ML18264A341

LLC - Response to NRC Request for Additional Information No. 441 (Erai No. 9485) on the NuScale Design Certification Application

ML18264A341
Person / Time
Site:	NuScale
Issue date:	09/20/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
Shared Package
ML18264A340	List: ML18264A341
References
AF-0918-61915, RAIO-0918-61914
Download: ML18264A341 (17)

Text

RAIO-0918-61914 September 21, 2018 Docket No.52-048 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.

441 (eRAI No. 9485) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

441 (eRAI No. 9485)," dated April 30, 2018 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 Question from NRC eRAI No. 9485:

• 15-6 is the proprietary version of the NuScale Response to NRC RAI No. 441 (eRAI No.

9485). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 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.

If you have any questions on this response, please contact Paul lnfanger 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 Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-0918-61914 : NuScale Response to NRC Request for Additional Information eRAI No. 9485, proprietary : NuScale Response to NRC Request for Additional Information eRAI No. 9485, nonproprietary : Affidavit of Zackary W. Rad, AF-0918-61915 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0918-61914 :

NuScale Response to NRC Request for Additional Information eRAI No. 9485, proprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0918-61914 :

NuScale Response to NRC Request for Additional Information eRAI No. 9485, 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.: 9485 Date of RAI Issue: 04/30/2018 NRC Question No.: 15-6 Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Section 47 requires a final safety analysis report (FSAR) to analyze the design and performance of the structures, systems, and components (SSCs). Safety evaluations, performed to support the FSAR, include accident analyses to (1) demonstrate that specified acceptable fuel design limits (SAFDLs) are not exceeded during normal operation, including the effects of anticipated operational occurrences (AOOs), and (2) determine the number of fuel failures associated with critical heat flux (CHF) that need to be included in the radiological consequences for postulated accidents.

As the return to power analysis in FSAR 15.0.6 can occur, assuming a stuck rod, within a few hours from either an AOO or postulated accident initiating event, the AOO acceptance criteria of General Design Criterion (GDC) 10 applies. GDC 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs.

Consistent with Regulatory Guide 1.203, Transient and Accident Analysis Methods, the adequacy of the evaluation model for the expected phenomena and range of conditions should be assessed and comprehensive documentation should be provided for staff review.

In response to RAI 8771, the applicant provided updated FSAR Section 15.0.6.3.1, Evaluation Models, which provides an overview of the methods used in the return to power analyses. The response to RAI 8771 indicates that the non-loss of coolant accident (LOCA) NRELAP5 model is used to determine the maximum return for a decay heat removal system (DHRS) cooldown while the LOCA NRELAP5 model is used to calculate the minimum critical heat flux ratio (MCHFR). The staff has determined the level of detail associated with the analysis methodology used in the return power analysis is not consistent with Regulatory Guide 1.203 and hence is unable to make a safety finding relative to GDC 10.

NuScale Nonproprietary

As such, the staff is requesting the applicant provide details associated with changes from the non-LOCA and LOCA NRELAP5 models for the staff to assess the adequacy to predict the peak return to power and MCHFR. Details should include, any changes to model nodalization, the methods used to determine the reactivity coefficients, hot rod/channel model, CHF correlations used and how the MCHFR is determined. Reference to existing non-LOCA and LOCA topical reports is acceptable for modeling details which remain unchanged in the return to power analyses.

In addition to providing the documentation associated with changes to the models, the staff is requesting justification of the adequacy of the return to power models to predict key figures of merit (peak power and MCHFR). As with the documentation request, the validation that supports the adequacy of the return to power models can reference the applicable non-LOCA and LOCA topical reports.

NuScale Response:

As described in FSAR section 15.0.6 the a return to power for a generic cooldown event is analyzed in two separate ways depending on whether or not the ECCS valves are assumed to open during the transient. First, the system was allowed to cool, resulting in a power overshoot, before reaching an eventual steady state power level. This scenario is consistent with the non-LOCA event progression where core cooling is provided by the decay heat removal system. The second scenario assumes the ECCS valves open simultaneous to the power peak in order to provide a conservative CHF analysis which is consistent with the inadvertent RPV valve opening scenario.

Nodalization Scheme Excluding Reactor Core The overcooling return to power (OCRP) event methodology prescribes the use of both the LOCA and non-LOCA methodologies depending on the event sequence. Since the focus of the event is analysis of CHF in the core using the inadvertent reactor pressure vessel (RPV) valve opening methodology, the modeling of the rest of the RCS loop consistent with either the LOCA or non-LOCA models is sufficient. The latest version of the OCRP analysis used an updated base model with a nodalization consistent with an inadvertent RPV valve opening model supporting the methodology presented in Appendix B of the LOCA Evaluation Model Topical Report (LTR), TR-0516-49422.

NuScale Nonproprietary

Core Modeling and Nodalization The core modeling and nodalization used in the OCRP analysis is consistent with that presented in the LOCA LTR. The hydraulic volumes and corresponding heat structures for the lower plenum, hot assembly, average core, core bypass, and upper plenum were copied from the NRELAP5 LOCA model specifically developed to be consistent with the LOCA LTR, and incorporated into the OCRP model. There are 37 assemblies in the NPM core, and given a uniform radial power distribution, each assembly would have a radial peaking factor of 1.0. For the OCRP analysis, ((2(a),(c) The OCRP analysis was analyzed with the generic bottom, middle, and top-peaked axial power profiles presented in the LOCA LTR. The top-peaked axial power shape produced the limiting results for the OCRP analysis. As a result of using the top-peaked axial power shape, MCHFR results for the OCRP event changed as shown in the markup of the FSAR, Section 15.0.6 at the end of this response and consistent with those submitted in response to eRAI 9487 in NuScale letter RAIO-0718-60857, dated July 13, 2018. Reactivity Coefficient Determination The methods for determining reactivity coefficients for input into the OCRP analysis differs depending on whether the ECCS valves are assumed to actuate. For the analysis without ECCS actuation, the OCRP transient behaves in a manner consistent with a non-LOCA transient. As such, the reactivity coefficients are determined using the methodology from the Non-LOCA Analysis Methodology LTR (TR-0516-49416). The moderator reactivity feedback is analyzed with the most negative moderator temperature coefficient (MTC) for reactor power below 25% RTP, at a value of -15 pcm/F. The fuel reactivity feedback is analyzed with the least negative doppler temperature coefficient (DTC) at -1.4 pcm/F, with the fuel temperature determined on a volume averaged basis. For the analysis with ECCS actuation, the OCRP transient behaves in a manner consistent with the inadvertent PRV valve opening transient following the actuation of the ECCS valves. Up the point of actuation the reactivity mechanisms are identical to those presented previously for the analysis without ECCS activation. However, following the actuation of the ECCS valves, modeling of the moderator reactivity feedback mechanism changes from being defined with MTC as specified by the non-LOCA LTR, to being defined with the moderator density coefficient NuScale Nonproprietary

(MDC) as specified by the LOCA LTR. The MDC is determined ((

                                                                          }}2(a),(c) The OCRP event is only possible due to the small amount of boron in the moderator; therefore the MDC is most appropriate for this analysis. The calculation of DTC is unchanged from the analysis without ECCS actuation.

CHF Correlations The NRELAP5 heat transfer option ((

                                                           }}2(a),(c) The CHF evaluation acceptance limits used are consistent with those explained in Appendix B, Section 5.3 of the LOCA LTR as submitted in the response to RAI 9536, question 15.6.6-2 in letter RAIO-0918-61859, dated September 21, 2018. The MCHFR margin is determined by comparing the calculated MCHFR against the evaluation limits for the appropriate flow range as specified in Appendix B, Section 5.3 of the LOCA LTR.

Adequacy of the Methodology Adequacy of modeling the return to power - Similar to non-LOCA events, the primary conservatism is derived from an evaluation of core reactivity balance to ensure a conservative peak power is generated. Peak power is confirmed to be conservative by ((

      }}2(a),(c) This evaluation demonstrates that NRELAP5 predicts a significant over prediction of the peak power due to the DHRS cooldown scenario.

Adequacy of modeling the MCHFR - The MCHFR calculation is performed using the methodology from Section 5.3 of Appendix B of the LOCA LTR. The appendix pertains to the analysis of the Inadvertent Opening of a RPV Valve (IORV). MCHFR in the OCRP analysis occurs at the time of peak power within seconds of the ECCS valves opening, which causes rapid depressurization and voiding in the core. The behavior of the OCRP is consistent with that of the IORV analysis. The IORV analysis presented in the Appendix of the LOCA LTR uses essentially the same NPM model and core as the OCRP analysis. Therefore the OCRP methodology is deemed adequate to calculate the MCHFR for that event particularly within the context of the conservatism of the peak power calculation. Furthermore, the analysis of MCHFR NuScale Nonproprietary

for the OCRP event is deemed sufficient on the basis that it is less limiting than the IORV event. By definition the OCRP occurs with significant inlet subcooling, and low core power levels, both of which contribute to significant margin to CHF. The MCHFR experienced in the OCRP analysis is higher than the CHFR calculated for the IORV event at steady state conditions prior to event initiation. Conclusion The OCRP model applies the relevant LOCA and nonLOCA core modeling methodologies, including nodalization, reactivity coefficients, power peaking/distribution and CHF models to conservatively predict both the power peak and the MCHF that occurs due to the ECCS transition event progression. Additionally, the 95/95 Hench Levy CHF limit is applied in the analysis consistent with the IORV methodology presented in the LOCA LTR. Impact on DCA: Section 15.0.6 has been revised as described in the response above and as shown in the markup provided with this response. NuScale Nonproprietary

NuScale Final Safety Analysis Report Transient and Accident Analyses presented: return to power with and without transition from DHRS cooling to ECCS cooling. The sequence of events for an overcooling return to power event with DHRS cooling is provided in Table 15.0-16 and with the transition to ECCS cooling is provided in Table 15.0-17. RAI 15-1 For the overcooling return to power event, it is assumed that a reactor trip occurs at end of cycle (EOC) with the most reactive control rod stuck out of the core. The subsequent DHRS cooldown is left unmitigated and boron addition does not occur. While there are simple operational means for mitigating the DHRS extended cooldown and thereby eliminating the need for boron addition, operator action is not credited for either mitigating the cooldown or adding boron, consistent with Section 15.0.0.6.4. RAI 15-1 The overcooling return to power event assumes a reactor trip coincident with the loss of normal AC power as the initiating event. This analysis concerns the post-reactor trip return to power; therefore, the MPS is not specifically credited. RAI 15-1, RAI 15-1S1 In the event that the highly reliable DC power (EDSS) is available, the reactor cools down on DHRS and ECCS is not actuated. If EDSS is unavailable concurrent with the initiating event, ECCS would be actuated while RCS pressure is above the IAB release pressure, and the ECCS valves would not initially open. During an extended DHRS cooling event, RCS pressure decreases due to reactor pressure vessel (RPV) heat loss and reactor coolant system (RCS) shrinkage causing an expansion of the pressurizer vapor space. Although unlikely, if the initial pressurizer pressure and level were sufficient, it is possible to postulate an IAB release concurrent with the overcooling return to power peak. This scenario generates the most challenging CHF conditions and is presented as the transition to ECCS cooling scenario. RAI 15-1 15.0.6.3 Thermal Hydraulic and Critical Heat Flux Analyses 15.0.6.3.1 Evaluation Models RAI 15-6 The transient evaluations are performed in separate parts. First, the peak power portion of the analysis, where EDSS is available, is analyzed using the non-LOCA NRELAP5 model modeling methodology discussed in Section 15.0.2described in Reference 15.0-5. The purpose of the peak power analysis is to demonstrate the limited magnitude of the return to power, to characterize the event should DHRS cooling be sustained and to examine the various sensitivities that influence the moderator temperature-driven power response to inform the CHF modeling of the appropriate case to simulate. RAI 15-1, RAI 15-6 Tier 2 15.0-40 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses The MCHFR portion of the analysis, where EDSS is unavailable, uses the LOCA NRELAP5 modelmodeling methodology. The CHF correlation applied in the LOCA evaluation model discussed in Reference 15.0-3 is evaluated against the 95/95 CHFR acceptance criterion of an AOO, as described in Section 15.6.6Reference 15.0-3. RAI 15-1 15.0.6.3.2 Input Parameters and Initial Conditions As stated above, this event is analyzed specifically for the parameters that generate the most severe overcooling return to power core power event. A bounding DHRS cooldown following a DBE is evaluated with conservative assumptions to maximize the rate of reactivity insertion during a return to power to maximize the peak power. The following assumptions, for the case with EDSS available, ensure that the results have sufficient conservatism. RAI 15-1

• The reactor is at hot zero power (HZP) for the initial condition. The core power response is due to the moderator temperature-driven reactivity insertion that creates a bounding power overshoot that is several times larger than the eventual steady state power level. From a HZP initialization, the RCS shrinkage does not impede cooldown rate due to the much higher initial RCS density.

Additionally, the HZP initial condition will tend to have lower decay heat levels and lower initial RCS temperature than higher power initializations resulting in a faster cooldown. RAI 15-1

• The most negative HZP moderator temperature coefficient (-15 pcm/°F) is used, as it will produce a bounding rate of increase in moderator reactivity worth during cooldown.

RAI 15-1

• The least negative Doppler coefficient (-1.40 pcm/°F) is used because it results in the least strong negative reactivity feedback during the return to power, bounding the maximum peak power for the transient.

RAI 15-1

• Uniform radial and axial moderator and Doppler reactivity feedback weighting is applied to ensure the power response is not suppressed due to the local heating effects.

RAI 15-1

• The reactor is shut down with an assumed minimum required shutdown margin of two percent at 420 degrees Fahrenheit. A minimum shutdown margin allows for a return to power early in the cooldown transient while the RCS cools down at a higher rate.

RAI 15-1

• The DHRS heat transfer is increased by 30 percent to ensure the consequences of the cooldown are maximized after DHRS actuation.

RAI 15-1 Tier 2 15.0-41 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses The overcooling return to power event begins with an initial negative reactivity insertion that is gradually removed as the transient progresses until a return to power occurs (Figure 15.0-8 and Figure 15.0-10). The biased initial conditions, with increased heat transfer, and low pool temperature results in a slightly larger return to power. A sensitivity analysis on background decay heat shows a minor sensitivity, where higher decay heat results in a slightly slower event progression and marginally decreased peak power. Initial negative reactivity insertion also has little impact on the calculated peak power. Therefore, it is concluded that protection of shutdown margin is insignificant for this event. RAI 15-1 The return to power (Figure 15.0-8 and Figure 15.0-10) occurs more than two hours from the start of the transient, meaning the initiating event has little impact on the event progression and results. These cases are initiated from HZP conditions, which subsumes initiation events from hot full power (HFP) conditions due to the nature of the reactivity balance. The time of return to power would be greatly delayed with realistic decay heat levels and an event initiating from HFP conditions. RAI 15-1 With the initial negative reactivity insertion, the average RCS temperature (Figure 15.0-11), average fuel temperature (Figure 15.0-12), and RPV pressure (Figure 15.0-13) decrease until the return to power occurs. At the time of the return to power, the temperature and pressure increase slightly and then either stabilize at a low value or continue to decline. RAI 15-1 For the limiting MCHFR event, which is a loss of highly reliable DC power (EDSS) resulting in a transition to ECCS cooling, the CHFR is presented in Figure 15.0-14. This analysis considers the short term transition period to demonstrate that even at elevated local power distributions, a transition to ECCS cooling is not a safety concern. Further analysis demonstrates that once ECCS equilibrium conditions are established, the density reactivity feedback is sufficient that even very low heat levels will suppress the critical power response. Therefore, it is concluded that the limiting condition for MCHFR is at the time of the ECCS transition when the core power levels are much higher than the equilibrium ECCS cooling power levels. Reactor power, RCS flow rate, and hot channel heat flux are provided in Figure 15.0-15, Figure 15.0-16, and Figure 15.0-17, respectively. RAI 15-1S1, RAI 15-6 For the peak power case, recriticality occurs at 6750approximately 6500 seconds and the peak power (1416 MW) occurs at 7850approximately 7900 seconds. As the temperature of the RCS increases, power decreases until equilibrium power (3 MW) is reached at 12000 seconds. This scenario results in the highest power level but is not limiting for MCHFR or other parameters. RCS Power is shown in Figure 15.0-18 and RCS flow in Figure 15.0-19. RAI 15-1 Tier 2 15.0-43 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses 15.0.6.3.4 Conclusions The AOO acceptance criteria outlined in Table 15.0-2 are used as the basis for the overcooling return to power event. The acceptance criteria, followed by how the NuScale design meets them, are listed below: RAI 15-1

1) Potential core damage is evaluated on the basis that it is acceptable if the minimum departure from nucleate boiling ratio (DNBR) remains above the 95/

95 DNBR limit. Minimum critical heat flux ratio is used instead of minimum DNBR, as described in Section 4.4.2. RAI 15-6 Fuel integrity is not challenged by an overcooling return to power event. The MCHFR is 2.251.9 and is shown in Figure 15.0-14. The MCHFR for evaluated cases occurs in the Hench-Levy correlation flow range, therefore the 95/95 design limit is 1.1221.13. The CHF analysis confirms that the DHRS overcooling return to power event can safely transition to ECCS cooling without challenging MCHFR limits. RAI 15-1

2) RCS pressure should be maintained below 110 percent of the design value.

Due to the nature of the overcooling return to power event, primary pressure is not challenged and is non-limiting for this event. RAI 15-1

3) The main steam pressure should be maintained below 110 percent of the design value.

Due to the nature of the overcooling return to power event, main steam pressure is not challenged and is non-limiting for this event. RAI 15-1

4) The event should not generate a more serious plant condition without other faults occurring independently.

The overcooling return to power analysis demonstrates that DBEs, where a most reactive control rod is assumed stuck out upon reactor trip, can be safely cooled by DHRS, or DHRS transitioning to ECCS cooling, without challenging MCHFR limits. Additionally, return to power scenarios with extended ECCS core cooling are limited by the density reactivity feedback as generated by the boiling in the core such that these scenarios are well bounded by the DHRS transition event due to the relative power levels in the core. RAI 15-1 The evaluation of an overcooling return to power event demonstrates that design limits are not exceeded and the overcooling return to power event is non-limiting with respect to DBEs. Tier 2 15.0-44 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6 Table 15.0-16: Sequence of Events for Overcooling Return to Power Event EDSS Available (Peak Power Case) Event Time [s]* Start of Transient 0.0 DHRS Actuation 2.1 Time of recriticality 65006750 Maximum Return to Power (164 MW) 79007850 Equilibrium power reached (3 MW) 12000 Note:

                     *Time is rounded.

Tier 2 15.0-72 Draft Revision 2

RAIO-0918-61914 : Affidavit of Zackary W. Rad, AF-0918-61915 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 performs its overcooling return to power analysis.

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-0918-61915

4. The information sought to be withheld is in the enclosed response to NRC Request for Additional Information No. 441, eRAI No. 9485. 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 September 21, 2018. Zackary W. Rad AF-0918-61915}}