ML20356A018
| ML20356A018 | |
| Person / Time | |
|---|---|
| Site: | 99902049 |
| Issue date: | 12/18/2020 |
| From: | Kenny F Holtec, SMR |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20356A015 | List: |
| References | |
| 160-USNRC-005 HI-2201064-NP, Rev 1 | |
| Download: ML20356A018 (40) | |
Text
SMR, LLC 160 Sponsoring Company Project No.
Company Record Number HI-2201064 1
Revision No.
18 Dec 2020 Issue Date Record Type Report Copyright Proprietary Classification Nuclear Quality Class A Holtec International Company Record
Title:
Non Proprietary Version of Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria Prepared by:
Reviewed by:
Approved by:
Export Control Status If Export Control is noted to be applicable to this record, control and distribution shall comply with all applicable laws, governmental rules and regulations, acts, orders, directives, guidance, or other statement by any legal or regulatory authority ³Laws' including, but not limited to, Laws related to export administered by the U.S. Department of Energy ³DOE' under 10 CFR Part 810 (the ³DOE Export Regulations' the U.S.
Nuclear Regulatory Commission ³NRC' or the U.S. Department of Commerce ³DOC' Any entity in possession of this record is obligated to fully comply with all such Laws.
Proprietary Classification This record does not contain confidential or Proprietary Information. The Company reserves all copyrights.
Signature histories are provided here for reference only. Company electronic signature records are traceable via the provided Verification QR Code and are available for review within the secure records management system. A valid Verification QR Code and the presence of this covering page indicates this record has been approved and accepted.
F.Kenny, 18 Dec 2020 T.Morin, 18 Dec 2020 R.Trotta, 18 Dec 2020 Yes Export Control Applicability
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 1 of 39 Revision Log Revision Description of Changes 0
Initial Issue.
1 Added clarifications based on page turn conducted with U.S. NRC on 12 2020. Proprietary designations added throughout. Revisions are indicated in the right hand margin.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 2 of 39 Executive Summary This report provides the requirements and the regulatory basis to eliminate the need to analyze a Large Break Loss of Coolant Accident (LOCA) for the SMR-160 Reactor Coolant System (RCS). An overall description of the RCS and its design requirements are provided. The reactor pressure vessel (RPV) and the reactor coolant boundary of the steam generator (SG) with the integral pressurizer (PZR) are designed and analyzed as one pressure vessel with no connecting piping. The RPV/SG connection and the SG riser are integral parts of one pressure vessel. The largest reactor coolant pressure boundary (RCPB) piping in the SMR-160 is ((
xxxxx )). Within the SMR-160 design, breaks of (( )) do not result in the large initial inventory loss and rapid depressurization of the RCS associated with large break LOCAs.
This report also provides a description of the Passive Core Cooling System (PCCS) and the Passive Containment Heat Removal System (PCHR), along with a qualitative description of the plant response to a LOCA.
Finally, this report establishes the LOCA acceptance criteria and the basis for how these criteria are more restrictive than the requirements in 10 CFR 50.46. The established LOCA acceptance criteria for the SMR-160 are that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or that fuel cladding temperature shall be maintained within the normal operating temperature range.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 3 of 39 Table of Contents Revision Log................................................................................................................................. 1 1.0 Acronyms and Abbreviations............................................................................................ 5 2.0 Introduction....................................................................................................................... 6 2.1 Purpose................................................................................................................... 6 2.2 Objective................................................................................................................. 6 3.0 Technical Evaluation of the Reactor Coolant System....................................................... 6 3.1 General Overview of Reactor Coolant System........................................................ 6 3.2 Design Requirements.............................................................................................. 8 3.3 RPV/SG Connection................................................................................................ 9 3.4 SG Riser................................................................................................................ 10 3.5 Identification of Potential Break Sizes and Locations............................................ 11 4.0 Description of PCCS and PCHR..................................................................................... 12 4.1 General Overview.................................................................................................. 13 4.2 Safety Requirements for the PCCS and PCHR.................................................... 17 4.3 Operational Requirements for the PCCS and PCHR During a LOCA................... 19 5.0 Technical Evaluation of LOCA Acceptance Criteria........................................................ 19 5.1 Emergency Core Cooling System Acceptance Criteria (10 CFR 50.46)............... 20 5.2 SMR-160 LOCA Acceptance Criteria.................................................................... 20 6.0 Description of a Design Basis LOCA.............................................................................. 21 6.1 Event Initiation....................................................................................................... 21 6.2 Event Detection and Mitigation.............................................................................. 21 6.3 Long-Term Cooling................................................................................................ 21 7.0 Regulatory Evaluation as Pertains to Elimination of Large Break LOCA and Establishment of LOCA Acceptance Criteria.............................................................................. 22 7.1 Applicable 10 CFR 50 Regulations and General Design Criteria.......................... 22 7.2 Applicable Regulatory Guides............................................................................... 33 7.3 Applicable NUREG-0800 Sections........................................................................ 35 8.0 Summary and Conclusions............................................................................................. 37 9.0 References...................................................................................................................... 39
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 4 of 39 List of Figures Figure 3-1: Configuration of Combined Vessel............................................................................. 7 Figure 3-2: General Arrangement of Reactor Coolant System with Flow Arrows......................... 9 Figure 3-3: RPV/SG Connection with Weld Joint........................................................................ 10 Figure 3-4: SG Riser Weld to the Tubesheets............................................................................ 11 Figure 4-1: PCCS and Containment General Layout.................................................................. 14 Figure 6-1: Sequence of events for a typical LOCA.................................................................... 22 List of Tables Table 3-1 Potential Break Sizes, By System and Type.............................................................. 12
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 5 of 39 1.0 ACRONYMS AND ABBREVIATIONS Term Definition ABS Auxiliary Boiler System ADS Automatic Depressurization System AR Annular Reservoir ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel CFR Code of Federal Regulations CISO Containment Isolation System CS Containment Structure CES Containment Enclosure Structure COL Combined License DBA Design Basis Accidents dc Direct current DC Design Certification DEG Double-ended Guillotine ECCS Emergency Core Cooling System FLEX Diverse and Flexible Coping Strategies ITAAC Inspections, Tests, Analysis, and Acceptance Criteria LOCA Loss of Coolant Accident NRC Nuclear Regulatory Commission PCCS Passive Core Cooling System PCHR Passive Containment Heat Removal System PCMWS Passive Core Makeup Water System PCMWT Passive Core Makeup Water Tank PCS Plant Control System PDHR Primary Decay Heat Removal System PORV Power Operated Relief Valve PSS Plant Safety System PWR Pressurized Water Reactor PZR Pressurizer RCPB Reactor Coolant Pressure Boundary RCSP Reactor Coolant Startup Pumps RG Regulatory Guide RPV Reactor Pressure Vessel SDHR Secondary Decay Heat Removal System SG Steam Generator SMR Small Modular Reactor SRP Standard Review Plan SSC Structures, Systems, and Components SWS Service Water System TS Technical Specifications
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 6 of 39
2.0 INTRODUCTION
2.1 Purpose The purpose of this report is to:
- 1. Provide the design requirements and regulatory basis for the SMR-160 Reactor Coolant System (RCS) to substantiate the elimination of the need to postulate a large break Loss of Coolant Accident (LOCA);
- 2. Provide a functional description of the Passive Core Cooling System (PCCS) and Passive Containment Heat Removal System (PCHR) and describe how these systems mitigate a LOCA event in the SMR-160 design; and
- 3. Establish acceptance criteria for medium and small break LOCA which bound the acceptance criteria in 10 CFR 50.46.
2.2 Objective The objective of this report is to receive:
- 1. NRC approval of the SMR-160 RCS design requirements and regulatory basis such that a large break LOCA is eliminated from the design basis.
- 2. NRC approval of the SMR-160 LOCA acceptance criteria.
3.0 TECHNICAL EVALUATION
OF THE REACTOR COOLANT SYSTEM 3.1 General Overview of Reactor Coolant System The SMR-160 is a single loop, 160 MWe, natural circulation, small modular reactor based on pressurized water reactor (PWR) technology. The reactor coolant system (RCS) features an integral primary heat transfer circuit instead of the traditional multi-loop circuit of large PWRs.
The basic philosophy of the RCS is to use natural circulation to promote heat transfer from the core to the steam generator, while using simple, diverse safety systems which employ natural, passive phenomena for emergency core cooling. The basic design principles used in the design of the RCS are as follows:
Use natural circulation to provide coolant flow in the RCS Eliminate large piping connections Eliminate penetrations in the lower two-thirds of the reactor pressure vessel, such that all the penetrations are above the top of active fuel Provide ample water above the core to prevent uncovering the core in the event of a LOCA Use proven materials for vessel construction Design pressure vessels in accordance with ASME BP&V Code The major components of the RCS include the reactor pressure vessel (RPV), steam generator (SG) with integral pressurizer (PZR), and the reactor coolant startup pumps (RCSPs). The offset
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 7 of 39 configuration of the SG and RPV (see Figure 3-1) enables the use of traditional external control rod drive mechanisms and refueling equipment. The SMR-160 PZR is integral to the SG, with an internal divider plate eliminating the need for large surge piping.
In order to facilitate natural circulation, the elevation difference between the top of the steam generator and the core has been maximized, resulting in a tall column of water (chimney) above the core. With no penetrations in the lower two thirds of the RPV, the water column in the chimney ensures sufficient reactor coolant inventory above the core which always ensures fuel cooling in the event of a LOCA. The RPV is a thick-walled cylindrical pressure vessel with an integrally welded bottom head and a removable flanged top head (see Figure 3-1). The RPV contains fuel assemblies, control rods, instrumentation, and other internals necessary for core support and directing reactor coolant flow. The upper portion of the RPV shell is equipped with a tapered hub, reverse flange. There are no penetrations in the lower two thirds of the vessel, eliminating the potential for pipe breaks that could lead to uncovering the core. The SG is a vertical shell and tube, once-through steam generator. Hot reactor coolant passes through a riser located in the center of the SG tube bundle and then passes down through tubes to reject thermal energy to the secondary fluid on the shell side. The use of an integral central riser eliminates loop piping typically required for once through steam generators to bring reactor coolant to the top of the tube bundle. The straight SG tubes provide easy access for in-service inspection and tube plugging as needed. The SG removes heat from the RCS during power operation and during initial cooldown after a shutdown or a reactor trip.
(a) Cutaway View (b) Isometric View Figure 3-1: Configuration of Combined Vessel
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 8 of 39 3.2 Design Requirements The RPV and the reactor coolant pressure boundary of the SG with the integral PZR (hereinafter designated as the Combined Vessel) are designed and analyzed as one ASME Boiler and Pressure Vessel (B&PV) Code Section III, Subsection NB Class 1 pressure vessel with no connecting piping. The design requirements for the Combined Vessel are as follows:
((
))
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 9 of 39 Figure 3-2: General Arrangement of Reactor Coolant System with Flow Arrows 3.3 RPV/SG Connection The SG is connected directly to the RPV (Figure 3-1) by a single forging with concentric fluid flow paths (see Figure 3-2). Coolant heated by the core flows through the inner duct (hot leg) and coolant returning to the core flows through the outer annulus (cold leg). The concentric flow paths eliminate large diameter coolant piping (hot leg and cold leg) between the RPV and SG. ((
))
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 10 of 39 The RPV/SG Connection is considered an integral part of one pressure vessel where the RPV is welded directly to the SG in accordance with ASME B&PV Code Section III, Subsection NB, Subarticle 3350 and Article 4000 code using a Category (( )) weld joint (Figure 3-3). Since no piping exists between the two vessels, a postulated pipe break is non-credible. Also, a break at this location in the Combined Vessel is non-credible since it is designed to the requirements described in Section 3.2.
((
))
Figure 3-3: RPV/SG Connection with Weld Joint 3.4 SG Riser The SG is designed for reactor coolant on the tube side and feedwater/steam on the shell side.
The SG features a tall riser that carries the hot coolant from the RPV to the primary flow turn-around region above the top tubesheet. Reactor coolant then flows down through the tubes, rejecting thermal energy to the secondary coolant and producing steam in a typical counter current flow arrangement.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 11 of 39 The riser is made from a low alloy forging and is welded to the top and bottom tubesheet (see Figure 3-4) in accordance with ASME B&PV Code Section III, Subsection NB, Subarticle 3350 and Article 4000 code using a Category (( )) weld joint (Figure 3-4). Hence, the riser forms an integral part of the SG. Therefore, postulating a break in the Combined Vessel at this location is non-credible when the vessel is designed to the requirements described in Section 3.2.
((
))
Figure 3-4: SG Riser Weld to the Tubesheets 3.5 Identification of Potential Break Sizes and Locations The reactor coolant system has piping and piping connections to other systems to support various plant operations. The maximum piping size is (( )) diameter nominal.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 12 of 39 3.5.1 Potential Break Sizes The largest potential break size for a Loss of Coolant Accident (LOCA) in the SMR-160 is based on a full double-ended guillotine (DEG) rupture of the largest pipe in the RCPB. Table 3-1 shows the RCS penetrations and their associated pipe size. The largest possible break is a ((x xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx)). The reason a (( xxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )) being the largest break that needs to be analyzed, the SMR-160 design has effectively eliminated the Large Break LOCA.
For the purposes of this report, the Large Break LOCA is defined in a classical sense as ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx)).
3.5.2 Potential Break Locations Potential break locations in the SMR-160 design include any piping, nozzle, or instrumentation penetrations in or connected to the RCS that are part of the RCPB. Breaks in the RPV/SG connection and the SG riser, as described above, are non-credible. The Direct Vessel Injection (DVI) piping connections to the RPV, with an elevation at least (( )) above the TAF, are at the lowest elevation for a potential break. Table 3-1 below indicates which potential breaks are DEG. A DEG break has potentially two break flow paths from the RCS. For example, a break from the (( )) could have flow from both ends of the break since this system forms a closed loop with the RCS. The breaks that are not DEG only have a single connection to the RCS.
Table 3-1 Potential Break Sizes, By System and Type RCS and Interfacing Systems Nominal pipe size [in]
DEG
((
))
4.0 DESCRIPTION
OF PCCS AND PCHR All passive safety features are integrated with the containment, which incorporates heat exchange devices to accommodate removal of core decay heat from the reactor coolant system (RCS) to containment and from the containment to the annular reservoir then the environment.
As a result, during a LOCA the RCS and containment of the SMR-160 interact in a more integral fashion than in current generation light-water reactor designs. The design and the safety
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 13 of 39 analyses of the SMR-160 takes into consideration the interactions among the RCS, containment, and the passive safety features.
4.1 General Overview 4.1.1 Passive Core Cooling System The PCCS is designed to provide emergency core heat removal and makeup water during postulated Design Basis Accidents (DBAs). The system uses passive means such as natural circulation, gravity injection, and compressed gas expansion for core makeup and cooling without the use of active components such as pumps for the cooling process. The PCCS consists of the following sub-systems:
Primary Decay Heat Removal System (PDHR)
Secondary Decay Heat Removal System (SDHR)
Automatic Depressurization System (ADS)
Passive Core Makeup Water System (PCMWS)
(( )). The PCCS sub-systems are designed to mitigate DBAs (( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxx )). In concert with the Passive Containment Heat Removal System (PCHR) and Containment Isolation System (CISO), the PCCS ensures safe and stable shutdown conditions can be maintained and decay heat removal occurs for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without power, make-up water, or operator actions.
Figure 4-1 shows a general layout of the PCCS and its sub-systems, in the context of its interface with the containment. The sub-systems are described in the following sections.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 14 of 39
((
))
Figure 4-1: PCCS and Containment General Layout 4.1.1.1 Sub-Systems of PCCS 4.1.1.1.1 Primary Decay Heat Removal System The primary function of the PDHR system is to provide passive core cooling for non-LOCA accidents by removing core decay heat directly from the RCS and rejecting it to the Annular Reservoir (AR). During a LOCA, the PDHR system is ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )). It does this by direct closed-loop decay heat transfer to the AR from the primary RCS to a PDHR heat exchanger (PDHR HX2) in the AR via an intermediate PDHR heat exchanger (PDHR HX1) inside the containment. Both loops of the PDHR operate passively with via natural circulation.
All system components are safely located and protected within the CES.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 15 of 39
((
))
4.1.1.1.2 Secondary Decay Heat Removal System The primary function of the SDHR system is to provide passive core cooling for non-LOCA accidents by removing core decay heat indirectly from the RCS via the SG shell side coolant and rejecting it to the AR via the SDHR heat exchanger located in the reservoir. During a LOCA, the SDHR system is (( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )). The SDHR system operates by condensing SG shell side steam and returning the condensate to the feedwater of the SG. All system components are securely located within the CES.
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
4.1.1.1.3 Automatic Depressurization System The Automatic Depressurization System (ADS) works in conjunction with the Passive Core Makeup Water System (PCMWS), discussed in paragraph 4.1.1.1.4. The ADS is designed to provide the following functions:
Depressurize the RCS to allow for safety injection of water from the accumulators (ADS Stage 1)
Fully depressurize the RCS (to equalize with containment pressure) to allow gravity injection of water from the passive core makeup water tanks (PCMWT) (ADS Stage 2)
The ADS has two stages of depressurization with two trains in each stage. Only one train of each stage is needed to perform the depressurization function. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
Each ADS Stage 2 train consists of ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )). Each train has two valves in series so that the valves can be independently tested.
The ADS reduces RCS pressure in a staged fashion to allow passive injection of makeup water to maintain the RCS inventory above the top of the core. The Stage 1 ADS valves reduce RCS
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 16 of 39 pressure to allow the accumulators to provide makeup water to the core. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
The Stage 2 ADS valves reduce the RCS pressure to equalize with the containment pressure.
This permits gravity injection from the PCMWTs which provide additional makeup water to the core. By venting directly to the containment, the second stage ADS valves provide a decay heat removal path, via the passive containment heat removal system (PCHR) for long-term cooling.
((xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx))
4.1.1.1.4 Passive Core Makeup Water System The PCMWS provides emergency make up water inventory to the RCS to ensure sufficient water is in the core to mitigate the LOCA. The PCMWS is designed to provide make-up water for the entire spectrum of potential pipe break sizes and locations discussed in Section 3.5. The PCMWS provides sufficient water to sustain long-term cooling in conjunction with PCHR following a LOCA.
The PCMWS includes three modes of passive injection:
- 1. Medium pressure injection: Accumulators, pressurized with nitrogen, inject coolant into the RPV at medium pressure.
- 2. Low-pressure injection: PCMWTs gravity inject coolant into the RPV when the RCS pressure equalizes with containment pressure.
- 3. Long-term cooling: PCMWT drains to the spent fuel pool and the SFP also collects coolant from the break and the condensation on the containment walls. This gravity injects coolant into the RPV to sustain long-term cooling.
The ADS works in conjunction with the PCMWS to ensure that the RCS is depressurized sufficiently to facilitate passive injection to maintain RCS water inventory in the core.
The PCMWS has two trains for redundancy. Water enters the RPV through the direct vessel injection (DVI) line. Each train of the PCMWS has a dedicated pressurized accumulator and a large atmospheric make-up tank, (( )). There is sufficient borated water in one accumulator and the PCMWTs to ensure both the core and the spent fuel in the spent fuel pool remain subcritical and covered during a LOCA for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without operator action.
The PCMWTs and accumulators are connected to the RPV through the DVI lines. Strainers located in the PCMWTs minimize the possibility of debris clogging the DVI lines. During a postulated LOCA, the accumulators and PCMWTs provide borated makeup water to the core via the DVI lines to the RPV as the RCS pressure reduces. The water level in the PCMWTs creates sufficient head pressure to overcome pipe frictional and strainer pressure losses and gravity injects water into the core. ((
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 17 of 39 xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxx)). This allows for borated water to be available for cooling both the core and spent fuel pool and for keeping the core subcritical, assuming one control rod of the highest reactivity worth does not insert, and the SFP subcritical.
The water that condenses on the inner surface of the Containment Structure (CS) because of a LOCA returns to the PCMWT (( )) to support long term cooling. The operation of this system is passive and driven by the pressure difference between the sources of makeup water and the RCS.
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx))
4.1.2 Passive Containment Heat Removal System The PCHR system passively maintains the containment atmospheric pressure and temperature within design limits in the event of a LOCA by utilizing the steel CS and the water inventory in the AR to dissipate heat. It does not require any signals or actuations to perform its functions.
During a LOCA, steam is released into the containment atmosphere. The steam condenses on the inside surface of the CS wall and the heat is transferred through the CS wall to the water in the AR. The large heat transfer area and high conductance of the steel CS results in rapid heat rejection from the containment atmosphere to the water in the AR. As the water in the AR is heated, it rejects heat to the environment through evaporative cooling via the vent at the top of the CES. Water level is maintained in the AR during normal operation to assure PCHR is available after a postulated accident without any operator action.
(( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
4.2 Safety Requirements for the PCCS and PCHR The PCCS is designed to the following requirements:
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx Xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xx xxxxxxxxxxxxxxxxxxxxx ))
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 18 of 39
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx Xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xx xxxxx ))
The PCHR system is designed to the following requirements:
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxx ))
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 19 of 39 4.3 Operational Requirements for the PCCS and PCHR During a LOCA The following operational requirements apply during a LOCA to the PDHR and SDHR:
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
The following operational requirements apply during a LOCA to the ADS and PCMWS:
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
The following operational requirements apply during a LOCA to the PCHR:
((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
5.0 TECHNICAL EVALUATION
OF LOCA ACCEPTANCE CRITERIA The objective of the LOCA analyses is to assess the adequacy of structures, systems, and components provided for the mitigation of the consequences of a LOCA, to assess the risk of operation of the plant on the health and safety of the public.
Plant features such as plant safety system (PSS) and engineered safety features (ESFs) are designed to detect and mitigate the consequences of postulated transients and accidents.
These features help to prevent or limit the release of radioactive material to protect the health and safety of the public. The integrity of the three radiological barriers (the fuel cladding, the RCPB, and the containment structure) prevent or limit the release of radioactivity. These barriers act in series to prevent or limit the release of radioactive material outside the plant during postulated events. The transient and accident analysis acceptance criteria are linked to the classification of the postulated event so that for the higher frequency events, the integrity of all three barriers is maintained.
A LOCA is a design basis accident (DBA) where a pipe break in the RCPB is postulated. For the SMR-160, LOCA acceptance criteria have been established to demonstrate continued fuel cladding integrity to maintain this important physical barrier to release. These acceptance criteria are based around and intended to bound the acceptance criteria for emergency core cooling systems (ECCS) in 10 CFR 50.46. The current NRC regulations are applicable to SMR-160 LOCA analysis because the materials used in the design and the expected steady-
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 20 of 39 state and transient conditions are within the experimental database that were used in establishing the regulations.
5.1 Emergency Core Cooling System Acceptance Criteria (10 CFR 50.46)
The acceptance criteria for Emergency Core Cooling Systems during a LOCA are specified in 10 CFR 50.46. These are the requirements related to emergency core cooling system (ECCS) equipment that refills the RPV in a timely manner following a LOCA. The 10 CFR 50.46 acceptance criteria are as follows:
The calculated maximum fuel element cladding temperature shall not exceed 2200 °F.
The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
Calculated changes in core geometry shall be such that the core remains amenable to cooling.
After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
5.2 SMR-160 LOCA Acceptance Criteria Acceptance criteria specific to a LOCA are established to ensure that the ECCS acceptance criteria in 10 CFR 50.46 can be met. These criteria are:
the collapsed liquid level in the reactor pressure vessel (RPV) shall be maintained at or above the top of the active fuel, or fuel cladding temperature shall be maintained within the normal operating temperature range.
The collapsed liquid level is a conservative water level. It does not account for two-phases of coolant which would be present in the core; therefore, the actual water level in the core would be higher than the collapsed liquid level. If the collapsed liquid level in the RPV falls below the top of the active fuel, then it must be demonstrated that the fuel cladding temperature stays within the normal operating temperature range. Meeting one of these criteria will ensure the acceptance criteria in 10 CFR 50.46 are met; fuel cladding temperature will be below 2200° F, there will be no significant cladding oxidization or hydrogen generation, the core geometry will be maintained, and long-term cooling will be achieved.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 21 of 39
6.0 DESCRIPTION
OF A DESIGN BASIS LOCA 6.1 Event Initiation The initiation of the LOCA starts with a break of any piping that is part of the RCPB. A LOCA results in a loss of coolant from the RCS resulting in an immediate reduction in RCS inventory and pressure. Potential break locations are discussed in Section 3.5. Some break locations will result in liquid coolant flowing out of the break and collecting in lower containment. Other break locations will discharge steam into the containment atmosphere where it condenses on the CS surfaces and drains back into the lower containment. Each of the LOCA breaks will result in a similar sequence of events but they will vary in duration based on the break size and location.
6.2 Event Detection and Mitigation The reduction in RCS pressure and inventory causes a reactor trip signal to be generated which results in a reactor scram. This is followed shortly after by the generation of a Safeguards Actuation Signal (S-Signal) which ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
When the RCS further depressurizes as a result of the break and the level falls, a low RCS inventory setpoint is reached. This setpoint initiates the Automatic Depressurization System (ADS) Stage 1 actuation logic. Each stage of ADS has two redundant valves that open to further reduce pressure in the RCS. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx)) as described in paragraph 4.1.1.1.3.
After actuation of the ADS Stage 1 valves, the RCS pressure continues to decrease until the pressure falls below the pressure in the PCMWS accumulators. There are two redundant accumulators at the same pressure. They inject into the RPV downcomer via the DVI lines. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
The pressure and level in the RCS continue to decrease until a second, lower RCS inventory setpoint is reached. This initiates the ADS Stage 2 actuation logic. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
6.3 Long-Term Cooling Once ADS Stage 2 has actuated and the RCS pressure equalizes with the containment pressure, the two PCMWTs contain large inventories of water stored inside the CS. They are also connected to the DVI lines and passively inject into the RPV downcomer. ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx)) This inventory sharing allows water to passively recirculate between the containment and the RPV for long-term cooling.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 22 of 39 Figure 6-1 illustrates a typical LOCA by providing a schematic of the sequence of events. The sequence and order of events are similar for most break sizes and locations with the primary difference being the time duration between events.
((
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Figure 6-1: Sequence of events for a typical LOCA
7.0 REGULATORY EVALUATION
AS PERTAINS TO ELIMINATION OF LARGE BREAK LOCA AND ESTABLISHMENT OF LOCA ACCEPTANCE CRITERIA 7.1 Applicable 10 CFR 50 Regulations and General Design Criteria 7.1.1 10 CFR 50.34(f) 10 CFR 50.34(f), Additional TMI-related requirements [2], requires that each applicant for a design certification, design approval, combined license, or manufacturing license under Part 52 of this chapter shall demonstrate compliance with the technically relevant portions of the requirements in paragraphs (f)(1) through (3) of this section, except for paragraphs (f)(1)(xii),
(f)(2)(ix), and (f)(3)(v). The following requirements are evaluated as they are related to the LOCA and LOCA acceptance criteria
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 23 of 39 Regulatory Requirement: 10 CFR 50.34(f)(1)(iii) requires an evaluation of the potential for and impact of reactor coolant pump seal damage following small-break LOCA with loss of offsite power. If damage cannot be precluded, provide an analysis of the limiting small-break loss-of-coolant accident with subsequent reactor coolant pump seal damage.
Statement of Compliance: The SMR-160 does not include reactor coolant pumps for normal operation, as this plant relies on natural circulation. Therefore, this requirement is not technically relevant to the SMR-160.
Regulatory Requirement: 10 CFR 50.34(f)(1)(iv) requires an evaluation of the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV). If this probability is a significant contributor to the probability of small-break LOCA's from all causes, provide a description and evaluation of the effect on small-break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened.
Statement of Compliance: The SMR-160 design does not include any PORVs on the pressurizer. The pressurizer is sufficiently sized such that the plant can accommodate normal power maneuvers without needing a PORV. The ADS Stage 1 valves are connected to the pressurizer, however each of the two ADS Stage 1 trains have two valves in series, so that a single failure of one of the valves would not result in inadvertent depressurization or a failure to isolate the pressurizer in the event that ADS is terminated. Therefore, this requirement is not technically relevant to the SMR-160.
Regulatory Requirement: 10 CFR50.34(f)(2)(vi) requires the provision of the capability of high point venting of non-condensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity.
Statement of Compliance: For the SMR-160 the capability for remotely operated high point venting of the reactor coolant system is provided by ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxx )). During loss of cooling accident events, the ADS automatically depressurizes the RCS so that the PCCS may effectively deliver core cooling flow. Depressurization via the ADS results in creation of a gas-steam volume in the upper region of the vessel. This vapor volume ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxx )). This process provides an open injection and steam venting flow path through the reactor vessel, maintaining required core cooling flow. Each train of ADS has two valves in series that are powered by different buses thus minimizing the potential for inadvertent actuation.
Regulatory Requirement: 10 CFR50.34(f)(2)(x) requires the provision of a test program and associated model development and to conduct tests to qualify reactor coolant system relief and
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 24 of 39 safety valves and, for PWR's, PORV block valves, for all fluid conditions expected under operating conditions, transients, and accidents. Consideration of anticipated transients without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed Statement of Compliance: The SMR-160 reactor coolant system design does not include PORVs and their associated block valves. However, the safety valve used in the SMR-160 design will be tested in accordance with the guidelines of Item [II.D.1] of NUREG-0737 [3].
Accident analyses will be performed for the SMR-160 to determine fluid conditions expected under operating conditions, transients, and accidents, and the postulated system responses to these conditions, including the operation of pressurizer safety valves. Anticipated transients without scram events will also be analyzed. Appropriate valve qualification documentation will be obtained and maintained in future licensing activities.
Regulatory Requirement: 10 CFR50.34(f)(2)(xi) Provide direct indication of relief and safety valve position (open or closed) in the control room.
Statement of Compliance: The SMR-160 design does not include PORVs and their associated block valves from the RCS. (( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxx ))
Based on the above discussions, the SMR-160 either will comply with these requirements or they are not technically relevant to the SMR-160.
7.1.2 10 CFR 50.46 10 CFR 50.46, Acceptance criteria for Emergency Core Cooling Systems (ECCS) for light-water nuclear power reactors [4], includes the following requirements:
Regulatory Requirement: 10 CFR 50.46 (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section.
Statement of Compliance: The SMR-160 is equipped with a passive core cooling system (PCCS) that serves the purposes of an ECCS as defined by the NRC. The systems of the PCCS that mitigate a LOCA are the ADS and PCMWS. These are described further in Section 4.1.1. A set of SMR-160 acceptance criteria for a LOCA have been developed that bound the acceptance criteria in 10 CFR 50.46(b). The PCCS has been designed such that each of these bounding criteria will be met for all possible break sizes and locations even under limiting conditions and crediting a limiting single failure. Therefore, this requirement is met through PCCS design and the implementation of SMR-160 specific acceptance criteria.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 25 of 39 Regulatory Requirement: 10 CFR 50.46 (a)(1)(i) ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model.
10 CFR 50.46 (a)(1)(ii) Alternatively, an ECCS evaluation model may be developed in conformance with the required and acceptable features of Appendix K ECCS Evaluation Models.
Statement of Compliance: ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx
)) The evaluation model will then be used to perform a deterministic safety analysis of a number of postulated loss-of-coolant accidents of different sizes and locations, as described in Section 3.5 to provide assurance that the most severe postulated loss-of-coolant accidents are calculated using conservative methods. The ECCS model will conform with the required and acceptable features of Appendix K.
Regulatory Requirement: 10 CFR50.46(b)(1), Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200° F.
Statement of Compliance: The SMR-160 acceptance criteria require that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within normal operating temperature range. Meeting one of these criteria will ensure that the calculated maximum fuel element cladding temperature does not exceed the 2200 °F. The LOCA safety analysis for a licensing application will demonstrate compliance with this requirement.
Regulatory Requirement: 10 CFR50.46 (b)(2), Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. As used in this subparagraph total oxidation means the total thickness of cladding metal that would be locally converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the oxidation, beginning at the calculated time of rupture. Cladding thickness before oxidation means the radial distance from inside to outside the cladding, after any calculated rupture or swelling has
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 26 of 39 occurred but before significant oxidation. Where the calculated conditions of transient pressure and temperature lead to a prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding thickness shall be defined as the cladding cross-sectional area, taken at a horizontal plane at the elevation of the rupture, if it occurs, or at the elevation of the highest cladding temperature if no rupture is calculated to occur, divided by the average circumference at that elevation. For ruptured cladding the circumference does not include the rupture opening.
Statement of Compliance: The SMR-160 acceptance criteria require that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within normal operating temperature range. Meeting one of these criteria will ensure that the total oxidation of the cladding will not exceed 0.17 times the total cladding thickness before oxidation. The LOCA safety analysis for a licensing application will demonstrate compliance with this requirement.
Regulatory Requirement: 10 CFR50.46 (b)(3), Maximum Hydrogen Generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
Statement of Compliance: The SMR-160 acceptance criteria require that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within normal operating temperature range. Meeting one of these criteria will ensure that the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam does not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. The LOCA safety analysis for a licensing application will demonstrate compliance with this requirement.
Regulatory Requirement: 10 CFR50.46 (b)(4), Coolable Geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
Statement of Compliance: The SMR-160 acceptance criteria require that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within normal operating temperature range. Meeting one of these criteria will ensure that no significant changes in core geometry occur, such that the core remains amenable to cooling. The LOCA safety analysis for a licensing application will demonstrate compliance with this requirement.
Regulatory Requirement: 10 CFR50.46 (b)(5), Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 27 of 39 Statement of Compliance: The SMR-160 acceptance criteria require that the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within normal operating temperature range. Meeting this criterion will ensure that the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. The SMR-160 PCCS design consists of passive systems without the need for offsite power or operator action. The only electrical power requirements are supplied by Class 1E dc batteries. The steam generated in the RPV is released into the containment through the ADS valves, where it is condensed on the surface of the CS. This latent and sensible heat in the containment is dissipated by the AR. The condensed steam in the containment structure is returned to the PCMWT ((
xxxxxxxxxxxxxxxx )), thus assuring supply of water to maintain the level in the core for long term cooling. The LOCA safety analysis for a licensing application will demonstrate compliance with this requirement.
7.1.3 10 CFR 50.55a Regulatory Requirement: 10 CFR 50.55a, Codes and Standards. The regulation lists the ASME Codes that are approved for incorporation by reference, and the use and conditions for the standards. This rule establishes minimum quality standards for the design, fabrication, erection, construction, testing, and inspection of the vessel and certain components of nuclear power plants by requiring conformance with appropriate editions of specified published industry codes and standards.
Statement of Compliance: The design, fabrication, erection, construction, testing, and inspection of the SMR-160 Combined Vessel, PCCS, and PCHR will use the standards approved in 10 CFR 50.55a in effect within six months of any license application, including any application for a construction permit under 10 CFR 50 or a design certification application under 10 CFR 52. Further details are to be described during future licensing activities.
7.1.4 10 CFR 50 Appendix A, GDC 1 Regulatory Requirement: 10 CFR 50 Appendix A GDC 1, Quality standards and records [5].
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
Statement of Compliance: The SMR-160 RCPB, PCCS, and PCHR are to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 28 of 39 safety functions to be performed in accordance with generally recognized codes and standards, and under an approved quality assurance program with approved control of records. Therefore, the SMR-160 design will meet the requirements of the 10 CFR 50 Appendix A GDC 1.
7.1.5 10 CFR 50 Appendix A, GDC 2 Regulatory Requirement: 10 CFR 50 Appendix A GDC 2, Design bases for protection against natural phenomena [5]: Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1)
Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Statement of Compliance: The SMR-160 RCPB, PCCS, and PCHR SSCs important to safety are to be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. Specific design requirements for the RCPB, PCCS and PCHR to verify the capability to perform their safety functions, and the natural phenomena and effects evaluated, will be provided during future licensing activities. Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 2.
7.1.6 10 CFR 50 Appendix A, GDC 4 Regulatory Requirement: 10 CFR 50 Appendix A GDC 4, Environmental and dynamic effects design bases [5]: Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
Statement of Compliance: Safety-related structures, systems, and components of the SMR-160 will be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 29 of 39 The SMR-160 structures, systems, and components will be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. Details of the design, environmental testing, and construction of these structures, systems, and components will be provided in future licensing activities.
7.1.7 10 CFR 50 Appendix A, GDC 14 Regulatory Requirement: 10 CFR 50 Appendix A GDC 14, Reactor coolant pressure boundary: 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.
Statement of Compliance: As discussed in Section 3.0, the Combined Vessel is designed to meet the requirements of ASME Boiler and Pressure Vessel Code Section III, Division 1, Subsection NB for vessel design. This results in an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture, in compliance with this criterion.
Reactor coolant piping is designed in accordance with ASME Boiler and Pressure Vessel Code Section III [1], Division 1, Subsection NB for Class 1 Components. The reactor coolant piping maintains the reactor coolant pressure boundary and is designed to withstand the maximum RCS pressure and temperature under all expected modes of plant operation, anticipated operational occurrences, and design basis accidents. Reactor coolant pressure boundary piping is all welded construction. Seamless piping is used to avoid in-service inspection of longitudinal piping welds. This results in an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture, in compliance with this criterion.
Further design details are to be described during future licensing activities. Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 14.
7.1.8 10 CFR 50 Appendix A, GDC 15 Regulatory Requirement: 10 CFR 50 Appendix A GDC 15, Reactor coolant system design:
The reactor coolant system and associated auxiliary, control, and protection systems shall 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.
Statement of Compliance: The combination of Plant Safety System (PSS), PDHR, and SDHR assure the design conditions of the RCPB are not exceeded for any normal operating conditions. In addition, the Pressurizer Safety Valves are sized such that they prevent the RCPB from exceeding 110% of the design pressure in accordance with ASME B&PV Code,Section III, Article NB-7000 for the most limiting anticipated operational occurrence. Further design details will be described during future licensing activities.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 30 of 39 Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 15.
7.1.9 10 CFR 50 Appendix A, GDC 30 Regulatory Requirement: 10 CFR 50 Appendix A GDC 30, Quality of reactor coolant pressure boundary: Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.
Statement of Compliance: The components of the RCPB are to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed in accordance with generally recognized codes and standards, and under an approved quality assurance program with approved control of records, as required by 10 CFR 50.55a and 10 CFR 50 Appendix A, GDC 1.
The SMR-160 design will incorporate a means to detect reactor coolant leakage for components of the RCPB, and to the extent practicable identify the location of the source of reactor coolant leakage. Further design details will be described during future licensing activities.
Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 30.
7.1.10 10 CFR 50 Appendix A, GDC 31 Regulatory Requirement: 10 CFR 50 Appendix A GDC 31,Fracture prevention of reactor coolant pressure boundary: The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.
Statement of Compliance: The components of the RCPB are to be designed with sufficient margin to assure that these requirements are met, with further design details to be described during future licensing activities.
Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 31.
7.1.11 10 CFR 50 Appendix A, GDC 33 Regulatory Requirement: 10 CFR 50 Appendix A GDC 33, Reactor coolant makeup: A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 31 of 39 specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.
Statement of Compliance: The design of the PZR accommodates expansion and contraction of the reactor coolant volume during controlled power maneuvers (( ))
percent without the need to makeup or letdown of coolant. The pressurizer water level control function of the PCS is designed to provide automatic water level control from startup to nominal full power operating conditions, using the chemical and volume control system (CVCS) to makeup water if the level falls below the low level of the dead band. In addition, the pressurizer has sufficient volume to accommodate minor reactor coolant system leakage.
The PCMWS provides safety-related makeup to accommodate small leaks when the normal makeup system is unavailable and to accommodate larger leaks resulting from loss of coolant accidents. Safety-related reactor coolant makeup and safety injection are provided by two accumulators and two PCMWTs. The water that condenses on the inner surface of the Containment Structure (CS) because of a LOCA returns to the PCMWT (( )) to support long term cooling. The safety-related reactor coolant makeup relies on the Class 1E dc power. Neither onsite nor offsite ac power is required.
In addition, the non-safety-related CVCS automatically provides inventory control ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )).
Details of the CVCS design and RCS inventory control will be provided in future licensing activities.
7.1.12 10 CFR 50 Appendix A, GDC 34 Regulatory Requirement: 10 CFR 50 Appendix A GDC 34, Residual heat removal [5]: A system to remove residual heat shall 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 specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 32 of 39 Statement of Compliance: For non-LOCA events, the PDHR and SDHR systems provide a safety-related means to remove decay heat and residual heat from the reactor core, ultimately transferring the heat to the annual reservoir. (( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )), thus they do not rely on offsite or onsite power and the systems; safety function can be accomplished with an assumed single failure.
For LOCA events, the ADS in conjunction with the PCMWS provides the means to remove the residual and decay heat from the core. These systems meet the definition of an ECCS as described in 10 CFR 50.46(a)(1)(i) whose cooling performance following a postulated LOCA will be in compliance with the SMR-160 acceptance criteria in response to a LOCA, which bound the acceptance criteria set forth in 10 CFR 50.46(b). Details will be described in future licensing activities.
7.1.13 10 CFR 50 Appendix A, GDC 35 Regulatory Requirement: 10 CFR 50 Appendix A GDC 35, Emergency core cooling [5]: A system to provide abundant emergency core cooling shall 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. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Statement of Compliance: As previously described, the combined design features of the PCCS meet the definition of an ECCS as described in 10 CFR 50.46(a)(1)(i) that has a calculated cooling performance following postulated LOCAs in compliance with the SMR-160 acceptance criteria in response to a LOCA which bound the acceptance criteria set forth in 10 CFR 50.46(b). In addition, the combination of the ADS and PCMWS are effective as an ECCS for breaks in pipes in the RCPB up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the RCS in compliance with the definition of a LOCA in 10 CFR 50.46(c)(1). The SMR-160 acceptance criteria in response to a LOCA is that reactor water level is maintained at or above TAF or fuel cladding temperature is maintained within normal operating temperature range, such that the performance of the ADS and PCMWS is sufficient to ensure 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. These passive systems (( xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )); safety function can be accomplished with an assumed single failure.
The analyses to demonstrate compliance with the SMR-160 acceptance criteria in response to a LOCA (i.e., reactor water level is maintained at or above TAF or fuel cladding temperature is maintained within normal operating temperature range) will be provided during future licensing activities.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 33 of 39 Therefore, the SMR-160 design will meet the requirements of 10 CFR 50 Appendix A, GDC 35.
7.1.14 10 CFR 50 Appendix A, GDC 38 Regulatory Requirement: 10 CFR 50 Appendix A GDC 38, Containment Heat Removal [5]: A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss of coolant accident and maintain them at acceptably low levels.
Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
Statement of Compliance: The SMR-160 design uses passive systems for post-LOCA core and containment heat removal and for the prevention of over pressurization failure of the containment structure (CS). As described in Section 4.1.2, the PCHR system passively maintains the containment atmospheric pressure and temperature within design limits in the event of a LOCA by utilizing the steel CS and the water inventory in the AR to dissipate heat. It does not require any signals or actuations to perform its functions.
The SMR-160 PCHR system is designed with sufficient capacity to prevent the containment from exceeding its design pressure with no operator action or outside assistance for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The PCHR system is a passive system that uses gravity and natural circulation as driving forces. The design of the SMR-160 PCHR system does not require the use of any pumps, and it functions independent of any power sources. Therefore, the PCHR system can function during loss of offsite or onsite power. The PCHR system will meet the requirements of GDC 38.
Details will be provided during future licensing activities.
7.2 Applicable Regulatory Guides 7.2.1 Regulatory Guide 1.26 Regulatory Guide (RG) 1.26, Quality Group Classifications and Standards for Water-, Steam-,
and Radioactive-Waste-Containing Components of Nuclear Power Plants [6], describes a quality classification system related to specified national standards that may be used to determine quality standards acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for satisfying GDC 1, Quality Standards and Records, as set forth in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10, Part 50, of the Code of Federal Regulations (10 CFR 50), Domestic Licensing of Production and Utilization Facilities, for components containing water, steam, or radioactive material in light-water-cooled nuclear power plants.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 34 of 39 The SMR-160 design will comply with the requirements of 10 CFR 50 Appendix A, GDC 1.
Therefore, the structures, systems, and components (SSCs) are classified in conformance with the guidance and regulatory positions provided in RG 1.26.
7.2.2 Regulatory Guide 1.29 RG 1.29, Seismic Design Classification for Nuclear Power Plants [7], describes a method that the staff of the NRC considers acceptable for use in identifying and classifying those features of light-water-reactor (LWR) nuclear power plants that must be designed to withstand the effects of the safe-shutdown earthquake (SSE).
The SMR-160 design will comply with the requirements of 10 CFR 50.48, 10 CFR 50.55a(h), 10 CFR 50 Appendix A, GDC 2, and 10 CFR 50 Appendix S. The RCPB, reactor core, reactor internals, safety systems, systems needed to shut down the reactor are classified as Seismic Category I. Therefore, the SMR-160 design is in conformance with the guidance and regulatory positions provided in RG 1.29.
7.2.3 Regulatory Guide 1.45 RG 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage [8],
describes methods that the staff of the NRC considers acceptable for use in implementing the regulatory requirements on GDC 14, GDC30, and 10 CFR 50.55a. These regulations deal with ensuring an extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture of the RCPB. They also require plants to provide the means for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.
Additionally, inservice inspection and testing is required to provide defense in depth for assuring structural integrity of the RCPB is maintained.
Sections 7.1.4 and 7.1.7 describe how the design of the Combined Vessel, PCCS and PCHR will comply with the requirements of 10 CFR 50 Appendix A, GDC 1 and GDC 14, respectively.
The SMR-160 design will incorporate a means to detect reactor coolant leakage for components of the RCPB, and to the extent practicable identify the location of the source of reactor coolant leakage, thus will be compliant with 10 CFR 50 Appendix A. The ISI and IST of the RCS, PCCS and PCHR components will meet the requirements of 10 CFR 50.55a, to be demonstrated in future licensing activities. Therefore, the SMR-160 design conforms to the guidance, including regulatory positions of RG 1.45.
7.2.4 Regulatory Guide 1.84 RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III [9],
lists the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code,Section III Code Cases that the NRC has approved for use as voluntary alternatives to the mandatory ASME B&PV Code provisions that are incorporated by reference into 10 CFR 50.
These regulations implement the requirements of GDC 1 for quality stands and records, GDC 30 for quality of RCPB, and 10 CFR 50.55a(c). This requires that components of the RCPB
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 35 of 39 must be design, fabricated, erected, and tested in accordance with the requirements for the Class 1 components of ASME B&PV Code,Section III, or equivalent quality standards.
Sections 7.1.4 and 7.1.9 describe how the design of the Combined Vessel, will comply with the requirements of 10 CFR 50 Appendix A, GDC 1, and 10 CFR 50 Appendix A, GDC 30, respectively. Section 7.1.3 describes how the design of the Combined Vessel will comply with the requirements of 10 CFR 50.55a. Compliance with the requirements of 10 CFR 50.55a, including the use of ASME B&PV Section III Code Cases endorsed in RG 1.84 where necessary, is to be demonstrated during future licensing activities. Therefore, the SMR-160 design will conform to the guidance and regulatory positions of RG 1.84.
7.2.5 Regulatory Guide 1.147 RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 [10],
lists the ASME B&PV Code,Section XI Code Cases that the NRC has approved for use as voluntary alternatives to the mandatory ASME B&PV Code provisions that are incorporated by reference into 10 CFR 50.
The requirements of ASME B&PV Code,Section XI, Division 1, apply during construction and operation activities of nuclear power plants for performance of inservice inspection activities and do not apply during the design phase. Therefore, these regulatory requirements are not technically relevant to the SMR-160 in this phase of the design. Compliance with the requirements of inservice inspection discussed in RG 1.147 will be demonstrated in future licensing activities.
7.2.6 Regulatory Guide 1.192 RG 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code [11], lists Code Cases associated with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) that the NRC has approved for use as voluntary alternatives to the mandatory ASME OM Code provisions that are incorporated by reference into 10 CFR 50.
The requirements of ASME OM Code apply during operation and maintenance activities of nuclear power plants for performance of inservice testing activities and do not apply during the design phase. Therefore, these regulatory requirements are not technically relevant to the SMR-160 in this phase of the design. Compliance with the requirements for the conduct of inservice testing discussed in RG 1.192 will be demonstrated during future licensing activities.
7.3 Applicable NUREG-0800 Sections 7.3.1 Standard Review Plan 6.3 SRP 6.3, Emergency Core Cooling System [12], states the following areas of review are as follows:
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 36 of 39
- 1. The design bases for the ECCS are reviewed to assure that they satisfy applicable regulations, including the general design criteria and the requirements of 10 CFR 50.46 regarding ECCS acceptance criteria.
- 2. The design basis for the automatic depressurization systems (ADS) are also reviewed for compliance with TMI Action Plan Items and associated guidance. This applies to BWRs and the advanced passive reactors (both PWRs and BWRs).
- 3. For advanced passive reactors which rely on gravitational head to provide ECCS injection to the RCS, the RCS must be designed with an ADS such that the available gravitational head is sufficient to provide adequate core cooling when depressurized.
- 4. For advanced reactors which rely on passive safety-related systems and equipment to automatically establish and maintain safe-shutdown conditions for the plant, these passive safety systems must be designed with sufficient capability to maintain safe shutdown conditions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, without operator actions and without non-safety-related onsite or offsite power.
- 5. The design of the ECCS is reviewed to determine that it is capable of performing all of the functions required by the design bases.
- 6. The preoperational and initial startup test programs for the ECCS are reviewed to determine if they are sufficient to confirm the performance capability of the ECCS. The need for special design features to permit the performance of adequate test programs should also be reviewed.
- 7. The proposed technical specifications (TSs) are reviewed to assure that they are adequate in regard to limiting conditions of operation and periodic surveillance testing.
- 8. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC). For design certification (DC) and combined license (COL) reviews, the staff reviews the applicant's proposed ITAAC associated with the structures, systems, and components (SSCs) related to this SRP section in accordance with SRP Section 14.3, "Inspections, Tests, Analyses, and Acceptance Criteria." The staff recognizes that the review of ITAAC cannot be completed until after the rest of this portion of the application has been reviewed against acceptance criteria contained in this SRP section. Furthermore, the staff reviews the ITAAC to ensure that all SSCs in this area of review are identified and addressed as appropriate in accordance with SRP Section 14.3.
- 9. COL Action Items and Certification Requirements and Restrictions. For a DC application, the review will also address COL action items and requirements and restrictions (e.g.,
interface requirements and site parameters).
For a COL application referencing a DC, a COL applicant must address COL action items (referred to as COL license information in certain DCs) included in the referenced DC. Additionally, a COL applicant must address requirements and restrictions (e.g.,
interface requirements and site parameters) included in the referenced DC.
The first five areas of review deal with the design concepts of the ECCS and are applicable to the SMR-160 in the current design phase. The last four areas of review deal with testing requirements and will be demonstrated during future licensing activities. For the first five review areas, the SMR-160 design conforms with the guidance of SRP 6.3 based on the following:
- 1. The SMR-160 ECCS, or the PCCS, should maintain the collapsed liquid level in the RPV at or above the top of the active fuel or maintain fuel cladding temperature within normal operating temperature range. These acceptance criteria were developed to be more restrictive such that the requirements in both NUREG-0800 and 10 CFR 50.46 can be met.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 37 of 39
- 2. The SMR-160 design includes two stages of ADS to provide fast depressurization following a LOCA and sufficient venting for long-term cooling. The ADS consists of ((
xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx )). The ADS has been designed such that compliance with TMI Action Plan Items are addressed.
- 3. The first stage of ADS valves are sized to reduce RCS pressure sufficiently to allow the accumulators to provide makeup water to the core. The second stage of ADS valves are sized to reduce the RCS pressure sufficiently to equalize with the containment pressure and allow gravity injection from the PCMWTs to provide makeup water to the core.
- 4. The PDHR and SDHR are passive safety systems that can maintain the SMR-160 in a safe shutdown condition by removing sensible and decay heat. The PCMWTs hold a sufficient inventory of makeup water to provide cooling for the core and the spent fuel pool in the event of a LOCA. There is enough water in the PCMWTs to keep the reactor core shutdown and the spent fuel covered for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. These safety systems will operate without the need for operator action or any non-safety-related onsite or offsite power.
- 5. Based on the above functional design requirements, the PCCS can perform all of the functions required by the design bases as outlined in the SRP.
7.3.2 Standard Review Plan 15.6.5 SRP 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary [12], addresses the requirements associated with the safety analysis of a design basis LOCA. This includes the requirements of 10 CFR 50 Appendix A, GDC 35, 10 CFR 50.46, 10 CFR 50 Appendix K, and SRP 6.3.
SRP 15.6.5 states, The piping breaks are postulated to occur at various locations and include a spectrum of break sizes, up to a maximum pipe break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant pressure boundary. As discussed in Section 3.5, the largest RCPB piping in the SMR-160 is in the connected systems, and not in the RCS.
The SMR-160 PCCS should maintain the collapsed liquid level in the RPV at or above the top of the active fuel or maintain fuel cladding temperature within normal operating temperature range.
These acceptance criteria were developed to be more restrictive such that the requirements in SRP 15.6.5 can be met.
8.0
SUMMARY
AND CONCLUSIONS The RPV and the reactor coolant side of the SG with the integral PZR are designed and analyzed as one pressure vessel with no connecting piping, meeting the requirements of the ASME Boiler and Pressure Vessel Code Section III, Division 1, Subsection NB for vessel design, as listed in Section 3.2. Since no piping exists between the RPV and the SG, postulating a break in the Combined Vessel at this location is non-credible when the vessel is designed to the requirements described in Section 3.2. Likewise, the SG riser forms an integral part of the SG, hence, postulating a break in the Combined Vessel at this location is non-credible when it is designed to the requirements described in Section 3.2. As a result, the piping to be considered for postulated breaks is (( )) in diameter, therefore there are no potential large breaks in the SMR-160 design.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 38 of 39 The established LOCA acceptance criteria for the SMR-160 are the collapsed liquid level in the RPV shall be maintained at or above the top of the active fuel or fuel cladding temperature shall be maintained within the normal operating temperature. This will ensure the ECCS acceptance criteria in 10 CFR 50.46 can be met. Meeting either of these criteria will result in meeting the acceptance criteria in 10 CFR 50.46 as no significant fuel cladding heat up, oxidization, or hydrogen generation will occur. It would also ensure that there is no significant change in core geometry and long-term cooling can be achieved.
The SMR-160 RCS design and LOCA acceptance criteria will be in compliance with applicable regulations, GDCs, and regulatory guidance.
Elimination of the Large Break Loss of Coolant Accident (LOCA) and Establishment of LOCA Acceptance Criteria for SMR-160 Report No. HI-2201064-R1-NP Copyright Holtec International © 2020, all rights reserved Page 39 of 39
9.0 REFERENCES
[1] ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Facility Components," 2017 Edition.
[2] U.S. Code of Federal Regulations, 10 CFR 50.34(f), "Additional TMI-related requirements".
[3] NUREG-0737, "Clarification of TMI Action Plan Requirements (NUREG-0737, Final Report)," 1980.
[4] U.S. Code of Federal Regulations, 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors".
[5] U.S. Code of Federal Regulations, 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants".
[6] U.S. NRC Regulatory Guide, RG 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," Revision 5.
[7] U.S. NRC Regulatory Guide, RG 1.29, "Seismic Design Classification for Nuclear Power Plants," Revision 5.
[8] U.S. NRC Regulatory Guide, RG 1.45, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," Revsion 1.
[9] US NRC Regulatory Guide, RG 1.84, "Design, Fabrication, and Materials Code Case Acceptability, ASME Section III," Revisioin 38.
[10] U.S. NRC Regulatory Guide, RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 19.
[11] U.S. NRC Regulatory Guide, RG 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code," Revision 3.
[12] NUREG-0800, "Standard Review Plant for the Review of Safety Analysis Reports for Nuclear Power Plants," Revision 3.
[13] U.S. NRC Regulatory Guide, RG 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance," 1989.
[14] U.S. Code of Federal Regulations, 10 CFR 50, Appendix K, "ECCS Evaluation Models".