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LLC Submittal of Technical Report NuScale Generic Technical Guidelines, TR-1117-57216
	ML17334B822
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Site:	NuScale
Issue date:	11/30/2017
From:	Bergman T; NuScale
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	LO-1117-57387 TR-1117-57216-NP
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LO-1117-57387 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com Docket: 52-048 November 30, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Technical Report NuScale Generic Technical Guidelines, TR-1117-57216

REFERENCES:

Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application, dated December 31, 2016 (ML17013A229).

NuScale Power, LLC (NuScale) has submitted a Design Certification Application for its Integral Small Modular Reactor design (Reference 1). The purpose of this letter is to provide NuScale Technical Report NuScale Generic Technical Guidelines, TR-1117-57216. This technical report will be incorporated by reference in a future revision of the NuScale Final Safety Analysis Report. contains the proprietary version of TR-1117-57216. NuScale requests that the proprietary version 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 TR-1117-57216.

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

I you have any questions, please contact Steve Mirsky at 240-833-3001 or at smirsky@nuscalepower.com if you have any questions.

Sincerely, Thomas A. Bergman Vice President, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury, NRC, OWFN-8G9A : NuScale Generic Technical Guidelines, TR-1117-57216-P, Revision 0, proprietary version : NuScale Generic Technical Guidelines, TR-1117-57216-NP, Revision 0, nonproprietary version : Affidavit of Thomas A. Bergman, AF-1117-57373

LO-1117-57387 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com :

NuScale Generic Technical Guidelines TR-1117-57216-P, Revision 0, proprietary version

LO-1117-57387 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com :

NuScale Generic Technical Guidelines TR-1117-57216-NP, Revision 0, nonproprietary version

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 Licensing Technical Report

NuScale Generic Technical Guidelines November 30, 2017 Revision 0 Docket: 52-048 NuScale Confidential, Proprietary Class 2 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 Licensing Technical Report

© Copyright 2017 by NuScale Power, LLC ii COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC, and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S.

Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

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

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 Licensing Technical Report

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

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© Copyright 2017 by NuScale Power, LLC iv CONTENTS Abstract....................................................................................................................................... 1 Executive Summary.................................................................................................................... 2 1.0 Introduction..................................................................................................................... 3 1.1 Purpose................................................................................................................. 3 1.2 Scope.................................................................................................................... 3 1.3 Abbreviations......................................................................................................... 3 2.0 Background..................................................................................................................... 6 3.0 Methodology.................................................................................................................... 7 4.0 Procedure Development................................................................................................. 9 4.1 Symptom-Based Procedures................................................................................ 9 4.2 Critical Safety Functions........................................................................................ 9 4.2.1 Containment Integrity.......................................................................................... 10 4.2.2 Reactivity............................................................................................................. 11 4.2.3 Core Heat Removal............................................................................................. 12 4.3 Structure and Use............................................................................................... 13 4.4 Manual Actions.................................................................................................... 18 4.5 Entry Conditions.................................................................................................. 24 4.6 Exit Conditions.................................................................................................... 24 4.7 Setpoint Selection............................................................................................... 25 5.0 Generic Technical Guidelines and Basis.................................................................... 26 5.1 Containment Integrity Safety Function................................................................ 27 5.1.1 Containment Atmosphere.................................................................................... 28 5.1.2 Containment Bypass........................................................................................... 29 5.1.3 Containment Isolation Signal Actuation............................................................... 32 5.1.4 Chemical and Volume Control System Isolation Actuation.................................. 34 5.1.5 Primary to Secondary Leakage........................................................................... 37 5.2 Reactivity Safety Function................................................................................... 39 5.2.1 Reactor Trip System Actuation............................................................................ 40 5.2.2 Reactivity Verification.......................................................................................... 41 5.2.3 Demineralized Water Isolation............................................................................. 43 5.3 Core Heat Removal Safety Function................................................................... 46

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5.3.1 Ultimate Heat Sink Inventory............................................................................... 47 5.3.2 Fuel Clad Protection............................................................................................ 48 5.3.3 Passive Cooling................................................................................................... 51 5.3.4 Emergency Core Cooling System (ECCS) Actuation.......................................... 52 5.3.5 Decay Heat Removal Actuation........................................................................... 55 5.3.6 Pressurizer Heater Trip Actuation........................................................................ 61 5.4 Defense-in-Depth Actions.................................................................................... 63 5.4.1 Electrical.............................................................................................................. 64 5.4.2 Other Defense-in-Depth Actions.......................................................................... 69 6.0 Validation....................................................................................................................... 71 7.0 References..................................................................................................................... 72 7.1 Source Documents.............................................................................................. 72 7.2 Referenced Documents....................................................................................... 72 8.0 Appendix A - Event Progression Comparison........................................................... 73 9.0 Appendix B - NuScale Design Severe Accident Management Guideline Development.................................................................................................................. 76 10.0 Appendix C - Instrumentation Utilized in the General Technical Guidelines.......... 79 TABLES Table 1-1 Abbreviations........................................................................................................... 3 Table 1-2 Definitions................................................................................................................ 5 Table 4-1 PRA credited operator actions to mitigate core damage........................................ 19 Table 4-2 PRA credited operator actions to mitigate large radiological release..................... 19 Table 4-3 Operator actions described in FSAR for Chapter 20 beyond-design-basis events 20 Table 4-4 Operator actions to support critical safety functions............................................... 21 Table 8-1 Loss of coolant accident......................................................................................... 73 Table 8-2 Steam generator tube failure.................................................................................. 74 Table 8-3 Loss of feedwater................................................................................................... 74 Table 8-4 Inadequate core cooling/loss of forced circulation................................................. 75 Table 9-1 EPRI Candidate high-level action comparison to NuScale design......................... 76 Table 10-1 List of reactor trip parameters................................................................................ 79 Table 10-2 List of general technical guidelines decision variables........................................... 80

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© Copyright 2017 by NuScale Power, LLC vi FIGURES Figure 4-1 Example safety function status display on the HSI.............................................. 16 Figure 4-2 Example safety function display on manual electronic interface......................... 17 Figure 4-3 Unit overview display........................................................................................... 24 Figure 5-1 Comparison of riser level vs. average RCS temperature for bounding cases..... 30 Figure 5-2 ATWS response under different conditions......................................................... 41

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Abstract This document describes the generic guidance that is provided to emergency procedure writers to develop procedures used by plant operators to ensure plant safety during an accident. This guidance covers the content provided in existing industry procedure classes including:

emergency operating procedures, severe accident management guidelines, extended loss of AC power and loss of large area guidance.

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Executive Summary TMI Action Plan Item I.C.1, a post-TMI requirement approved by the Commission for implementation, requires the preparation of emergency procedure technical guidelines for development of the emergency operating procedures (EOPs). Preparation of the technical guidelines is conducted in accordance with Clarification of TMI Action Plan Requirements, NUREG-0737, and Supplement 1, Requirements for Emergency Response Capability, NUREG-0737, that also specify submittal of the technical guidelines to the NRC for review and approval.

Meeting the requirements of TMI Action Plan Item I.C.1 as prescribed in NUREG-0737,Section I.C.1, and Supplement 1 to NUREG-0737, Section 7 is an acceptance criterion in Operating and Emergency Operating Procedures, SRP Chapter 13.5.2.1, NUREG-0800. Design-specific generic technical guidelines (GTGs) will be used by combined license applicants to develop their plant-specific technical guidelines, from which their EOPs will be developed. The GTGs are described in Guidelines for the Preparation of Emergency Operating Procedures, NUREG-0899.

The GTGs are documents that identify the equipment or systems to be operated and list the steps necessary to mitigate the consequences of transients and accidents and restore safety functions. The GTGs represent the translation of engineering data derived from transient and accident analyses and probabilistic risk assessment (PRA) into information presented in such a way that it can be used to write EOPs.

The NuScale Power, LLC plant design is both simple and passive with limited manual actions required to ensure nuclear safety. There are no operator actions credited in the evaluation of NuScale design-basis accidents. All of the analyzed operator actions are used to mitigate beyond-design-basis events.

This document describes the NuScale generic technical guidance, including:

1.

the process used to develop GTGs.

2.

the assumptions used during development.

3.

a description of the supportive analysis.

4.

a description of the validation process.

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1.0 Introduction 1.1 Purpose This document describes the generic guidance that is provided to the emergency procedure writers to develop procedures used by plant operators to ensure plant safety during an accident. This guidance covers the content provided in existing industry operating procedure classes including: emergency operating procedures, severe accident management guidelines, extended loss of AC power and loss of large area guidance.

1.2 Scope This document describes the actions assumed within analysis of accident conditions that occur outside the design basis. The NuScale transient and accident analyses conducted for design-basis events (DBEs) do not identify any required operator actions. The generic technical guidelines (GTGs) contain those actions that have been credited by probabilistic risk assessment (PRA) and described in the final safety analysis report (FSAR) to respond to beyond-design events such as severe accidents, extended loss of AC power, and loss of large area. Additionally, some actions have been included solely based on task analysis performed as part of the human factors engineering program.

This document is intended to provide guidance and basis for the emergency operating procedures but does not describe actions to the level of detail expected for a final operating procedure.

1.3 Abbreviations Table 1-1 Abbreviations Term Definition AAPS auxiliary AC power source ATWS anticipated transient without scram BDG backup diesel generator CET core exit thermocouple CFDS containment flood and drain system CHLA candidate high-level actions CIS containment isolation signal CIV containment isolation valve CNV containment vessel CVCI chemical volume control isolation CVCS chemical and volume control system DBA design-basis accident DBE design-basis event DHRS decay heat removal system DW demineralized water DWIS demineralized water isolation signal DWS demineralized water system

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Term Definition ECCS emergency core cooling system EDSS highly reliable DC power system EOF emergency operations facility EOP emergency operating procedure ESFAS engineered safety features actuation system FSAR final safety analysis report GTG generic technical guideline HSI human-system interface I&C instrumentation and controls IHA important human action ISV integrated system validation LOCA loss of coolant accident LOCA-IC loss of coolant accident inside containment LOCA-OC loss of coolant accident outside containment (containment bypassed)

MCC motor control center MCR main control room MCS module control system MPS module protection system PCS plant control system PORV power operated relief valve PRA probabilistic risk assessment PZR pressurizer RCS reactor coolant system RPV reactor pressure vessel RRV reactor recirculation valve RTS reactor trip system RVV reactor vent valve SAMG severe accident management guidance SDI safety display and indication SGTF steam generator tube failure SME subject matter expert TMI Three Mile Island UHS ultimate heat sink UTB under the bioshield WR wide range

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Table 1-2 Definitions Term Definition defense-in-depth function A function defined by the plant design that verifies and restores nonsafety related equipment to mitigate beyond-design-basis events, provides additional margin to safety system actuation or maintains parameters within normal operating ranges. The defense-in-depth functions are: (1) electrical, (2) reactor coolant system (RCS) makeup, (3) secondary heat removal, and (4) post trip actions.

design basis accident A postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures and components necessary to ensure public health and safety.

design basis event Postulated events used in the design to establish the acceptable performance requirements for the structures, systems, and components.

safety function A function defined by the plant design that maintains the unit in a safe shutdown condition and provides mitigating actions for accidents that cause core damage or release of radioactivity. The NuScale design has three safety functions defined: (1) maintain containment integrity, (2) reactivity control, and (3) remove fuel assembly heat.

maintain containment integrity Design features used to prevent fission products from escaping the containment boundary. This function is referred to as containment integrity for ease of communication.

reactivity control Design features used to control reactivity to maintain core and RCS integrity in order to prevent the release of radioactive material to the environment.

remove fuel assembly heat Design features used to remove heat from the fuel assemblies via passive convection and conduction. This function maintains fuel integrity and prevents the release of radioactive material to the environment. This function is referred to as core heat removal for ease of communication.

Operations subject matter expert A person who has completed the NuScale human factors engineering (HFE)/Operations initial company training program, has previous licensed operating nuclear plant experience, and has performed task analysis or NuScale system reviews so they are familiar with the NuScale plant design.

secondary heat sink Use of the steam generators to accomplish core heat removal. Each of the two steam generators acts as a secondary heat sink to facilitate core heat removal. Heat removal can be accomplished by supplying feedwater to the steam generator where it is converted to steam and subsequently transferred to the circulating water system via the main condenser, or to the DHRS where core heat is transferred to the reactor pool.

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2.0

Background

The generic technical guidance is described in the Standard Review Plan, NUREG-0800, Chapter 13.5.2.1. The Standard Review Plan is the parent document providing guidance for the review of submittal documents and states that a procedures generation package should be submitted. The procedures generation package describes the applicants program for developing the emergency operating procedures as well as the content. Vendor-supplied generic technical guidance is developed to be used by applicants or licensees to develop emergency operating procedures.

The Three Mile Island accident in 1979 provided the impetus for renewed guidance and clarity on the development of emergency operating procedures. The nuclear steam supply system vendors at the time developed generic guidance that was then given to the plant owners to develop specific plant technical guidance. NUREG-0899, Guidelines for the Preparation of Emergency Operating Procedures, provides guidance that the U.S. Nuclear Regulatory Commission uses in evaluating whether an applicant or licensee meets the requirements for Emergency Operating Procedures of Title 10 Code of Federal Regulations, Part 50.34(b)(6)(ii).

The NuScale GTGs have been developed by referring to the content and format that legacy vendor guidelines have used. Traditionally, generic guidance has been arranged in four volumes: (1) the actual guidelines, (2) an annotated version of the guidelines with reference to a basis document, (3) the guidelines basis, and (4) an implementation guide. NuScale is presenting the GTGs by consolidating the content that would traditionally have been contained in Volumes 1 through 3 into a single volume. An implementation guide is not included with this current guidance, but may be developed later when a combined operating license applicant is identified.

Additionally, four events are typically described in past vendor-submitted generic technical guidance: (1) loss of coolant events, (2) steam generator tube rupture events, (3) loss of feed events, and (4) inadequate core cooling events. The NuScale generic technical guidance supports a symptom-based approach and is, therefore, not event specific.

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3.0 Methodology Technical guidance is necessary to develop the emergency operating procedures that operators use during accident conditions. Multiple inputs are evaluated to determine the appropriate mitigating strategies and actions based on plant design. The following topics are reviewed to include any assumed operator actions:

1. The plant design bases response to DBEs as described in FSAR Chapter 15

2. The instrumentation and controls(I&C) failure defense-in-depth analysis as described in FSAR Chapter 7

3. The operator actions assumed in the beyond-design-basis PRA as described in FSAR Chapter 19

4. The operator actions assumed in beyond-design-basis event evaluation as described in FSAR Chapter 20

5. Multi-unit design considerations as described in FSAR Chapter 21

6. Plant technical specifications as described in Part 4 of the NuScale design certification application (DCA)

7. System requirements and limitations as defined in system description documents

8. HFE task analyses results as described in FSAR Chapter 18.4 and its associated reference Single operator errors of both omission and commission were considered and analyzed in the transient and accident analysis of FSAR Chapter 15. The NuScale design limits operator error consequences to be less severe than the worst-case component single failure. The NuScale plant is a passive design that requires no operator action for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for any design basis event. Automated actions place and maintain the unit in a safe-state for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after a DBE even with assumed failures. Operator actions directed by procedure make the consequences of an event less severe, but failure to take one of these actions cannot make the consequences worse than the bounding FSAR Chapter 15 analysis.

Safety analysis of DBEs are performed from event initiation until a safe, stabilized condition is reached. A safe, stabilized condition is reached when the initiating event is mitigated, the FSAR Chapter 15 acceptance criteria are met, and system parameters are trending in the favorable direction. No operator action is required to reach or maintain a safe, stabilized condition for any DBE.

Instrumentation and controls (I&C) diversity and defense-in-depth analysis documented in FSAR Chapter 7 did not result in any required operator actions.

Multiple operator errors or errors that result in common mode failures are beyond-design-basis and analyzed in the PRA.

The PRA assessment documented in FSAR Chapter 19 identified seven human actions to prevent core damage in beyond-design-basis accident sequences. These actions are listed in Section 4.4 of this report.

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Additional evaluations for beyond design-basis events, performed and documented in FSAR Chapter 20, identified two human actions that would mitigate core damage once it has occurred or minimize the radioactive release. These actions are also listed in Section 4.4 of this report.

Other actions have been included within the generic technical guidelines that do not originate from analysis but are considered prudent and promote consistent operator performance. For instance, the PRA analysis only credits manual initiation of emergency core cooling system (ECCS) or containment isolation signal (CIS) when those systems fail to actuate properly. Any failed engineered safety features actuation system (ESFAS) actuation will be manually actuated to provide prudent and consistent guidance for operators.

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4.0 Procedure Development 4.1 Symptom-Based Procedures The structure of the guidelines is symptom-based. Symptom-based procedures are used to allow the operator to respond directly to the indications presented as part of the accident progression. Symptom-based procedures do not require the operator to attempt to diagnose the accident in progress. Symptom-based procedures allow the operator to respond to an event without knowledge of the initiating event or equipment status. These procedures also allow the operator to respond to unanticipated events, because they evaluate key parameters and direct actions to maintain them within the prescribed limits rather than responding in a predetermined sequence based on a diagnosed accident.

Legacy generic guidelines have included event-based descriptions. These events were based on the transient and accident analysis events and associated operator actions described in those designs Final Safety Analysis Report Chapter 15. Because the NuScale design has no FSAR Chapter 15 manual actions credited, the symptom-based approach allows for mitigating strategies to be effective with multiple failures regardless of the combination.

4.2 Critical Safety Functions The evaluation of symptoms is grouped into critical safety functions. This guidance is developed to maintain critical safety functions for the NuScale plant design. Evaluation of the NuScale design, in addition to performing a comparison with traditional light water reactor safety functions, was used to determine the appropriate NuScale safety functions. These functions are accomplished by maintaining the following:

containment integrity reactivity core heat removal Additional safety functions are not needed due to the simplicity and reliance on passive systems in the NuScale design. For example, current fleet pressurized-water reactors (PWRs) typically have a critical safety function of maintaining a secondary heat sink.

Heat sink maintenance exists in other PWR designs because its loss can lead to core damage. Timely assessment and recovery or mitigation is critical to preventing core damage and, therefore, a separate critical safety function is warranted. The key difference is that in the NuScale design, loss of secondary heat sink, by itself, does not result in core damage. Mitigation of a complete loss of secondary heat sink has been analyzed as part of PRA and, as such, is a best estimate analysis. This analysis demonstrates that loss of secondary heat sink is mitigated passively with the reactor safety valves and heat removal through containment. The ECCS is also fully capable of removing decay heat in all required operating conditions.

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NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 18 4.4 Manual Actions There are limited operator actions required in the NuScale design. All of the analyzed operator actions are used to mitigate beyond-design-basis events. There are no operator actions credited in the evaluation of NuScale DBEs. Automated actions place and maintain the unit in a safe state for at least 72 hours after a DBE even with assumed failures. Operator errors of both omission and commission were analyzed. Operator actions directed by guidance are intended to mitigate the consequences of the event. Failure to take action cannot make the consequences more severe than the bounding FSAR Chapter 15 analysis. The I&C analysis of diversity and defense-in-depth identified no manual operator actions for a DBE. ((

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The PRA analysis examined beyond-design-basis events that lead to core damage and identified seven human actions, including two that are important human actions (IHAs) that are highlighted in Table 4-1. A single action to mitigate large radiological releases is listed separately in Table 4-2; although it is identical to the first item listed in Table 4-1.

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 19 Table 4-1 PRA credited operator actions to mitigate core damage Name Description Context CFDS-HFE-0001C-FOP-N Operator fails to unisolate and initiate CFDS injection (IHA) Used for LOCA-OC, SGTF, and transients important human action CVCS-HFE-0001C-FOP-N Operator fails to unisolate and initiate CVCS injection (IHA) Used for LOCA-IC, LOCA-OC (letdown), transients and secondary steam line break, upon failure of ECCS, and SGTF important human action CNTS-HFE-0001C-FTC-N Operator fails to isolate containment Backup action to MPS auto function failure CVCS-HFE-0002C-FOP-N Operator fails to locally unisolate and initiate CVCS injection Local unisolation due to lack of control from a partial loss of DC power ECCS-HFE-0001C-FTS-N Operator fails to open ECCS valves Backup action to MPS auto function failure EHVS-HFE-0001C-FTS-N Operator fails to start/load combustion turbine generator Backup local action to control room initiation failure during loss of offsite power ELVS-HFE-0001C-FTS-N Operator fails to start/load backup diesel generator Backup local action to control room initiation failure during loss of offsite power The important human actions of injecting water into the NuScale Power Module are successful to mitigate core damage when taken as assumed by supporting PRA analysis. When emergency makeup is initiated, the top of active fuel may be uncovered for a period of time, but peak cladding temperature remains below the temperature at which damage occurs (Reference 7.2.1, Section 19.2.5). Table 4-2 PRA credited operator actions to mitigate large radiological release Name Description Context CFDS-HFE-0001C-FOP-N Operator fails to unisolate and initiate CFDS injection (IHA) Used for LOCA-OC, SGTF, and transients important human action Additional evaluations for beyond-design-basis events, performed and documented in FSAR Chapter 20, identified the following two actions not covered in other programs that are meant to prevent core damage during an extended loss of AC power (ELAP), or to mitigate core damage and minimize the spread of radioactive release during a loss of large area (LOLA) event.

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 20 Table 4-3 Operator actions described in FSAR for Chapter 20 beyond-design-basis events Name Description Context LOLA phase 2/ELAP phase 3 Add inventory to the UHS through the spent fuel pool assured makeup line Long-term extended loss of AC power action (> 30 days) LOLA phase 3 Mitigate damage to fuel in the reactor vessel and radiological release - evaluate safety functions, provide a means for water spray scrubbing using fog nozzles and available water sources. Address runoff water containment issues (sandbags, dikes, etc.) Long-term action loss of large area to support a reduced emergency planning zone Firefighting activities are not included within the generic technical guidelines and are addressed through other procedures. This practice is consistent with the current industry practice where separate fire response plans and procedures are maintained specific to each site. In addition to the actions specifically required by analysis, manual actions have been included when deemed appropriate by subject matter expert experience or to maintain consistency among mitigating strategies. For example, the PRA only assumes manual actuation of two of seven ESFAS functions. For consistent performance and standardization for operators, when any ESFAS function does not actuate as designed, the manual action is taken to actuate that function from the control room. Manual actions are prioritized based on both the time allowed to perform the action and the location where the action is taken. The allowed times are based on PRA analysis of operator actions to mitigate core damage or large radiological releases. The following examples demonstrate how this prioritization is taken into account: Actions in the main control room are prioritized over actions taken locally in the plant. Actions with the shortest analyzed time for success will be prioritized first. Table 4-4 provides a comprehensive list of all actions determined to be included in the generic technical guidance from sources described in Tables 4-1, 4-2, 4-3, and from subject matter expert task analysis. These are grouped by critical safety function.

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 24 4.5 Entry Conditions (( Figure 4-3 Unit overview display 4.6 Exit Conditions The generic technical guidance is intended to inform the writing of the emergency operating procedures to enable operators to mitigate accident conditions. The GTGs are exited and normal operating procedures commence when the event has been mitigated to the extent that parameters allow return to normal operation. }}2(a),(c)

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Instrumentation requirements have been provided based on NuScale requirements, regulatory requirements, or vendor recommendations, but must be refined once the actual instrumentation is selected or purchased. Emergency procedures developed from these GTGs will need to reference the plant specific equipment values, ranges, and accuracies.

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The GTGs are arranged by safety function and are meant to be used by utilizing the flowchart in conjunction with the text. Each section contains a flowchart showing the functional logic and is accompanied by a narrative separated into the logic decision and endpoints. The decision points are described first for that function followed by the manual actions supporting that function.

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NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 71 6.0 Validation This revision of the GTGs is issued with the understanding that the selected setpoints are based on analysis and engineering judgement. Instrument selection is not complete in all cases and final instrument selection may impact the listed setpoints as tolerances and accuracy are considered. This revision of the GTGs has been incorporated into the NuScale simulator. It has undergone limited scenario testing to ensure the expected operator cues prompt operators to perform the credited actions within expected times, but is not considered to be validated. The current actions and logic have been used in simulations to provide high confidence in the mitigating strategies. The generic technical guidance provided in this document will be validated during the performance of the NuScale integrated system validation testing. An update to the generic technical guidelines will occur to document any changes resulting from the integrated system validation testing.

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 72 7.0 References 7.1 Source Documents 7.1.1 U.S. Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the TMI-2 Accident, NUREG-0660, May 1980. 7.2 Referenced Documents 7.2.1 NuScale Power Standard Plant Design Certification Application, Rev. 0. 7.2.2 U.S. Nuclear Regulatory Commission, Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980. 7.2.3 U.S. Nuclear Regulatory Commission, Requirements for Emergency Response Capability, NUREG-0737, Supplement No. 1, December 1982. 7.2.4 U.S. Nuclear Regulatory Commission, Standard Review Plan, Operating and Emergency Operating Procedures, NUREG-0800, Rev. 2, Chapter 13, Section 13.5.2.1. 7.2.5 U.S. Nuclear Regulatory Commission, Guidelines for the Preparation of Emergency Operating Procedures, NUREG-0899, August 1982. 7.2.6 Combustible Gas Control, TR-0716-50424, Rev. 0. 7.2.7 Electric Power Research Institute, Severe Accident Management Guidance Technical Basis Report Volume 1: Candidate High-Level Actions and Their Effects, EPRI #1025295, EPRI, Palo Alto, CA, 2012. 7.2.8 Electric Power Research Institute, Severe Accident Management Guidance Technical Basis Report Volume 2: The Physics of Accident Progression, EPRI

1025295, EPRI, Palo Alto, CA, 2012.

7.2.9 Electric Power Research Institute, Steam Generator Management Program: PWR Primary-to-Secondary Leak GuidelinesRevision 4, EPRI #1022832, EPRI, Palo Alto, CA, 2011. 7.2.10 Electric Power Research Institute, Hydrogen Detection in Nuclear Power Plants: Comparison of Potential, Existing, and Innovative Technologies, EPRI

3002002107, EPRI, Palo Alto, CA, 2013.

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 73 8.0 Appendix A - Event Progression Comparison This appendix is intended to illustrate the differences in how a loss of coolant accident, steam generator tube rupture, loss of feedwater, and inadequate core cooling are mitigated between existing PWR technology and the NuScale design. The list of actions is provided at a high level for illustrative purposes only. These compare events between the two designs for those events within the scope of design-basis events. Table 8-1 Loss of coolant accident Typical PWR Mitigating Strategy/Actions NuScale Mitigating Strategy/Actions Maximize safety injection flow into the RCS There is no injection phase Manually isolate the source of RCS leakage Containment Isolation occurs automatically Manually control RCS pressure to maintain subcooling RCS pressure control is not required. Passive cooling systems are designed so that subcooling is not jeopardized Manually control RCS inventory to maintain subcooling No additional makeup is necessary as long as the RCS leak is being captured in containment Manually perform a controlled cooldown using the steam generators DHRS passively performs cooldown within design cooldown rates Manually secure RCS recirculation pumps No recirculation pumps Manually align shutdown cooling system if RCS leakage is small or has been isolated ECCS occurs automatically to provide additional passive cooling Transition from injection to containment sump recirculation There is no injection phase and recirculation occurs naturally using the ECCS Manually alternate hot and cold leg recirculation There is no injection phase Manual actions required No manual actions required

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 74 Table 8-2 Steam generator tube failure Typical PWR Mitigating Strategy/Actions NuScale Mitigating Strategy/Actions Manually confirm the affected steam generator Not required as both are isolated Manually perform a controlled cooldown using the steam generators DHRS passively performs cooldown within design cooldown rates Manually isolate the most affected steam generator Decay heat removal occurs automatically and isolates both steam generators Manually control RCS pressure to maintain subcooling and match steam generator pressure RCS pressure quickly matches DHRS pressure and effectively stops the RCS leak in the affected steam generator Manually control RCS inventory to maintain subcooling DHRS passively performs cooldown within design cooldown rates Manually secure RCS recirculation pumps as needed No recirculation pumps exist Manually align shutdown cooling system if RCS leakage is small or has been isolated ECCS occurs automatically to provide additional passive cooling if necessary Manual actions required No manual actions required Table 8-3 Loss of feedwater Typical PWR Mitigating Strategy/Actions NuScale Mitigating Strategy/Actions Manually minimize heat input to the RCS by securing RCS recirculation pumps No recirculation pumps exist Manually conserve steam generator inventory by isolating steam generator blowdowns Decay heat removal occurs automatically that isolates the inventory in both steam generators Manually regain a source of feedwater to at least one steam generator Decay heat removal occurs automatically and isolates both steam generators and retains it in the DHRS Manually control RCS pressure to maintain subcooling RCS pressure control is not required. Passive cooling systems are designed so that subcooling is not jeopardized Manually control RCS inventory to maintain subcooling No additional RCS inventory is required If feed cannot be restored, manually align injection to the RCS and open PORVs for once through core cooling ECCS occurs automatically to provide additional passive cooling if necessary Manual actions required No manual actions required

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 75 Table 8-4 Inadequate core cooling/loss of forced circulation Typical PWR Mitigating Strategy/Actions NuScale Mitigating Strategy/Actions Verify natural circulation established No action, natural circulation is the normal cooling mechanism Manually perform a controlled cooldown using the steam generators DHRS passively performs cooldown within design cooldown rates Manually control RCS pressure to maintain subcooling RCS pressure control is not required. Passive cooling systems are designed so that subcooling is not jeopardized Manually control RCS inventory to maintain subcooling No additional RCS inventory is required Manually restore electrical power to de-energized AC buses No AC power is required to obtain safe shutdown conditions Manually start RCS recirculation pumps as needed No recirculation pumps exist Manually align shutdown cooling system ECCS occurs automatically to provide additional passive cooling if necessary Manual actions required No manual actions required

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 76 9.0 Appendix B - NuScale Design Severe Accident Management Guideline Development As part of development of the severe accident management guidelines, EPRI Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects (Reference 7.2.7) was reviewed and the 20 candidate high-level actions (CHLAs) identified in the report were categorized by their applicability to the NuScale design. Given the unique design features of the NuScale Power Module, the applicability review was not limited to PWR actions only. Table 9-1 EPRI Candidate high-level action comparison to NuScale design }}2(a),(c) ((

NuScale Generic Technical Guidelines TR-1117-57216-NP Rev. 0 © Copyright 2017 by NuScale Power, LLC 79 10.0 Appendix C - Instrumentation Utilized in the General Technical Guidelines The following instrumentation is used as inputs to the RTS. A valid reactor trip signal is processed with two of four of the input channel parameters exceed the analytical limit setpoint. A reactor trip is the initiator for entry into the generic technical guidance. The table lists the parameter, limit, number of channels, and the actuation logic with an arrow indicating if the parameter is rising or lowering to actuate. Each of the reactor trip parameters are described in the Safety Analysis Analytical Limits Report (Reference 7.2.1, Table 7.1-4). Table 10-1 List of reactor trip parameters Parameter Analytical Limit Channels Logic High Power Range Linear Power High-1 = 25% RTP High-2 = 120% RTP 4 2/4 High Intermediate Range Log Power Rate 3 dpm 4 2/4 High Power Range Positive and Negative Rate +/- 15% RTP/minute 4 2/4 High Source Range Count Rate 5x105 cps 4 2/4 High Source Range Log Power Rate 3 dpm 4 2/4 High Range RCS Thot (NR RCS Thot) 610°F 4 2/4 High Range Containment Pressure 9.5 psia 4 2/4 High Pressurizer Pressure 2000 psia 4 2/4 Low Pressurizer Pressure 1720 psia (when Thot > 600°F) 4 2/4 Low Low Pressurizer Pressure 1600 psia 4 2/4 High Pressurizer Level 80% 4 2/4 Low Pressurizer Level 35% 4 2/4 High Main Steam Pressure 800 psia 4 2/4 Low Main Steam Pressure 300 psia 4 2/4 Low Low Main Steam Pressure 100 psia 4 2/4 High Steam Superheat (MS Temperature and Pressure) 150°F 4 2/4 Low Steam Superheat (MS Temperature and Pressure) 0.0°F 4 2/4 Low Low RCS Flow 0.0 ft3/sec 4 2/4 Low ELVS 480 VAC to EDSS Battery Chargers Actuation Delay of 60 seconds 4 2/4 High Under-the-Bioshield Temperature 250°F 4 2/4 The following parameters are used within the generic technical guidance as decision points for the operators. The parameters are arranged by safety function and, in most cases, are the same as the actuation setpoints for the associated ESFAS functions.

LO-1117-57387 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com : Affidavit of Thomas A. Bergman, AF-1117-57373

AF-1117-57373 Page 1 of 2 NuScale 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, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for 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 to withhold 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 a license 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 the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors 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, production capabilities, budget levels, or commercial strategies of NuScale. (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 to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the process by which NuScale develops its generic technical guidelines. NuScale has performed significant research and evaluation to develop a basis for this process 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 NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled NuScale Generic Technical Guidelines. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies

AF-1117-57373 Page 2 of 2 upon 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) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from 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 of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the 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 the information, and then only in accordance with appropriate regulatory provisions or contractual 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 public sources. All disclosures to third parties, including any required transmittals to NRC, have been 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 competitive position 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 information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on November 30, 2017. Thomas A. Bergman}}

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