ML19215A006

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LLC - Supplemental Response to NRC Request for Additional Information No. 441 (Erai No. 9485) on the NuScale Design Certification Application
ML19215A006
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Site: NuScale
Issue date: 08/02/2019
From: Bergman T
NuScale
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RAI 0-0819-66538
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RAI 0-0819-66538 August02,2019 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 Supplemental Response to NRC Request for Additional Information No. 441 (eRAI No. 9485) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 441 (eRAI No. 9485)," dated April 30, 2018
2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 441 (eRAI No.9485)," dated September 21, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosures to this letter contain NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9485:

  • 15-6 is the proprietary version of the NuScale Supplemental 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 Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.

Sincerely, Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-0819-66538 : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9485, proprietary : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9485, nonproprietary : Affidavit of Thomas A. Bergman, AF-0819-66539 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0819-66538 :

NuScale Supplemental 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-0819-66538 :

NuScale Supplemental 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 NuScale Nonproprietary

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.

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:

The original NuScale response as submitted in NuScale correspondence RAIO-0918-61781 and dated September 13, 2018, was augmented in NuScale correspondence RAIO-0319-64867 dated March 14, 2019 is further being supplemented with the following return to power evaluation of the full 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> long term cooling coping period.

  • Return to Power Overview Given an unmitigated cooldown after reactor trip at late in cycle reactor coolant system (RCS) boron conditions, it is possible that a limited return to power could occur if the highest worth control rod is assumed stuck out of the core due to the slow decay of xenon. A description of the analysis of this condition is presented to support the conclusion that the return to power is not a safety concern.

The mechanisms for loss of shutdown margin following an event that leads to a reactor trip are due to either moderator cooling or boron dilution in the core. The design of the module protection system includes consideration for boron dilution and automatically isolates the normal chemical volume control system (CVCS) dilution source upon reactor trip precluding the possibility of an inadvertent boron dilution due to CVCS malfunction or Operator error. The possibility of boron redistribution is also a consideration and has been evaluated in response to NuScale Nonproprietary

RAI 8930 with the conclusion that under emergency core cooling system (ECCS) conditions, boron will preferentially be transported to the core region due to the boiling and condensing heat removal design of the ECCS.

Examination of both boron dilution and boron redistribution supports the conclusion that net loss of boron in the RCS core region is not possible for the design basis coping period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

This necessarily reduces the period of time in which an extended cooling return to power could occur. Early in the fuel cycle there is excess reactivity in the fuel that must be balanced with boron in the moderator. As the core exposure increases, the excess reactivity in the fuel decreases, and less boron is required in the moderator to maintain a critical core configuration.

The moderator temperature coefficient (MTC) varies with boron concentration; at higher boron concentrations the MTC is less negative or slightly positive and prevents the reactor from achieving recriticality as it cools. This behavior is observed by comparing the critical boron concentration (CBC) for a nominal cycle depletion at hot full power (HFP) to the critical boron concentrations with an assumed worst rod stuck out (WRSO) presented in Table 1. Note that except for the nominal values, the CBC values presented in Table 1 are evaluated at zero power, with no xenon, a WRSO, subcooled moderator conditions, and include the ((2(a),(c) CBC nuclear reliability factor (NRF) for additional conservatism. Analysis presented in later sections demonstrates that within the 72 hour coping window, the decay heat maintains the moderator temperature above (( }}2(a),(c) precluding a recriticality for BOC or MOC burnups. However, recriticality is possible for an EOC core where initial boron concentrations are very low, under this combination of very conservative assumptions. Table 1 Critical Boron Concentration Comparison Boron Concentration (ppm) BOC MOC EOC (( 

                                                                                                    }}2(a),(c)

The low nominal boron concentration for the EOC core (5 ppm) corresponds with the equilibrium negative reactivity from the buildup of fission products, primarily xenon. The EOC core has approximately (( }}2(a),(c) of reactivity due to the presence of xenon. If there was an extended shutdown and xenon decayed away, a reactor restart would require an additional approximately (( }}2(a),(c) of boron to balance the previous reactivity from xenon no longer in the core . A return to power would be precluded from these initial conditions since the boron NuScale Nonproprietary

concentration is larger than the EOC (( }}2(a),(c) WRSO critical boron concentration presented in Table 1. Therefore it is concluded that maximizing the initial xenon concentration will result in the maximum available post-trip reactivity insertion and resultant critical power level. A similar analysis can be extended to a gradual down power to hot zero power (HZP) from HFP at EOC conditions. Assuming the core is at EOC HFP conditions, any reduction in power or moderator temperature will add positive reactivity due to moderator temperature and Doppler feedback effects. In order to maintain exact critical conditions at HZP, it would be necessary to increase the boron concentration. The power dependent insertion limits prevent the shutdown bank from being inserted when the reactor is in Mode 1 (critical); the regulating bank may be partially inserted in accordance with the insertion limits. The available negative reactivity to be inserted by the control rods while remaining operational is not sufficient to down power the core to HZP conditions. CRA insertion must be supplemented with boron addition; any increase in boron concentration reduces the magnitude and increases the time to achieve any potential recriticality. For the reasons described in the preceding paragraphs, it is concluded that the return to power event is most limiting if: (i) the core is at EOC exposure levels, (ii) equilibrium fission products are present initially (no preceding power maneuvers), and (iii) the reactor is initially at HFP conditions. An initiation from any other conditions either leads to a lower equilibrium power, or recriticality is not possible.

  • Methodology The evaluation of the return to power is performed in the following steps and is discussed in greater detail in the following paragraphs:
1. A spectrum of (( }}2(a),(c) NRELAP5 state point calculations are performed in order to characterize the RCS temperature a function of available decay heat, pool temperature, and passive cooling mode (i.e. DHRS or ECCS).
2. A spectrum of (( }}2(a),(c) SIMULATE5 state points are calculated to characterize the WRSO critical power level as a function of RCS temperature, flow rate, and void fraction.
3. The NRELAP5 and SIMULATE5 calculation results are compared to conservatively approximate the critical power level for both ECCS and DHRS overcooling events.
4. The critical power level and peaking are used to perform a conservative evaluation of CHF margin using (( }}2(a),(c).

NuScale Nonproprietary

5. The relative time to recriticality is determined using the transient xenon worth following a reactor scram.

Step one of this analysis considers three separate evaluations of DHRS cooling and an evaluation of cooling while in ECCS heat removal. (( 

                                                }}2(a),(c). Under nominal cooldown conditions from HFP, the coolant shrinkage results in the collapsed water level falling below the top of the riser referred to as the riser uncovery phenomenon. For the DHRS cooling cases, maintaining the water level in the pressurizer is conservative from a recriticality perspective as the loss of natural flow path increases the thermal resistance and increases the RCS temperature in the core region by reducing the RCS flow rate. In addition, both the adiabatic riser and the default riser uncovery modeling approaches described previously in RAI-9508 are evaluated for the potential of a return to power.

The cooldown modes are evaluated in the NRELAP5 state-point calculations (( 

                            }}2(a),(c). Of the ECCS and DHRS decay heat removal modes evaluated, the covered riser DHRS cases produce the maximum flow rate. The data from these cases is used to develop ((                                             }}2(a),(c) which is used in developing the critical power line from SIMULATE5 state-point calculations.

The second step of this analysis uses SIMULATE5 calculations to iteratively solve for the critical power level that, for a given set of inlet conditions, results in an exactly critical core configuration. The cases model an EOC core with WRSO, zeroed xenon and iodine concentrations, and a boron concentration of 5 ppm (nominal CBC at EOC). (( 

                                                         }}2(a),(c).

For a given core inlet temperature, the critical power calculated by SIMULATE5 (( 

                                                                                                       }}2(a),(c).

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(( 

                                                                                                }}2(a),(c). Evaluating the equilibrium power for the cooldown modes, using the critical power line developed from the riser-covered cases results in a very conservative equilibrium power level for use in the final MCHFR determination. The line of criticality is compared to the NRELAP5 data as shown in Figure 1.

For step three, the critical power line generated from the SIMULATE5 results is plotted with the minimum core moderator temperature results from the NRELAP5 state-point calculations. By identifying the intersection of the data, the equilibrium power for a given pool temperature and cooldown mode is determined. In step four, the NRELAP5 thermal-hydraulic conditions at the equilibrium power levels are investigated to determine the minimum pressure in the core. The minimum pressure is used to calculate saturation properties that are used to evaluate the (( 

            }}2(a),(c). While the critical power level is determined from subcooled thermal hydraulic conditions, the CHF evaluation applies ((                  }}2(a),(c) void fraction as a conservatism. This corresponds with the amount of voiding that would be required to keep the core subcritical. The peak heat flux is determined from considering a conservative dynamic overshoot of the equilibrium power level, and considering a conservative total pin peaking factor associated with the WRSO.
  • Calculations Figure 1 summarize the results from steps 1 and 2 for the DHRS scenarios. The data points are the NRELAP5 state point calculation results which show a decreasing equilibrium RCS temperature as decay heat decreases. For comparison, a 72 hour DHRS transient NRELAP5 calculation was completed with the riser covered to confirm the state-point calculations conservatively represent the relationship between core heat and RCS temperature. Additional state-point results are included on the figure to demonstrate the conservatively lower temperatures observed for the covered riser DHRS cooling. As described above, the two modeling approaches were used to evaluate uncovered riser cases. Figure 1 presents the results using an (( }}2(a),(c); however, the ((
               }}2(a),(c) is reported instead of the core average value to ensure conservatism.

The black line represents the (( 

              }}2(a),(c). The point of the intersection determines the ((
                                                                       }}2(a),(c). If the natural flow path is NuScale Nonproprietary

maintained, the equilibrium power level is approximately 0.5% rated thermal power (RTP) with an average moderator temperature of 180oF. Both the (( 

          }}2(a),(c) and the ((                }}2(a),(c) conditions demonstrate recriticality is not observed for at least 72 hours. Note that these results conservatively assume ((
                                   }}2(a),(c).

(( }}2(a),(c) Figure 1. RCS temperature as a function of core heat level for DHRS cooling. Figure 2 presents the results for the ECCS cooling scenario. Similar to the DHRS cases, the state-point calculations provide lower (conservative) temperatures relative to the transient calculation. During the NRELAP5 convergence process, the lowest power cases experience a temporary decrease in temperature before recovering and stabilizing at a higher temperature, causing the temperature behavior change at about (( }}2(a),(c) in Figure 2. For the sake NuScale Nonproprietary

of conservatism, the coldest temperature is recorded for each case, including the temporary minimum observed. This analysis assumes (( }}2(a),(c), which is conservative given the (( }}2(a),(c) present in at the equilibrium power level, and the very strong (( }}2(a),(c). The equilibrium power level is approximately 1.0% RTP with a corresponding moderator temperature of 170°F. (( }}2(a),(c) Figure 2. RCS temperature as a function of core heat level for ECCS cooling. The (( }}2(a),(c) calculated in the ECCS mode cooling and the conservative treatment of (( }}2(a),(c) results in the ECCS cooldown mode having the highest calculated critical power level. Regardless of coodown mode, CHF limits are NuScale Nonproprietary

not challenged. For confirmation, (( }}2(a),(c) is used to calculate the CHF ratio at equilibrium power levels for each mode. Results are presented in Table 2 for each of the equilibrium scenarios. The equilibrium power levels considered are 0.8 and 1.6 MWth respectively for the, covered riser and ECCS cooldown modes. When considering the SIMULATE5 CBC NRF of (( }} for subtraction, the maximum resultant equilibrium power level, corresponding with the ECCS cooldown mode, increases to 2.9 MWth. An additional (( }}2(a),(c) multiplier is then added to bound possible dynamic power overshoot of the equilibrium condition. The dynamic return to power factor is conservative given the very slow rate of reactivity insertion (( }}2(a),(c) at the time of recriticality. A total local peaking factor (FQ) of (( }}2(a),(c) is conservatively assumed to calculate peak heat flux to account for potential peaking from the stuck out rod. Given the presence of core voiding for the ECCS case, a bounding void fraction of (( }}2(a),(c) is applied for the CHF analysis. This is a bounding value since recriticallity is demonstrated in the nominal SIMULATE5 results to not be possible given this amount of void present in the core at moderator temperatures consistent with those resulting from the 72 hour decay heat levels. The evaluated pressure is the (( }}2(a),(c) at the thermal-hydraulic conditions from the equilibrium power NRELAP5 case. (( 

                                        }}2(a),(c). Results are presented in Table 2 below.

Table 2. CHF results for the bounding critical power level. Parameter DHRS Covered ECCS ECCS w/ SIM5 Riser Uncertainty[1] (( 

                                                                                                                 }}2(a),(c)

CHFR 29 8 4 [1] The pressure and CHF are conservatively copied from the default ECCS results. [2] The specified void fraction is conservatively assumed for the calculation of the CHF. The specified void fraction did not influence the determination of the critical power level. For the equilibrium ECCS cooling results presented in Figure 2, the temperature required for the recriticality without xenon and iodine present is associated with the decay heat levels approximately (( }}2(a),(c) while the transient temperature results are NuScale Nonproprietary

associated with decay heat levels approximately (( }}2(a),(c). However, negative reactivity from xenon is still increasing within this time frame and is sufficient to keep the reactor subcritical. The nominal evaluation of the xenon worth for an RCS temperature of 140°F and a WRSO results in recriticality (keff=1.0) at 48 hours after reactor trip. Results from Figure 2 show that at decay heat levels corresponding with 48 hours after reactor trip, that the RCS temperature is approximately 140°F. The equilibrium power analysis is also performed for the ECCS and DHRS cooldown modes with a 140°F pool temperature. The equilibrium power decreases with increasing pool temperature, and with a 140°F pool, subcriticality is demonstrated at decay heat levels at 72 hours regardless of cooldown mode. The results are summarized in Figure 3. Figure 3 RCS cooling as a function of core heat level with a 140°F pool NuScale Nonproprietary

Conclusions Over 200 NRELAP5 state-point calculations were evaluated to identify the quasi-steady thermal-hydraulic response for a full spectrum of specified initial conditions. The NRELAP5 thermal-hydraulic results were compared to SIMULATE5 results identifying the critical power level for specified core inlet conditions. The equilibrium power levels for different return to power scenarios were identified at conditions where the NRELAP5 results intersect the SIMULATE5 results. Based on the analysis provided in the previous sections the following conclusion are made:

  • The limiting DHRS return to power occurs if RCS level is maintained in the pressurizer as this results in high core flow and lower RCS temperature conditions.
  • The limiting return to power occurs during ECCS cooling.
  • The simple evaluation of MCHFR demonstrates that the return to power is not a CHF concern and is well bounded by full power transients.
  • Nominal xenon decay alone prevents a return to power for about 50 hours after reactor trip.
  • Pool temperature sensitivities show that the return to power is mitigated by pool heat up.

Impact on DCA: FSAR Section 5.4 and Section 15.0.6, including all tables and figures 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 Reactor Coolant System Component and Subsystem Design Return to Power Event RAI 15-6S1 In the event of an extended DHRS cooldown (post-reactor trip), when the RCS is at low boron concentrations and the CVCS is unavailable to add boron, it may be possible for DHRS to cool the core to the point of reestablishing some level of critical neutron power, if it is assumed that the most reactive control rod is stuck out. This event is caused by increased net reactivity post-reactor trip, due to a stuck or partially withdrawn control rod, followed by the DHRS overcooling the RCS. See Section 15.0.6 for further discussion of DHRS performance during a return to power event. RAI 05.04.07-6, RAI 15-6S1 Figure 15.0-8 shows the increase in reactor power due to the reduced negative reactivity. The DHRS mitigates this event by continuing to remove both residual decay heat and fission heat for the duration of the event until a new equilibrium reactor power is reached. As shown in Figure 15.0-11, during this event, the average RCS temperature does not increase above 420 degrees F, safe shutdown temperature, and fuel and thermal hydraulic acceptance criteria are met. See Section 15.0.6 for further discussion of DHRS performance during a return to power event. 5.4.3.4 Tests and Inspections RAI 05.02.01.01-7 Preservice and inservice inspection requirements of Section XI are met for Class 2 components of the BPVC are applicable to the DHRS components including the steam piping, actuation valves, condensers, and condensate piping. RAI 09.03.06-2S1 The DHRS actuation valves are classified as Category B valves in accordance with OM Code Subparagraph ISTC-1300(b) because seat leakage in the closed position is inconsequential for fulfillment of the required function(s). Exercising the actuation valves while at power is not practicable. Therefore, the valves are full-stroke exercised during the equivalent of cold shutdown conditions as allowed by OM Code, Subparagraph ISTC-3521(c). As described in Section 3.9.6, NuScale Mode 3 safe shutdown with reactor coolant temperatures < 200 degrees Fahrenheit is considered to be the equivalent of cold shutdown as defined in the OM Code ISTA-2000. The DHRS actuation valves that are fully cycled as part of a plant shutdown satisfy the exercising requirements provided they meet the observation requirements for testing in accordance with ASME OM Code, Paragraph ISTC-3550. In addition, loss of valve actuator power and position verification testing is performed in accordance with OM Code, Paragraphs ISTC-3560 and ISTC-3700, respectively. The DHRS automatic actuation testing and valve actuation testing, including position verification testing, is performed in accordance with plant technical specifications. Tier 2 5.4-32 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses 15.0.6 Evaluation of a Return to Power RAI 15-6S1 Having all control rods inserted provides the safety-related means to maintain the reactor shut down for internal events and for hazards such as floods and fires in the plant, earthquakes, severe weather conditions, external fires, and external floods. With all control rods inserted, a return to power is precluded. For design basis analysis of internal events for which the worst control rod is assumed stuck out, a return to power is highly unlikely. However, a return to power is evaluated for variousa generic cooldown progressionstransient in order to demonstrate that fuel design limits are not challenged. As described in Section 4.3, a failure in reactivity control system reliability to ensure long term shutdown is calculated to be less than 1E-5 per NPM-reactor year. With the highest worth control rod assembly stuck out and the chemical and volume control system unavailable, subcritical core conditions (keff<1.0) are demonstrated, for 72 hours30 days after a DBE using nominal analysis assumptions, except for the unlikely condition where initial boron concentration isand decay heat are very low. The probability of reactivity control systems failing during the first 72 hours after shutdown within the small window of initial conditions that can lead to a return to power is conservatively calculated to be less than 1E-6 per NPM-reactor year. In the unlikely event of a return to power, shutdown with margin for stuck rods is not required to demonstrate adequate fuel protection. Fuel is protected through physical processes inherent to the NuScale design that control reactivity and limit power compared to a design in which shutdown is required to limit power production to protect fuel integrity. In the NPM design, additional protection is provided by limiting power and passively removing heat. The means for limiting the power produced if the reactor does not remain shut down is dependent on the heat removal system used. RAI 15-6S1 For those events that rely on heat removal using natural circulation flow through the RVVs and RRVs, the heat produced from a return to power with a nominal value for shutdown margin will be limited to less than 100 kW (0.06 percent of rated power) by negative reactivity feedback from moderator density. The low core heat level increases moderator temperature and generates voiding in the core, which in combination with the elevated RCS temperature due to the heatup of the reactor pool from the residual heat of multiple reactors being shut down upon a loss of all AC power, provides negative feedback to keep the core power very low. If decay heat exceeds 100 kW, the reactor will be maintained with keff<1 even with a worst rod stuck out because of negative density reactivity. Therefore, the heat produced after a return to power with the RVVs and RRVs is bounded by maximum decay heat. Consequently, the maximum decay heat curve, with the reactor at keff<1, is used to evaluate maintaining fuel integrity and ECCS performance. Maintaining fuel integrity is addressed in Sections 15.6.5 and 15.6.6 and ECCS performance is addressed in Section 6.3. Tier 2 15.0-40 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6S1, RAI 15-11 15.0.6.1 Identification of Causes and Accident Description Design basis events are analyzed with an assumed highest worth control rod stuck fully withdrawn in order to evaluate the immediate shutdown capability of the negative reactivity insertion due to a reactor trip with the control rods inserting into the core, consistent with GDC 26 (See Section 3.1). In the event of an extended DHRS cooldown, when the RCS is at low boron concentrations and the CVCS is unavailable to add boron, it may be possible for DHRS to cool the core to the point of reestablishing some level of critical neutron power if the most reactive control rod stuck out is assumed. This potential overcooling could cause a unique reactivity event similar to a steam line break for traditional multi-loop PWRs. Therefore, this event is specifically evaluated for specified acceptable fuel design limits (SAFDLs). RAI 15-1, RAI 15-6S1 The purpose of this analysis is to evaluate the thermal hydraulic and core neutronic response of the NPM for an extended DHRS overcooling return to power. This analysis is intended to provide a generic bounding evaluation of the extended DHRS cooling that could result following any DBE, therefore AOO acceptance criteria and conservative analysis assumptions are applied. This event is analyzed specifically for the parameters that generate the most severe overcooling return to power core power overshoot so that it can be concluded that the postulated overcooling return to power following a DBE would not be more limiting than the event presented in this analysis. Additionally this event is analyzed for the postulated transition from DHRS cooling to ECCS cooling in order to demonstrate that transition to ECCS cooling can be safely accomplished.The limiting return to power event occurs when operating conditions are biased to maximize initial core fission product poisons which gradually decay resulting in reactivity insertion. The timing of this reactivity insertion occurs well after equilibrium DHRS or ECCS passive cooling modes will have been established following an initial transient and reactor trip. Therefore, analysis of the return to power is limited to the equilibrium thermal hydraulic and neutronic conditions with appropriate biases and conservatisms to ensure a conservative CHF analysis is performed. RAI 15-1, RAI 15-6S1 15.0.6.2 Sequence of Events and Systems Operation Consistent with the AOO acceptance criteria, the parameters of interest for the return to power event are reactor vessel pressure, secondary pressure and minimum critical heat flux ratio (MCHFR). To appropriately characterize the event, two cases are 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. Tier 2 15.0-41 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6S1 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 decay of xenon slowly adds positive reactivity during the cooldown. 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, RAI 15-6S1 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, RAI 15-6S1 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, RAI 15-6S1 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 modeling methodology described 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.The overcooling return to power analysis is performed using the following analysis procedure: RAI 15-6S1

  • The core average RCS temperature is determined using the long term cooling statepoint analysis approach described in the LTC technical report.

Tier 2 15.0-42 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses

  • The worst rod stuck out, EOC critical power level is determined using the SIMULATE5 core physic analysis model.
  • CHF margin is evaluated using the zero flow CHF correlation described in the LOCA EM topical report.

RAI 15-1, RAI 15-6, RAI 15-6S1 The MCHFR portion of the analysis, where EDSS is unavailable, uses the LOCA NRELAP5 modeling methodology. The CHF correlation applied in the LOCA evaluation model is, evaluated against the 95/95 CHFR acceptance criterion of an AOO, as described in Reference 15.0-3. RAI 15-6S1 SIMULATE5 is an advanced three-dimensional (3D), steady-state, multi-group nodal reactor analysis code capable of multi-dimensional nuclear analyses of reactors. A discussion of SIMULATE5 is provided in Section 4.3. RAI 15-1, RAI 15-6S1 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 equilibrium power results have sufficient conservatism. RAI 15-1, RAI 15-6S1

  • The core is assumed to be at hot full power and end of cycle (5 ppm boron concentration) conditions prior to the transient initiation.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, RAI 15-6S1

  • 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, RAI 15-6S1

  • 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.

Tier 2 15.0-43 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6S1

  • A critical boron concentration (CBC) nuclear reliability factor (NRF) is used in this analysis.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, RAI 15-6S1

  • The ECCS valve capacity is maximized to increase the efficiency of heat transfer from the RPV to the UHS.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, RAI 15-6S1

  • A reactor pool level of 69 feet and a temperature of 4065 degrees Fahrenheit is used leading to a conservatively high cooldown rate, which adds the maximum positive reactivity.

RAI 15-6S1

  • A time-dependent xenon worth is used in this analysis for the purposes of calculating timing of return to power only. The core is assumed to be at EOC conditions at the time of event initiation with equilibrium fission products. The xenon worth specified in the input is determined from the time dependent decay of the fission products that are present in the EOC core.

RAI 15-1 No single failure is assumed. Failure of the main steam or feedwater isolation valves to close could result in a reduction of DHRS cooling, which would be non-conservative for the overcooling return to power event. Full ECCS actuation will be more limiting for CHF, therefore, an ECCS valve failure to open is not considered. RAI 15-1, RAI 15-6S1 For the limiting MCHFR portion of the analysis, a loss of highly reliable DC power (EDSS) is assumed at the time of DHRS initiation, resulting in ECCS actuation. The timing of the ECCS valve opening is near the power peak as determined by a timing sensitivity analysis. Tthe following conservatisms are applied to the MCHFR portion of the analysis: RAI 15-1, RAI 15-6S1

  • A dynamic return to power factor of 2.0 is applied to the equilibrium power level to bound any potential overshoot of the equilibrium power. IAB release timing is sequenced with the timing of the power peak in order to evaluate the most limiting ECCS transition event sequence.

Tier 2 15.0-44 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6S1

  • The maximum radial peaking (FHh) due to the stuck control rod is 67.5. The return to power is driven by the lack of necessary negative reactivity insertion due to the postulated most reactive control rod stuck in a fully withdrawn position. The critical power will be localized in this location generating higher than normalregion of large radial peaking.

RAI 15-1, RAI 15-5, RAI 15-6S1

  • A maximum FQ was chosen with additional penalty for variation in axial peaking.The top shaped axial power distribution is applied consistent with the LOCA EM.

RAI 15-1, RAI 15-6S1

  • A critical boron concentration (CBC) nuclear reliability factor (NRF) is used in the determination of the critical power level for the limiting MCHFR analysis.

The ECCS valve characteristics are conservatively set to maximize the depressurization effect on MCHFR. RAI 15-1, RAI 15-5, RAI 15-6S1

  • Uniform radial density reactivity feedback is used to conservatively bound the localized reactivity suppression due to the localized power generated around the stuck rod location. Flux squared density reactivity feedback weighting is used axially.

RAI 15-1, RAI 15-6S1 15.0.6.3.3 Results This analysis provides a conservative characterization of the equilibrium power and corresponding critical heat flux ratio, should a return to power occur. Additionally, the time of return to power is evaluated based on time-dependent xenon and thermal-hydraulic conditions. RAI 15-6S1 For several different cooldown modes and pool temperature conditions, the nominal equilibrium power level and MCHFR are summarized in Table 15.0-16. The limiting equilibrium power level and MCHFR are provided in Table 15.0-17. The nominal results for the limiting pool temperature are included in Figure 15.0-8. The results for a pool temperature of 140 degrees F are provided in Figure 15.0-10. RAI 15-6S1

  • The maximum equilibrium power level occurs for the ECCS cooldown mode with a 65 degrees F pool.

RAI 15-6S1

  • The maximum equilibrium power is approximately 2.9 MW.

RAI 15-6S1

  • Several of the cases do not return to a critical condition within the 72 hour window analyzed. The earliest return to power occurs at approximately 40 hours post scram.

Tier 2 15.0-45 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-6S1

  • The timing of the initial recriticality demonstrates that the return to power event does not occur during the short-term RCS de-energizing phase, but instead is the result of the slow decay of xenon in the long-term equilibrium phase between decay heat and RCS temperature.

RAI 15-6S1

  • Results show that the equilibrium power decreases with increasing pool temperature.

RAI 15-6S1

  • The MCHFR is well above the analytical limit, therefore it is concluded that the SAFDLs are ensured should a limited return to power occur following an unmitigated cooldown, regardless of initiating event or time in cycle in which it occurs.

RAI 15-6S1 The sequence of events for the DHRS overcooling event is provided in Table 15.0-16. Figure 15.0-8 provides the power response on the return to power. Figure 15.0-10 through Figure 15.0-13 show the transient behavior of key parameters. RAI 15-1, RAI 15-6S1 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, RAI 15-6S1 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, RAI 15-6S1 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. Tier 2 15.0-46 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6S1 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, RAI 15-6S1 For the peak power case, recriticality occurs at approximately 6500 seconds and the peak power (16 MW) occurs at approximately 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 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, RAI 15-6S1

1) Potential core damage is evaluated on the basis that it is precludedacceptable 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, RAI 15-6S1 Fuel integrity is not challenged by an overcooling return to power event. The limiting MCHFR is 1.9 and is shown in Table 15.0-17Figure 15.0-14. The MCHFR isfor evaluated usingcases occurs in the stagnant flow CHFHench-Levy correlation flow range, therefore the 95/95 design limit is 1.3713. The CHF analysis confirms that the DHRS overcooling return to power event does not challengecan 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. Tier 2 15.0-47 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses 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, RAI 15-6S1

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. 15.0.7 References 15.0-1 NuScale Power, LLC, Subchannel Analysis Methodology, TR-0915-17564-P-A, Rev. 21. 15.0-2 NuScale Power, LLC, NuScale Power Critical Heat Flux Correlations, TR-0116-21012-P-A, Rev. 1. 15.0-3 NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422, Rev. 0. 15.0-4 NuScale Power, LLC, Accident Source Term Methodology, TR-0915-17565, Rev. 23. 15.0-5 NuScale Power, LLC, Non-Loss-of-Coolant Accident Transient Analysis Methodology, TR-0516-49416, Rev. 1. 15.0-6 Nuclear Energy Institute, Small Modular Reactor Source Terms, [Position Paper] December 27, 2012, Washington, DC. 15.0-7 NuScale Power, LLC, Long-Term Cooling Methodology, TR-0916-51299, Rev. 01. 15.0-8 U.S. Environmental Protection Agency, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, 1988. Tier 2 15.0-48 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-6, RAI 15-6S1 Table 15.0-16: Return to Power Calculation Nominal Results SummarySequence of Events for Overcooling Return to Power Event EDSS Available(Peak Power Case) Cooldown ModeEvent Pool Temperature Equilibrium Power MCHFR (°F)Time [s]* (%RTP) DHRS Uncovered RiserStart of 065.0 Subcritical N/A Transient DHRS Covered RiserDHRS 65.02.1 0.49 29 Actuation DHRS Covered RiserTime of 65100.0 0.43 33 recriticality DHRS Covered RiserMaximum 790140.0 Subcritical N/A Return to Power (16 MW) ECCS ActuatedEquilibrium 120065.0 1.01 >8 power reached (3 MW) ECCS Actuated 100.0 0.65 13 ECCS Actuated 140.0 Subcritical N/A Note: 

              *Time is rounded.

Tier 2 15.0-75 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 15-1, RAI 15-5, RAI 15-6S1 Table 15.0-17: Return to Power Calculation Limiting ResultsSequence of Events for Overcooling Return to Power Event EDSS unavailable(MCHFR Case) EventEquilibrium Power (%RTP) Time [s]*Peak Heat Flux (kW/m2) MCHFR Time of power peak1.84 7897164 >4 Time of IAB release (ECCS actuation) 7897 Time of minimum critical heat flux ratio 7900 Note: 

              *Time is rounded.

Tier 2 15.0-76 Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-5, RAI 15-6S1 Figure 15.0-8: Nominal Return to Power Results - 65 Degree F Pool Temperature (15.0.6)Power Response on a Return to Power 

                                                                              
                                                                              
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Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-10: Nominal Return to Power Results -140 Degree F Pool Temperature (15.0.6)Return to Power - Net Reactivity (Peak Power Case) 

                                                                                       
                                                                                       
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Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-11: Not UsedReturn to Power - Average Reactor Coolant System Temperature (Peak Power Case) 15.0-92 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-12: Not UsedReturn to Power - Volume Average Fuel Temperature (Peak Power Case) 15.0-93 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-13: Not UsedReturn to Power - Reactor Pressure Vessel Lower Plenum Pressure (Peak Power Case) 15.0-94 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-14: Not UsedReturn to Power ECCS Transition Case - Critical Heat Flux Ratio (MCHFR Case) 15.0-95 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-15: Not UsedReturn to Power ECCS Transition Case - Reactor Power (MCHFR Case) 15.0-96 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-16: Not UsedReturn to Power ECCS Transition Case - RCS Flowrate (MCHFR Case) 15.0-97 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1, RAI 15-5, RAI 15-6S1 Figure 15.0-17: Not UsedReturn to Power ECCS Transition Case - Hot Channel Heat Flux (MCHFR Case) 15.0-98 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1S1, RAI 15-5, RAI 15-6S1 Figure 15.0-18: Not UsedReactor Power EDSS Available - Peak Power Case 15.0-99 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-1S1, RAI 15-5, RAI 15-6S1 Figure 15.0-19: Not UsedRCS Flowrate (Peak Power Case EDSS Available) 15.0-100 Transient and Accident Analyses Draft Revision 3

RAIO-0819-66538 : Affidavit of Thomas A. Bergman, AF-0819-66539 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 Thomas A. Bergman I, Thomas A. Bergman, state as follows:

1. I am the Vice President, 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 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-0819-66539

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 August 2, 2019. AF-0819-66539}}

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