ML18190A519
| ML18190A519 | |
| Person / Time | |
|---|---|
| Site: | PROJ0769, NuScale |
| Issue date: | 07/09/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| Shared Package | |
| ML18190A518 | List: |
| References | |
| AF-0718-60782, RAIO-0718-60781 | |
| Download: ML18190A519 (26) | |
Text
RAIO-0718-60781 July 09, 2018 Docket: PROJ0769 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No.
9513 (eRAI No. 9513) on the NuScale Topical Report, "Non-Loss of Coolant Accident Analysis Methodology," TR-0516-49416, Revision 1
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 9513 (eRAI No. 9513)," dated May 08, 2018
- 2. NuScale Topical Report, "Non-Loss of Coolant Accident Analysis Methodology," TR-0516-49416, Revision 1, dated August 2017 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The Enclosures to this letter contain NuScale's response to the following RAI Questions from NRC eRAI No. 9513:
15.00.02-14 15.00.02-17 15.00.02-18 15.00.02-19 15.00.02-20 15.00.02-21 The response schedule for the remaining questions of RAI No. 9513, eRAI 9513 were provided in emails to the NRC (Greg Cranston) dated June 19 and June 26, 2018. is the proprietary version of the NuScale Response to NRC RAI No. 9513 (eRAI No. 9513). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The proprietary enclosures have been deemed to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.
This letter and the enclosed responses make no new regulatory commitments and no revisions to any existing regulatory commitments.
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
RAIO-0718-60781 If you have any questions on this response, please contact Paul Infanger at 541-452-7351 or at pinfanger@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Rani Franovich, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9513, proprietary : NuScale Response to NRC Request for Additional Information eRAI No. 9513, nonproprietary : Affidavit of Zackary W. Rad, AF-0718-60782 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Response to NRC Request for Additional Information eRAI No. 9513, proprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Response to NRC Request for Additional Information eRAI No. 9513, nonproprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
Response to Request for Additional Information Docket No.52-048 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-14 General Design Criterion (GDC) 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs). In addition, GDC 15, Reactor coolant system design, requires that the reactor coolant system (RCS) 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 (RCPB) are not exceeded during any condition of normal operation, including AOOs.
Topical report (TR) TR-0516-49416-P, Non-Loss-of-Coolant Accident [Non-LOCA] Analysis Methodology, supports the conclusions relative to GDC 10 and 15 in the NuScale Final Safety Analysis Report (FSAR). Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods, describes a process that the staff finds acceptable for developing and assessing evaluation models. Section 1.1.4 of RG 1.203 describes the identification and ranking of key phenomena and processes and states: An optimum analysis reduces candidate phenomena to a manageable set by identifying and ranking the phenomena with respect to their influence on figures of merit. Each phase of the transient scenario and system components are separately investigated. The outcome of this process is a phenomena identification and ranking table (PIRT), which should be used to determine the requirements for physical model development, scalability, validation, and sensitivity studies.
TR Sections 5.1.4.28, ((2(a),(c), and 5.1.4.32, (( 2(a),(c)
}} , state that these phenomena are highly ranked during
(( }}2(a),(c), which the staff notes is a figure of merit during Phase 3 (stable natural circulation) as well. To provide a consistent understanding of the presented PIRT high-ranked phenomena, explain why these phenomena are not highly ranked during Phase 3, and update TR-0516-49416-P as appropriate. NuScale Nonproprietary
NuScale Response: The effects of the ((
}}2(a),(c), were evaluated as high importance in the non-LOCA PIRT during Phases 1 and 2 (i.e., pre-trip transient and post-trip transition) of non-LOCA transients based on ((
}}2(a),(c) During Phase 3 (i.e., stable natural circulation) of non-LOCA transients, however, those phenomena were evaluated as ((
}}2(a),(c) during Phase 3 were deemed to be relatively less significant compared to those during (( }}2(a),(c) due to the significantly reduced core decay power level. These importance rankings, as documented in TR-0516-49416-P, were based upon preliminary non-LOCA PIRT development evaluations consistent with current DBE evaluations, which showed ((
}}2(a),(c) , due to the reduced primary flow rate in conjunction with a relatively high core decay power level. The ((
}}2(a),(c), due to the continuously decreasing core decay power level and the settled-down primary flow rate.
Since only highly ranked phenomena are discussed in the Non-LOCA topical report, no change was made to the topical report due to this response. Impact on DCA: There are no impacts to the DCA as a result of this response. NuScale Nonproprietary
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-17 GDC 10 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs. In addition, GDC 15 requires that the RCS and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs. TR-0516-49416-P supports the conclusions relative to GDC 10 and 15 in the NuScale FSAR. TR Section 7.2, Event Specific Methodology, provides representative results of sensitivity studies and states that such studies are performed to identify plant conditions that result in bounding transient analyses. It is the staffs understanding based on audit discussions with the applicant that these sensitivity studies are examples only and should be conducted for each new licensing basis calculation to establish bounding analyses. However, TR Section 1.2, Scope, is unclear about the purpose of these sensitivity studies. While the TR states that representative analysis results are provided to illustrate results from application of the EM, the TR also states that the scope of the evaluation model includes the general and event-specific analysis methodologies of the EM in Section 7.0 of the TR (which is where the representative sensitivity study results reside). Because the scope of the evaluation model should be clearly defined such that the staff is able to support its GDC 10 and 15 findings, and because the purpose of the representative sensitivity studies may be unclear to a user of the methodology, please clarify the intent of the representative sensitivity studies. Update TR-0516-49416-P to clearly reflect this intent. NuScale Response: NuScale considers the sensitivity studies presented in Section 7.2 of TR-0516-49416-P to be adequate to justify the biasing directions selected for each event. As such, these studies do not need to be repeated as they are representative of the NPM design. In contrast, the results presented in Section 8.0 of TR-0516-49416-P are examples only, and will be performed for each new application. The examples in Section 8.0 are presented for illustration only using NuScale Nonproprietary
representative initial conditions and biasing. Section 1.2 of TR-0516-49416-P was revised as presented at the end of this response to clarify the intent of the sensitivity studies. Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 1.0 Introduction 1.1 Purpose The purpose of this report is to present the NuScale evaluation model (EM) used to evaluate the NuScale Power Module (NPM) system transient response to non-loss-of-coolant accident (non-LOCA) events with the NRELAP5 code. This report summarizes the NuScale plant design and identifies the potential non-LOCA initiating events for the NPM analyzed by this EM. The classification of these non-LOCA events and relevant acceptance criteria that are prescribed in the NRC standard review plan (SRP) and the NuScale design specific review standard (DSRS) are discussed in this report. The purpose of the non-LOCA evaluation model is to model the NPM response to a non-LOCA design basis event. The non-LOCA system transient evaluation model was developed following a graded approach to the guidance provided in Regulatory Guide (RG) 1.203 (Reference 1). The non-LOCA phenomena identification and ranking table (PIRT) is described, including a summary of the high-ranked phenomena and how they are assessed. The applicability of NRELAP5 for non-LOCA system transient analysis is assessed. The scope of the non-LOCA system transient analysis is described in this report, as well as interfaces to other safety analysis methodologies. This report describes the selection of appropriately conservative input when applying this EM to perform non-LOCA system transient analyses. Representative transient calculations from application of the EM for the range of non-LOCA events are presented. 1.2 Scope NuScale requests U.S. Nuclear Regulatory Commission (NRC) approval to use the EM described in this report for analyses of NPM design basis non-LOCA events that require system transient analysis. Representative analysis results are provided in Section 8.0 of this report to illustrate results from application of the EM. These representative cases are not necessarily based on final NuScale NPM design inputs, and NRC approval of the representative results is not requested. The scope of this report includes the applicability and acceptability of this methodology to evaluate the primary and secondary system pressure acceptance criteria found in Chapter 15 of the NuScale DSRS and the SRP. This report also describes how the non-LOCA evaluation model interfaces with other analyses that will evaluate acceptance criteria that are not evaluated by the non-LOCA evaluation model. The scope of the NuScale non-LOCA system transient analysis EM is summarized below:
- The non-LOCA evaluation model uses the NRELAP5 code to perform system transient analysis of the NPM design basis events listed in Table 4-1. The general and event-specific analysis methodologies of the EM are presented in Section 7.0.
Sensitivity studies justifying the selected biasing direction are presented in Section 7.2 as part of the event-specific analysis methodology of the EM. The NRELAP5 code is described in the LOCA Evaluation Model (Reference 2). © Copyright 20187 by NuScale Power, LLC 6
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-18 GDC 15 requires that the RCS and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs. TR-0516-49416-P supports the conclusions relative to GDC 15 in the NuScale FSAR. The staff notes that several initial conditions, biases, and conservatisms tables in Section 7.2 of the TR state that the bounding ((
}}2(a),(c)
The staff notes that a (( }}2(a),(c) may not lead to limiting primary and secondary pressure responses due to reduced heat transfer capability in the primary and reduced heat transfer to the secondary. Furthermore, it may be possible that a transient could produce a primary overpressure condition such that (( 2(a),(c)
}} . Provide further justification for the
(( }}2(a),(c) for most transients, and update TR-0516-49416-P as appropriate. NuScale Response: Justification that biased-low initial RCS flow rate is limiting for Chapter 15 events when evaluating peak reactor coolant system pressure and peak steam generator pressure was provided in the response to eRAI 9473, Question 15.02.06-4, in Letter RAIO-0618-60620, dated June 26, 2018. The reactor coolant system pressure response demonstrates that the capacity of the RSV is sufficient to limit peak primary pressure. NuScale Nonproprietary
Impact on DCA: There are no impacts to the DCA as a result of this response. NuScale Nonproprietary
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-19 GDC 10 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs. TR-0516-49416-P supports the conclusions relative to GDC 10 in the NuScale FSAR. The initial conditions, biases, and conservatisms tables in TR Section 7 discuss moderator temperature coefficient, but do not discuss the Doppler temperature coefficient (DTC) as part of the event specific methodologies. The staff notes that the value chosen for DTC affects the reactivity feedback during the transient, which may increase or decrease the severity of the transient and the margin required by GDC 10. Therefore, DTC is an assumption that should be included in consideration of each event-specific methodology. Please update the event-specific methodologies to address the bounding direction for DTC. NuScale Response: As stated in the third paragraph of Section 7.2 of TR-0516-49416-P, only the significant inputs are presented. Thus, while it is agreed that the value chosen for DTC does affect the total reactivity feedback, the magnitude of that effect on the design basis events (DBEs) evaluated using TR-0516-49416-P is minimal. The events that have a greater sensitivity to DTC are the fast reactivity initiated events, such as the control rod ejection event, which are outside the scope of TR-0516-49416-P (Section 1.2). The insensitivity to DTC for the events in the scope of the TR-0516-49416-P methodology arises from the design of the MPS. Specifically, the MPS incorporates multiple signals of high pedigree for each source range and intermediate range channel (Section 7.2.13.1 of TR-0516-49416-P) to trip the reactor before any significant change in fuel temperature occurs following a reactivity initiated event. For all other DBEs evaluated using TR-0516-49416-P, the change in fuel temperature for the average channel is slow relative to the change in moderator temperature. Thus, reactivity feedback from DTC is relegated to a second or third order effect, which is not significant to the transient progression. NuScale Nonproprietary
Impact on DCA: There are no impacts to the DCA as a result of this response. NuScale Nonproprietary
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-20 GDC 10 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs. GDC 13, Instrumentation and control, requires the provision of instrumentation to monitor variables and systems over their anticipated ranges of normal operation, including the effects of AOOs, and of appropriate controls to maintain listed variables and systems within prescribed operating ranges. TR-0516-49416-P supports the conclusions relative to GDC 10 and 13 in the NuScale FSAR. TR Sections 7.2.13, Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions, and 7.2.14, Uncontrolled Control Rod Assembly Bank Withdrawal at Power, state that low power startup conditions exist until the reactor power reaches (( }}2(a),(c). In addition, TR Section 8.3.1 of the same title assumes an initial core power of 15 percent rated thermal power. However, FSAR Tier 2, Section 15.4.1.3.2, Input Parameters and Initial Conditions, for the uncontrolled control rod assembly bank withdrawal from subcritical or low power startup conditions, states that the maximum initial power considered in the analysis is 25 percent, as the high power trip is set at 25 percent of full power for startup conditions. To ensure that future analyses encompass the appropriate operating ranges, address the apparent discrepancy in the scope of low power conditions, and update TR-0516-49416-P as appropriate. NuScale Response: As noted in FSAR Table 15.0-7, the high and low settings for the high power analytical limit are 120 percent rated thermal power (RTP) and 25 percent RTP, respectively. The power range over which each analytical limit is applicable is missing from this table. Thus, FSAR Table 15.0-7 was modified to indicate that the high setting is applicable whenever the core power level is (( }}2(a),(c). One additional modification was made to the first bullet of FSAR Section 15.4.1.3.2 to indicate NuScale Nonproprietary
the upper level for the event analysis is 15 percent RTP, consistent with use of the low setting for the high power analytical limit. Impact on DCA: Table 15.0-7 and Section 15.4.1.3.2 have been revised as described in the response above and as shown in the markup provided with this response. NuScale Nonproprietary
Tier 2 NuScale Final Safety Analysis Report RAI 15.00.02-20 Table 15.0-7: Analytical Limits and Time Delays Signal Analytical Limit Basis and Event Type Actuation Delay High Power 25%, 120% rated This signal is designed to protect against exceeding critical heat flux (CHF) limits for reactivity and 2.0 sec thermal power overcooling events. (RTP)120% rated thermal power (RTP) ( 15% RTP) 25% RTP (<15% RTP) Source and Intermediate 3 decades/min This signal is designed to protect against exceeding CHF and energy deposition limits during startup Variable Range Log Power Rate power excursions High Power Rate +/-15% RTP/min This signal is designed to protect against exceeding CHF limits for reactivity and overcooling events. 2.0 sec High Startup Range Count 5.0 E+05 counts per This signal is designed to protect against exceeding CHF and energy deposition limits during rapid 3.0 sec Rate second startup power excursions. 15.0-55 High Subcritical 32 This signal is designed to detect and mitigate inadvertent subcritical boron dilutions in operating modes 150.0 sec Multiplication 2 and 3. High Reactor Coolant System 610°F This signal is designed to protect against exceeding CHF limits for reactivity and heatup events. 8.0 sec (RCS) Hot Temperature High Containment Pressure 9.5 psia This signal is designed to detect and mitigate RCS or secondary leaks above the allowable limits to 2.0 sec protect RCS inventory and emergency core cooling system (ECCS) function during these events. High Pressurizer Pressure 2000 psia This signal is designed to protect against exceeding reactor pressure vessel (RPV) pressure limits for 2.0 sec reactivity and heatup events. High Pressurizer Level 80% This signal is designed to detect and mitigate chemical and volume control system (CVCS) malfunctions 3.0 sec to protect against overfilling the pressurizer. Transient and Accident Analyses Low Pressurizer Pressure 1720 psia(1) This signal is designed to detect and mitigate primary high energy line break (HELB) outside 2.0 sec containment and protect RCS subcooled margin for protection against instability events. Low Low Pressurizer Pressure 1600 psia(2) This signal is designed to detect and mitigate primary HELB outside containment and protect RCS 2.0 sec subcooled margin for protection against instability events. Low Pressurizer Level 35% This signal is designed to protect the pressurizer heaters from uncovering and overheating during 3.0 sec Draft Revision 2 decrease in RCS inventory events. Low Low Pressurizer Level 20% This signal is designed to detect and mitigate loss-of-coolant accidents (LOCAs) to protect RCS inventory 3.0 sec and ECCS functionality during LOCA and primary HELB outside containment events.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies
- Loss of normal DC power system (EDNS), in addition to loss of normal AC power:
power to the trip breakers is provided via the EDNS. This results in a reactor trip, which terminates bank withdrawal, and therefore is non-limiting.
- Loss of the highly reliable DC power system (EDSS), in addition to loss of both EDNS and Normal AC power: this scenario results in reactor trip and actuation of all ESFs.
This terminates the bank withdrawal, and therefore is non-limiting. There are no single failures that could occur during an uncontrolled CRA withdrawal from a subcritical or low power or startup condition event that would result in more severe conditions for the limiting case. 15.4.1.3 Thermal Hydraulic and Subchannel Analyses 15.4.1.3.1 Evaluation Models The thermal hydraulic analysis of the plant response to an uncontrolled CRA withdrawal from a subcritical or low power or startup condition is performed using NRELAP5. The NRELAP5 model is based on the design features of a NuScale module. The non-loss-of-coolant accident (LOCA) NRELAP5 model is discussed in Section 15.0.2. The relevant boundary conditions from the NRELAP5 analyses are provided to the downstream subchannel critical heat flux (CHF) analysis. RAI 15.04.01-6 The subchannel core CHF analysis is performed using VIPRE-01. VIPRE-01 is a subchannel analysis tool designed for general-purpose thermal-hydraulic analysis under normal operating conditions, operational transients, and events of moderate severity. Limiting axial and radial power shapes are used in the subchannel analysis to ensure a conservative MCHFR result, in accordance with the methodology described in Reference 15.4-1. See Section 15.0.2.3 for a discussion of the VIPRE-01 code and evaluation model. 15.4.1.3.2 Input Parameters and Initial Conditions RAI 15.04.01-3 A spectrum of initial conditions is analyzed to find the limiting reactivity insertion due to an uncontrolled CRA withdrawal from a subcritical or low power or startup condition. The initial conditions of the transient evaluation result in a conservative calculation. Table 15.4-2 and Table 15.4-26 provide key inputs and the associated biases for the limiting MCHFR and RCS pressure cases, respectively. The following initial conditions and assumptions ensure that the results have sufficient conservatism. RAI 15.00.02-20
- The minimum initial power assumed for this analysis is 1 Watt. The transient analyses for this event evaluate cases with initial powers ranging from this minimum power of 1 Watt to 2515% of full power (consistent with use of the low setting for the high power analytical limit). The SRP guidance states that minimizing initial power provides the most conservative conditions for a CRA withdrawal from a subcritical or low power because it provides the maximum Tier 2 15.4-3 Draft Revision 2
Response to Request for Additional Information Docket: PROJ0769 eRAI No.: 9513 Date of RAI Issue: 05/08/2018 NRC Question No.: 15.00.02-21 GDC 10 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs. GDC 15 requires that the RCS and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs. TR-0516-49416-P supports the conclusions relative to GDC 10 and 15 the NuScale FSAR. The information in TR-0516-49416-P needs to be accurate and consistent so the staff is able to make a reasonable assurance finding. The staff noted that TR-0516-49416-P contains the following apparent typographical errors that affect technical meaning or details: Word(s) seem to be missing from a sentence in the last paragraph of TR Page 29, which states, After DHRS is actuated, the in the DHRS flows into the SG. Section 5.1.4.29, (( }}2(a),(c), states that the primary pressure is one of the FOMs [figures of merit] in Phase 3. However, Section 5.1.3 of the TR states that primary pressure is only a FOM in Phases 1 and 2. Figures 5-30 and 5-31 appear to be identical despite claiming to show mid and upper temperatures, respectively. The same is true for and Figures 5-36 and 5-37. It appears the title of Section 5.3.5.2, 5.3.5.2 Helical Coil Steam Generator Modeling (( }}2(a),(c) is incorrect, as the section discusses modeling at ((
}}2(a),(c).
It appears that the numbering of components, junctions, etc. on the nodalization diagram in Figure 6-2 is incorrect based on comparison to the NRELAP5 basemodel described in EC-A010-1782, NuScale NRELAP5 Module Basemodel. Please address the above items by either (1) updating TR-0516-49416-P to correct them or (2) justifying why the information is accurate. NuScale Nonproprietary
NuScale Response: The sentence in question, which appears in the sixth paragraph of Section 4.2 to TR-0516-49416-P, was corrected by adding the word "fluid" after "After DHRS is actuated, the" as presented at the end of this RAI response. As submitted, TR-0516-49416-P, Section 5.1.4.29 showed that both the Upper Plenum and the Pressurizer were applicable for Phase 3 (third sentence). However, a review of the PIRT indicates that the two-phase level swell high ranking phenomenon applies for the Upper Plenum during Phase 3 and the Pressurizer during Phase 2, which is consistent with Section 5.1.3 of TR-0516-49416-P. Therefore, the description provided in Section 5.1.4.29 of TR-0516-49416-P was revised as presented at the end of this RAI response. A review of TR-0516-49416-P indicates that Figures 5-30 and 5-31 are identical. Figure 5-31 of TR-0516-49416-P was replaced with the correct figure as presented at the end of this RAI response. The title to Section 5.3.5.2 of TR-0516-49416-P is incorrect for the reasons noted. The title of Section 5.3.5.2 of TR-0516-49416-P has been corrected as presented at the end of this RAI response. As noted in the first paragraph of Section 6.1 of TR-0516-49416-P, Figure 6-2 is meant to convey the overall model structure rather than show nodalization details of any particular component. Therefore, no correction to TR-0516-49416-P is necessary as a result of this response. Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss of Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 4.2 Design Basis Event Acceptance Criteria Safety analyses are performed to demonstrate that a nuclear power plant can meet applicable acceptance criteria for a limiting set of AOOs, IEs, and accidents. If the risk of an event is defined as the product of the events frequency of occurrence and its consequences, then the design of the plant should be such that events produce about the same level of risk. The acceptance criteria indicated by the GDC for nuclear power plants (Reference 4) reflect the risk of an event. Relatively frequent events such as AOOs are prohibited from resulting in serious consequences, but relatively rare events (postulated accidents) are allowed to produce more severe consequences. Design basis events for the NPM are categorized as AOOs, IEs, or postulated accidents. Table 4-2, Table 4-3, and Table 4-4 summarize the acceptance criteria applied for AOOs, IEs, and postulated accidents, respectively. The applicable acceptance criteria identified for each event are based on the event classification as identified in Table 4-1. For a limited number of events, a more conservative acceptance criterion may be applied than required based on the event classification. For many non-LOCA transient events, the specific acceptance criterion will not be challenged during the event progression. For example, events that result in an increase in heat removal from the RCS may have a maximum RCS pressure higher than the initial operating pressure, but will not challenge the margin to the maximum RCS pressure acceptance criterion. In contrast, events that result in a decrease in heat removal from the RCS may result in an RCS pressurization that could challenge the maximum RCS pressure acceptance criterion. In Section 7.2, the acceptance criteria of interest for each non-LOCA event are identified. The acceptance criteria of interest are those where margin to the limit may be challenged during the event progression. In the event-specific transient analysis, sensitivity calculations are performed as necessary to ensure that the event meets acceptance criteria that may be challenged. These sensitivity calculations are performed to confirm that appropriately conservative inputs are specified to identify the case that results in minimum margin to the acceptance criterion of interest. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed are sufficient to demonstrate that margin to the acceptance criterion is maintained. A prime example of an acceptance criterion where the NPM design has significant margin is the maximum secondary system pressure. Unlike in typical PWR designs, in the NuScale design, the design pressure of the SG secondary side up to the second containment isolation valves is equal to the RCS design pressure. This feature supports the design and operation of the SG and DHRS. In a non-LOCA event that results in DHRS actuation, typically the maximum secondary side pressure occurs in the first minutes of the transient progression, following DHRS actuation. After DHRS is actuated, the fluid in the DHRS flows into the SG. Heat is transferred from the RCS primary system to the SG, where the DHRS loop inventory boils in the SG tubes. The steam flow © Copyright 20187 by NuScale Power, LLC 30
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 5.1.4.28 ((
}}2(a),(c) 5.1.4.29 ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 61
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 ((
}}2(a),(b),(c),ECI Figure 5-30 NIST-1 HP-04-02 mid cooling pool vessel fluid temperatures
((
}}2(a),(b),(c),ECI Figure 5-31 NIST-1 HP-04-02 upper cooling pool vessel fluid temperatures © Copyright 20187 by NuScale Power, LLC 129
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Rev. 12 5.3.5 Steam Generator Modeling 5.3.5.1 Background NuScales LOCA Topical Report (Reference 2) Section 7.3 discusses the validation of NRELAP5 for helical coil SG modeling. The validation was mainly against SIET TF-1 and TF-2 test data. It was concluded that NRELAP5 showed reasonable to excellent agreement with test data for all phenomena at conditions that are important to non-LOCA analysis, as discussed in the following sections. The validation is further investigated in this report to ensure the unique characteristics of the non-LOCA transients (comparing to LOCA) are identified and evaluated. Specifically, this investigation ensures the operating ranges expected during the non-LOCA transients are covered by the validated ranges. The modeling issues identified in the LOCA topical report are evaluated to ensure they do not affect the non-LOCA transients. 5.3.5.2 Helical Coil Steam Generator Modeling ((
}}2(a),(c)
In the LOCA topical report (Reference 2), NRELAP5 shows reasonable to excellent agreement with test data on the helical coil SG primary side, except ((
}}2(a),(c)
Figure 5-154 through Figure 5-159 show the secondary and primary flow rates for several cases of steam line breaks outside the CNV. ((
}}2(a),(c)
Figure 5-160 and Figure 5-161 show the secondary and primary flow rates for double-ended guillotine steam line break inside the CNV. ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 222
RAIO-0718-60781 : Affidavit of Zackary W. Rad, AF-0718-60782 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com
NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows:
- 1. I am the Director, Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale.
- 2. I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following:
- a. The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale.
- b. The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit.
- c. Use by a competitor of the information requested to be withheld would reduce the competitor's expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
- d. The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale.
- e. The information requested to be withheld consists of patentable ideas.
- 3. Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScale's competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the method by which NuScale develops its non-loss of coolant accident analysis methodology.
NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. AF-0718-60782
- 4. The information sought to be withheld is in the enclosed response to NRC Request for Additional Information No. 9513, eRAI 9513. The enclosure contains the designation "Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document.
- 5. The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4).
- 6. Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld:
- a. The information sought to be withheld is owned and has been held in confidence by NuScale.
- b. The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale.
The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality.
- c. The information is being transmitted to and received by the NRC in confidence.
- d. No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence.
- e. Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry.
NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on July 9, 2018. Zackary WW. Rad AF-0718-60782}}