long term cooling is not considered a removal of a fission product barrier, and thus not an event escalation, because the functions of the RCS barrier are not lost. The RCS barrier continues to provide a confined volume for reactor coolant which allows a flow path for cooling the core and thus, confining fission products to the fuel. The basis for this response is as follows. As part of the analysis acceptance criteria for AOOs (p15.0-5, SRP 15.0), The reviewer applies a third criterion, based on the American Nuclear Safety (ANS) standards to ensure that there is no possibility of initiating a postulated accident with the frequency of occurrence of an AOO. This review is performed under Acceptance Criterion 2.A.iii, based on the ANS standards referenced in SRP 15.0, which states: An AOO should not generate a postulated accident without other faults occurring independently or result in a consequential loss of function of the RCS or reactor containment barriers.

Based on SRP 15.0, the intent of the non-escalation criterion is to ensure that the consequences associated with accidents do not occur at the frequency of an AOO. Such a condition would lead to an unacceptable risk to the public, due to frequent events with more significant consequences. The two parts of the non-escalation criterion, preventing accidents generated by an AOO and protecting the functions of barriers, are intended to prevent such an increase in risk. Thus, events that do not result in unacceptable consequences or significantly increase the risk for radiological release do not challenge the intent of the non-escalation criterion. With respect to whether opening of a valve to depressurize the RCS involves a "consequential loss of function of the RCS barrier," it is helpful to review the regulatory history of this acceptance criterion. Acceptance Criterion 2.A.iii and the definition of event categories were first introduced in Rev. 0 of the SRP and were derived from the PWR and BWR ANS standards for nuclear safety. The PWR standard referred to in Rev. 0 of the SRP, ANSI

N18.2, was reviewed to clarify what was meant with this acceptance criterion. The SRP cites this acceptance criterion for ANSI N18.2 Condition II ("Incidents of Moderate Frequency") and Condition III ("Infrequent Incidents") events. ANSI N18.2 presents the following events as examples of Condition II and III events:

• Condition II example: "depressurization by spurious operation of an active element, forexample, relief valve, pressurizer spray valve."

LO-1116-51959 Revision 0 Page 14 of 15

© Copyright 2016 by NuScale Power, LLC

• Condition III example: "loss of reactor coolant, such as from a small ruptured pipe or from a crack in a large pipe, which would prevent orderly reactor shutdown and cooldown assuming makeup is provided by normal makeup systems only" (i.e., small-break loss of coolant accident (LOCA). These ANSI N18.2 examples are also included in examples of AOOs in SRP 15.0. The ANSI N18.2 design requirement that is the basis for Acceptance Criterion 2.A.iii (3) in SRP 15.0 is: A Condition III incident shall not, by itself, generate a Condition IV fault or result in a consequential loss of function of the reactor coolant system or reactor containment barriers.

Thus, Condition II and III events do not in themselves involve a consequential (significant) 2 loss of function of the RCS barrier. Of particular interest, the two examples given above that result in continuously blowing down reactor coolant from the RCS, either through a valve (Condition II) or through a small-break LOCA (Condition III), do not result in a consequential loss of function of the RCS barrier. A consequential loss of function of the RCS barrier is associated with only Condition IV ("Limiting Fault" or "Postulated Accident") events. Based on the Condition IV example provided in ANSI N18.2 and repeated in SRP 15.0, a loss of function of the RCS barrier is associated with a major pipe rupture. This interpretation is consistent with Appendix A to 10 CFR Part 50 Section II, "Protection by Multiple Fission Product Barriers" which includes general design criteria for the fission product barriers. GDCs 14 and 15 state the design criteria for the RCPB and the RCS, which are intended to have "an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture." That is, the function of the RCS, as implemented by these GDCs, is not to form a leak-tight radionuclide barrier to the environment; in contrast the function of the containment as stated in GDC 16 is to "-establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment -."

Rather, the RCPB functions, which are equivalent to that of the RCS barrier, are stated in SRP 5.2.3 under the technical rationale for GDC 14: The RCPB provides a fission product barrier, a confined volume for the inventory of reactor coolant, and flow paths to facilitate core cooling. A loss of these functions is also described in SRP Section 5.2.3 as a "gross failure of the RCPB resulting in substantial reduction in capability to contain reactor coolant inventory, reduction in capability to confine fission products, or interference with core cooling."

2 The word "consequential" can mean resultant or significant. "Significant" makes more sense in the context of Acceptance Criterion 2.A.iii. Under "Barrier Integrity Criteria" for the RCPB, ANSI N18.2 (3 rd criterion, p8, Reference 9) states that the RCPB "shall withstand Conditions I, II, III and IV, including thermal transients associated with the operation of the emergency core cooling system, without significant consequential rupture (that is, if consequential rupture occurs, it shall not appreciably worsen the safety consequences)." The terms "result in a consequential loss of function" (Acceptance Criterion 2.A.iii) and "significant consequential rupture" (ANSI N18.2) are equivalent and are interpreted to mean "result in a significant loss of function due to RCPB rupture." In comparison, the design criterion for the "Containment Barrier" is: "The design pressure, temperature, and leakage rate of the reactor containment shall not be exceeded as a result of Conditions I, II, III, or IV."

LO-1116-51959 Revision 0 Page 15 of 15

© Copyright 2016 by NuScale Power, LLC Thus, gross failure is a necessary condition for such a substantial loss of function. Further, gross failure resulting in a substantial reduction of any one of the three functions of the RCPB constitutes a substantial or consequential loss in function of the RCS barrier. This is because the function of fission product confinement is integrated with the functions of inventory control and heat removal; i.e., the function of fission product confinement is maintained if the functions of inventory control and heat removal are maintained. "Fission product barrier" does not mean that leakage from fuel defects or activation products in RCS coolant must be confined in the RCS after all design basis events. It refers to maintaining integrity of the cladding. Absent the potential for fuel cladding failure, there are no significant radiological consequences associated with the event, and therefore no "consequential loss of function" of

the RCS barrier. Thus, opening a valve to depressurize the RCS and establish long term cooling does not result in a consequential loss of function of the RCS barrier, i.e. a substantial reduction in capability to contain reactor coolant inventory, reduction in capability to confine fission products, or interference with core cooling. Accordingly, opening such valve during an anticipated operational occurrence is not considered an event escalation. Impact of NRC RAI Question Response on Topical Report 0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems": This RAI Response does not require Licensing Document revisions.

Attachments:

None Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report TR-0815-16497-NP-A Rev. 1 © Copyright 201 8 by NuScale Power, LLC Section D

LO-0217-52963 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com February 1, 2017 Docket: PROJ0769 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Topical Report TR-0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1 (CAC No. RQ6002)

REFERENCES:

1.Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission,"Safety Classification of Passive Nuclear Power Plant Electrical Systems,"Revision 0 TR-0815-16497, dated October 29, 2015 (ML 15306A126)2.NuScale Topical Report, "Safety Classification of Passive Nuclear Power PlantElectrical Systems," Revision 0, TR-0815-16497, dated October 29, 2015(ML 15306A126)3.Letter from U.S. Nuclear Regulatory Commission to NuScale Power, LLC,"Request for Additional Information Letter No. 8 for the Review of NuScale Topical Report (TR) 0815-17497, 'Safety Classification of Passive NuclearPower Plant Electrical Systems,' Revision 0 (CAC No. RQ6002)," datedOctober 7, 2016 ( ML16281A103).4.Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission,"Submittal of Response to Request for Additional Information Letter No. 8 for the review of Topical Report 0815-16497, "Safety Classification of PassiveNuclear Power Plant Electrical Systems," Revision 0, dated December 5, 2016(ML 16340D339)In a letter dated October 29, 2015 (Reference 1), NuScale Power, LLC (NuScale) submitted the topical report titled "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 0 (Reference 2). In a letter dated October 7, 2016 (Reference 3), the NRC Staff provided Request for Additional Information (RAI) No. 8 regarding the subject topical report. NuScale submitted responses to the NRC RAI in a letter dated December 5, 2016 (Reference 4). The purpose of this letter is to provide Revision 1 of the Topical Report TR-0815-16497 incorporating changes that resulted from the RAI responses in Reference 4. Changes are summarized with revision bars in the margin. Additional edits have also been made to the proprietary markings of the report, assuring consistency with the recently submitted NuScale Final Safety Analysis Report and other industry guidance. These proprietary marking revisions were not highlighted with revision bars. Enclosure 1 is the proprietary version of the report titled "Safety Classification of Passive Nuclear Power Plant Electrical Systems" Revision 1. Enclosure 2 is the nonproprietary version of the report titled "Safety Classification of Passive Nuclear Power Plant Electrical Systems" Revision 1. NuScale requests that the proprietary Enclosure 1 be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request.

LO-0217-52963 Page 2 of 2 02/1/2017NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com This letter and its enclosures make no regulatory commitments and no revisions to any existing regulatory commitments. Please feel free to contact Jennie Wike at (541) 360-0539 or at jwike@nuscalepower.com if you have any questions.

Sincerely, Thomas A. Bergman Vice President, Regulatory Affairs NuScale Power, LLC Distribution: Frank Akstulewicz, NRC, TWFN-6C20 Greg Cranston, NRC, TWFN-6E Omid Tabatabai, NRC, TWFN-6E Samuel Lee, NRC, TWFN-6C20 Enclosure 1: TR-0815-16497-P, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, proprietary version Enclosure 2: TR-0815-16497-NP, "Safety Classification of Passive Nuclear Power Plant Electrical Systems," Revision 1, nonproprietary version Enclosure 3: Affidavit of Thomas A. Bergman, AF-0217-52964 y, Thomas A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A A............Ber g manVice Presi de de de de de d d de de de e d d e e de e e d e e d e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e nt n n n n n n n n n n n n n n n n n , Re g ula to o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o o ry r r r r r r r r r r r r r Affairs Nu S cale Powe r, r r r r r r r r r r r r r r r r r r r r r r r L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L LC L L L L L LC LC L LC LC L LC LC L L L L LC LC LC C LC LC LC L LC LC LC LC L L LC LC LC C C C LC LC LC C LC LC LC L L LC C LC C C C C C C C C LC LC C C C L L L LC LC L LC LC LC C C C C C C LC LC LC L L C L C C C C C C C C C C LO-0217-52963 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com Enclosure : TR-0815-16497- P, "Safety Classification of Passive Nuclear Po wer Plant Electric al Systems,"

Revision 1, proprietary version

LO-0118-58309 Enclosure 3:

Affidavit of Thomas A. Bergman, AF-0118-58310

AF-0118-58310 Page 1 of 2NuScale Power, LLC AFFIDAVIT of Thomas A. Bergman I, Thomas A. Bergman, state as follows:

(1)I am the Vice President of Regulatory Affairs of NuScale Power, LL C (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in th is Affidavit that NuScale seeks to have withheld from public disclosure, and am authoriz ed to applyfor its withholding on behalf of NuScale.

(2)I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request towithhold information from public disclosure is driven by one or more of the following:(a)The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without alicense from NuScale, would constitute a competitive economic disadvantage to NuScale.(b)The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of thedata secures a competitive economic advantage, as described more fully in paragraph 3 ofthis Affidavit.(c)Use by a competitor of the information requested to be withheld would reduce thecompetitor's expenditure of resources, or improve its competitive position, in the design

, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.(d)The information requested to be withheld reveals cost or price information, productioncapabilities, budget levels, or commercial strategies of NuScal e.(e)The information requested to be withheld consists of patentable ideas.

(3)Public disclosure of the information sought to be withheld is likely to cause substantial harm toNuScale's competitive position and foreclose or reduce the availabilit y of profit-making opportunities. The accompanying Approved Version of Topical Report TR-0815-16497 Revision 1reveals distinguishing aspects about the process and method by which NuScale develops itsSafety Classification of Passive Nuclear Power Plant Electrical Systems Methodology.

NuScale has performed significant research and evaluation to develop a basis for this methodology and has invested significant resources, including the expenditure of a considerable sum of money

.The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScal e.If the information were disclosed to the public, NuScale's competitors would have access to theinformation without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropria tion of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitiveadvantage to seek an adequate return on its investme nt.(4)The information sought to be withheld is in Section B of Enclosure 1 to NuScale letter titled"Submittal of Approved Version of Topical Report TR-0815-16497, "Safety Classification of Passive Nuclear Power Plant Electrical Systems, Revision 1." Enclosure 1 contains thedesignation "Proprietary" at the top of each page containing proprietary information. Theinformation considered by NuScale to be proprietary is identified within double braces, "{{

}}" inthe document.

AF-0118-58310 Page 2 of 2(5)The basis for proposing that the information be withheld is that NuScale treats the information asa trade secret, privileged, or as confidential commercial or financial information. NuScale reliesupon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and9.17(a)(4).(6)Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided forconsideration by the Commission in determining whether the information sought to be withheldfrom public disclosure should be withheld:(a)The information sought to be withheld is owned and has been held in confidence by NuScale.(b)The information is of a sort customarily held in confidence by NuScale and, to the best ofmy knowledge and belief, consistently has been held in confidence by NuScale. Theprocedure for approval of external release of such information typically requires review bythe staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content,competitive effect, and determination of the accuracy of the proprietary designation.Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for theinformation, and then only in accordance with appropriate regulatory provisions orcontractual agreements to maintain confidentiality.(c)The information is being transmitted to and received by the NRC in confidence.(d)No public disclosure of the information has been made, and it is not available in publicsources. All disclosures to third parties, including any required transmittals to NRC, havebeen made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence.(e)Public disclosure of the information is likely to cause substantial harm to the competitiveposition of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The informationsought to be withheld is part of NuScale's technology that provides NuScale with acompetitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would bedifficult for others to duplicate the technology without access to the information sought to bewithheld.I declare under penalty of perjury that the foregoing is true and correct. Executed on February 2, 2018. _____________________________ Thomas A. Bergman

___________________________________________________________________Th o m as a a a a a a a a a a a a a a A. Ber g ma m m m m m m m m m m m m m m m n