ML18355A973
| ML18355A973 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 12/21/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| Shared Package | |
| ML18355A972 | List: |
| References | |
| RAIO-1218-63931 | |
| Download: ML18355A973 (202) | |
Text
RAIO-1218-63931 December 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.
483 (eRAI No. 9516) on the NuScale Design Certification Application
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 483 (eRAI No. 9516)," dated May 25, 2018
- 2. NuScale Technical Report Long-Term Cooling Methodology, dated January 2017, TR-0916-51299
- 3. NuScale Power, LLC Response to NRC "Request for Additional Information No. 483 (eRAI No. 9516)," dated August 10, 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 Questions from NRC eRAI No. 9516:
15-23 15-24 15-25 15-26 The response to RAI question 15-22 was provided in reference 2. This completes all responses to eRAI 9516. is the proprietary version of the NuScale Response to NRC RAI No. 483 (eRAI No.
9516). 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-1218-63931 If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@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. 9516, proprietary : NuScale Response to NRC Request for Additional Information eRAI No. 9516, nonproprietary : Affidavit of Zackary W. Rad, AF-1218-63932 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. 9516, 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. 9516, 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.: 9516 Date of RAI Issue: 05/25/2018 NRC Question No.: 15-23 10 CFR 50 Appendix A, GDC 34, Residual heat removal, and NuScale's PDC 34, in FSAR Section 3.1.4.5, state, "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded."
The long term cooling technical report (LTC-TR), TR-0916-51299, supports FSAR Section 15.0.5, Long Term Decay and Residual Heat Removal, when the ECCS is used for long term decay heat removal following either a non-LOCA or LOCA event up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The primary acceptance criteria for the analysis are 1) Collapsed liquid level is maintained above the active fuel and 2) fuel cladding temperature is maintained at an acceptable level such that the SAFDLs are preserved.
Section 4.2 of the LTC-TR discusses the validation and assessment of the LTC-EM using two NIST-1 facility tests, namely HP-19a and HP19b. The purpose of test HP-19a was to simulate the spurious opening of an RVV without DHRS activation.
Since the conditions at the time of ECCS actuation for tests HP-19a and HP-19b may be different for a non-LOCA event which transitions from the DHRS to the ECCS for LTC, the applicability of these tests to the LTC of non-LOCA events should be justified.
Justify the applicability of the NIST-1 HP-19a and HP-19b tests to validate the LTC EM used to simulate the non-LOCA transition from DHRS cooling to long term cooling with the ECCS and NuScale Nonproprietary
provide as discussion or reference to the scaling evaluation performed to show the applicability of these tests to the LTC for the non-LOCA events.
NuScale Response:
The primary purpose of the long term cooling evaluation model (EM) is to demonstrate that for all design basis events, the ECCS is appropriately designed to ensure acceptable core cooling is maintained. This is accomplished by ensuring the liquid mass in the reactor pressure vessel (RPV) is sufficient to keep the fuel assembles in the nucleate boiling heat transfer regime. RPV collapsed liquid level and fuel cladding temperatures are identified as the two mechanisms for demonstrating acceptable core cooling, although it is generally accepted that sufficient liquid level also ensures acceptable convective or boiling heat transfer and therefore acceptable fuel cladding temperatures. As such, specific or detailed evaluation of departure from nucleate boiling or onset of critical heat flux is not the focus of the LTC evaluation model. This detailed CHF (DNB) analysis scope is covered in the short term evaluation models for LOCA and non-LOCA analysis where the core energy is much higher than in the long term cooling phase.
Figure 1 provides clarification of the intended scope of evaluation for the LTC EM relative to the event types.
NuScale Nonproprietary
Figure 1 Illustration of the scope and analyses covered by long-term cooling methodology The specific evaluation of ECCS transition is covered in the LOCA LTR where a full break spectrum is analyzed both with and without DHRS, and the conclusion is made that it is conservative for both minimum level and CHF to not credit any DHRS cooling. For the smaller break LOCA event progressions, ((
.2(a),(c) For these smaller breaks it is concluded in the LOCA LTR that not including DHRS cooling is a substantial conservatism, as demonstrated in Figure 2. Therefore, the conservative LOCA event progression including the ECCS actuation transient will be more limiting from the perspective of minimum level and MCHFR than the initial ECCS actuation period in a non-LOCA transient where sustained DHRS cooling is achieved up to the point of IAB release and ECCS actuation. NuScale Nonproprietary
The long-term ECCS cooling for a non-LOCA transient that transitions from DHRS to ECCS cooling is non-limiting compared to the LOCA event progression, as discussed below. ((
}}2(a),(c)
Figure 2 Effect of DHRS operation on RCS injection line break RPV collapsed level transients (Left: without DHRS, Right: with DHRS) While detailed analysis of the ECCS transition transient due to either inadvertent actuation block (IAB) release or due to 24 hour actuation by the module protection system (MPS) is not performed as part of the LTC scope, it is simulated by the LTC EM to ensure pressure, temperature, decay heat levels, and inventory boundary conditions are reasonably predicted upon the onset of the LTC analysis. Of these parameters, decay heat and liquid inventory are of primary importance as described in the response to eRAI-9523 question 15-14, submitted in NuScale letter RAIO-1218-63922 dated December 21, 2018. Therefore, the decrease in RCS inventory in non-LOCA events are of primary focus to minimize the available liquid inventory for the purposes of demonstrating ECCS functionality. Sensitivity analysis performed within the LTC scope shows that the limiting LOCA injection line break event is bounding of the most severe non-LOCA decrease in RCS inventory events. Additionally, Figures 3 through 6 demonstrate that both LOCA and non-LOCA events are similar from a thermal hydraulics standpoint once ECCS cooling is established. NuScale Nonproprietary
Figure 3 - Riser collapsed level (SGTF) Figure 4 - Riser collapsed level (injection line break) NuScale Nonproprietary
Figure 5 - Peak cladding temperature (SGTF) Figure 6 - Peak clad temperature (injection line breaks) NuScale Nonproprietary
The intent of the HP-19a and 19b is to focus on the extension of the LOCA EM assessment and NRELAP5 validation of the break initiating event through the actuation of ECCS into the extended long term cooling behavior, with particular interest in the effect of non-condensables in the CNV. As described in section 4.2 of the LTC TR, ((
.}}2(a),(c) This conclusion is additionally justified by the discussion provided in this response.
In summary, the need for specific validation of the integral system response of the DHRS cooling to ECCS cooling transient is not required because this is a non-limiting transient from a short term MCHFR or minimum level perspective and is sufficiently similar to the LOCA or IORV transient once ECCS cooling has been established. 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.: 9516 Date of RAI Issue: 05/25/2018 NRC Question No.: 15-24 10 CFR 50 Appendix A, GDC 34, Residual heat removal, and NuScale's PDC 34, in FSAR Section 3.1.4.5, state, "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded." The long term cooling technical report (LTC-TR), TR-0916-51299, supports FSAR Section 15.0.5, Long Term Decay and Residual Heat Removal, when the ECCS is used for long term decay heat removal following either a non-LOCA or LOCA event up to 72 hours. The primary acceptance criteria for the analysis are 1) Collapsed liquid level is maintained above the active fuel and 2) fuel cladding temperature is maintained at an acceptable level such that the SAFDLs are preserved. Section 4.2.3.3 describes the NRELAP5 simulation of the HP-19a test. ((
}}2(a),(c) LTC-TR Section 4.2.3.3 stated the following:
NuScale Nonproprietary
((
}}2(a),(c)
The explanation for the ((
}}2(a),(c) The LTC-TR goes on to state:
((
}}2(a),(c)
The test simulation shows that NRELAP5 is ((
}}2(a),(c) However, the heat removal by the containment has a direct effect on the pressure and level within the containment which directly affect the level and pressure within the RPV. If the heat removal to the pool is not correctly calculated, the heat remove from the NuScale Nonproprietary
containment cannot be calculated correctly. If the containment level and pressure are well predicted some compensating error in the heat transfer either from the containment to the CNV or from the CNV to the pool is likely responsible. The applicant did not described an evaluation of the heat transfer coefficients to provide an explanation of the simulation results. While it is understood that the cooling pool temperatures may be set to bounding temperature and level values to assess which condition may be worse for an aspect of LTC, it does not assess the integrated long term effect of miscalculated heat transfer, nor quantify the uncertainty associated with the miscalculation of the heat transfer or explain potentially compensating error lead to accurate predictions of some results. Provide additional information as to the basis for the prediction of the reactor vessel water level in relation to the containment water level. Explain the physical basis for the good predictions of RPV and containment vessel pressures, while the observed thermal stratification and level in the cooling pool is not accurately predicted. Identify any areas of uncertainty which could lead to an accurate prediction of reactor vessel and containment pressures while other key phenomena are not predicted well. NuScale Response: First, additional information is provided regarding the reactor pressure vessel (RPV) and containment vessel (CNV) water levels. Then the NRELAP5 prediction of thermal stratification in the cooling pool is clarified, and additional information is provided about the cooling pool level to justify the physical basis for good prediction of the RPV and CNV pressures considering the prediction of thermal stratification and liquid level in the cooling pool. Level Calculation Background ((
}}2(a),(c)
NuScale Nonproprietary
((
}}2(a),(c)
NuScale Nonproprietary
- ((
}}2(a),(c)
NRELAP5 Prediction of Reactor Pressure Vessel and Containment Vessel Level Figure 1a, Figure 1b, Figure 2a, and Figure 2b show ((
}}2(a),(b),(c) Overall the NRELAP5 prediction in the mid-term and long-term of the RPV and CNV level is in reasonable or excellent agreement with the data.
NuScale Nonproprietary
((
}}2(a),(b),(c), ECI Figure 1a. HP19a mid-term transient reactor pressure vessel level comparison
((
}}2(a),(b),(c), ECI Figure 1b. HP19a long-term cooling reactor pressure vessel level comparison NuScale Nonproprietary
((
}}2(a),(b),(c), ECI Figure 2a. HP19a mid-term transient containment vessel level comparison
((
}}2(a),(b),(c), ECI Figure 2b. HP19a long-term containment vessel level comparison NuScale Nonproprietary
((
}}2(a),(b),(c), ECI Figure 3. HP19a long-term cooling reactor pressure vessel and containment vessel level comparison Prediction of Cooling Pool Temperatures Figure 4-8 in the long-term cooling (LTC) technical report is updated in the change package included in this response. The updated Figure 4-8 provides additional information regarding the measured and predicted fluid temperatures at the top of the cooling pool. This figure is included in this response as Figure 4 for convenience. ((
}}2(a),(c)
NuScale Nonproprietary
((
.}}2(a),(b),(c), ECI With this additional information and clarification, the reasonable predictions of NRELAP5 RPV and CNV pressure are consistent with NRELAP5 predictions of RPV level, CNV level, cooling pool level, and cooling pool temperature.
((
}}2(a),(b),(c), ECI Figure 4. HP19a transient long-term cooling upper cooling pool temperature comparison NuScale Nonproprietary
((
}}2(a),(b),(c), ECI Figure 5. HP19a longterm cooling average cooling pool temperature comparison Impact on DCA:
Technical Report TR-0916-51299, Long-Term Cooling Methodology, been revised as described in the response above and as shown in the markup provided with the response to question 15-26. NuScale Nonproprietary
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9516 Date of RAI Issue: 05/25/2018 NRC Question No.: 15-25 10 CFR 50 Appendix A, GDC 34, Residual heat removal, and NuScale's PDC 34, in FSAR Section 3.1.4.5, state, "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded." The long term cooling technical report (LTC-TR), TR-0916-51299, supports FSAR Section 15.0.5, Long Term Decay and Residual Heat Removal, when the ECCS is used for long term decay heat removal following either a non-LOCA or LOCA event up to 72 hours. The primary acceptance criteria for the analysis are 1) Collapsed liquid level is maintained above the active fuel and 2) fuel cladding temperature is maintained at an acceptable level such that the SAFDLs are preserved. LTC-TR Section 4.2.4 discusses the prediction of NIST-1 test HP-19b. The only difference between tests HP-19a and HP19b was the containment pressure initial condition. The initial conditions at 6,000 seconds were different from those in HP-19a, indicating that the initial containment backpressure at atmospheric conditions had some effect on the transient progression from 0 - 6,000 seconds. The LTC TR did not discuss how the effect of initial containment atmosphere condition had on the test response or how it would affect the predicted level response. The staff notes that Figure 4-10 shows the predicted RPV water level is ((
}}2(a),(c)
NuScale Nonproprietary
((
}}2(a),(c) It is expected that good agreement over the long-term in the RPV water level would also result in good agreement in containment water level, unless, the heat removal by the containment cooling pool is under predicted by NRELAP5.
As the expected agreement between the RPV and containment level is not demonstrated the staff is seeking to understand if model limitations exists which could require code modification or additional code validation. In addition, the prediction of the cooling pool temperature in the upper region in Figure 4-16, ((
}}2(a),(c)
The staff is requesting the applicant describe how: 1) the containment atmospheric initial condition affected the test transient response in comparison to the HP-19a test, 2) why the containment level (Figure 4-9) is ((
}}2(a),(c) and 3) and the change in the NRELAP5 cooling pool simulation of the upper temperature prediction at approximately 50,000 seconds in Figure 4-16.
NuScale Response: The description of the differences between the HP-19a and HP-19b test runs in the long-term cooling (LTC) technical report Section 4.2.1 was clarified as indicated at the end of this RAI response. ((
}}2(a),(b),(c),ECI NuScale Nonproprietary
((
}}2(a),(b),(c), ECI These differences between the HP-19a and HP-19b initial conditions contribute to slightly different initial system mass and slightly different test conditions at 6,000 sec after transient initiation. Overall, the NRELAP5 prediction of the results is consistent between HP-19a and HP-19b.
The discussion provided in response to RAI 9516, question 15-24, submitted concurrently with this response, regarding the calculation of the compensated liquid values in the HP-19a test is also applicable to the assessment of the HP-19b test. Therefore, the same revisions to text and figures that apply to the HP-19a test in the LTC technical report apply to the HP-19b test, and the technical report was revised as indicated with the response to question 15-26. The NRELAP5-calculated cooling pool fluid temperature inflection around 50,000 sec ((
}}2(a),(b),(c), ECI NuScale Nonproprietary
((
}}2(a),(b),(c), ECI Figure 1. HP-19b transient long-term cooling upper pool temperature comparison
((
}}2(a),(c)
Figure 2. HP-19b cooling pool and heat transfer plate temperature at approximately 50,000 seconds, at 228 inches from bottom of cooling pool NuScale Nonproprietary
((
}}2(a),(c)
Figure 3. HP-19b upper cooling pool axial temperature profile at 55,000 seconds Impact on DCA: Technical Report TR-0916-51299, Long-Term Cooling Methodology, has been revised as described in the response above and as shown in the markup provided with the response to question 15-26. NuScale Nonproprietary
Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9516 Date of RAI Issue: 05/25/2018 NRC Question No.: 15-26 10 CFR 50 Appendix A, GDC 34, Residual heat removal, and NuScale's PDC 34, in FSAR Section 3.1.4.5, state, "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded." The long term cooling technical report (LTC-TR), TR-0916-51299, supports FSAR Section 15.0.5, Long Term Decay and Residual Heat Removal, when the ECCS is used for long term decay heat removal following either a non-LOCA or LOCA event up to 72 hours. The primary acceptance criteria for the analysis are 1) Collapsed liquid level is maintained above the active fuel and 2) fuel cladding temperature is maintained at an acceptable level such that the SAFDLs are preserved. RG 1.203 describes the EMDAP, and provides guidance, which the NRC staff considers acceptable for use in developing and accessing EMs used to analyze transient and accident behavior. RG 1.203 provides guidance regarding the content of assessment reports: "With respect to a calculational device input model and the related sensitivity studies, assessment reports must achieve the following additional purposes: (10) Provide a nodalization diagram, along with a discussion of the nodalization rationale. (11) Specify and discuss the boundary and initial conditions, as well as the operational conditions for the calculations. NuScale Nonproprietary
(13) Discuss modifications to the input model (nodalization, boundary, initial or operational conditions) resulting from sensitivity studies." In LTC-TR Section 4.3, the Applicant describes the studies performed to assess the difference in the NRELAP5 nodalization for the LTC-EM and the NRELAP5 nodalization for the LOCA-EM, and stated that both nodalization schemes preserve the code validation pedigree. However, the LTC-TR did not explain what was different. LTC-TR Section 4.3 stated: ((
}}2(a),(c)
In addition, the changes described in Section 4.1 of the LTC TR were included in the LOCA EM nodalization calculation. The background, bases, and justification for these changed conditions is not discussed in the LTC-TR. In Figure 4-18 through Figure 4-23 of The TR, the applicant describes the differences in the results between the calculations. However, there is no explanation relating the differences in the nodalization to the differences in the results. The applicant concluded that the ((
}}2(a),(c) Without a firm understanding of the reasons for the differences in the calculations, it is not clear how it may be concluded that the differences in the pressures and temperatures would not affect the flows through the RVVs, RRVs, flow within the RPV, the natural circulation of coolant flow within the CNV and the heat transfer response during the long- term cooling calculation.
Discuss the analysis differences between the LTC-EM nodalization of the letdown line break and the LOCA-EM nodalization of the letdown line break, including the basis and justification for the changes to the LOCA-EM model for the conditions noted. Describe how the differences between the model nodalization resulted in the differences between the calculated results. Alternately, the applicant may provide a detailed technical basis justifying the LTC EM NuScale Nonproprietary
nodalization based on adequately capturing the uncertainity assoicated with the high ranked phenomena identified by the PIRT. NuScale Response: The long term cooling (LTC) technical report (TR-0916-51299) was revised to address the differences between the NPM modeling used in the LOCA LTR (TR-0516-49422) and that used for the LTC analysis as described at the end of this response. As described in the revision to Section 4.0, the updated LTC EM is developed from the coarser LOCA model described in Section 9.6.1 of TR-0516-49422. The coarser LOCA model is selected to improve calculation performance over the long term, quasi-steady state conditions where the fidelity of a finer model nodalization is not required. As described in the revision to Section 4.1, additional changes are applied to the coarser LOCA model to further improve calculation performance during LTC analysis: reduction from a three-channel core region to a single-channel average core, nodalization simplifications to the secondary system model, and modification of the injection line break model to a direction connection between the RPV riser and containment at the same elevation as the RRVs. As described in the revision to Section 4.3, the discharge line break under minimum level conditions is evaluated using the LOCA base model, the LOCA coarser model, and the updated LTC model. Results presented in Section 4.3 generally show similar LTC response between the three models, with minor differences being on the same order of magnitude as seen during LOCA model nodalization sensitivity cases presented in Section 9.6.1 of TR-0516-49422. Additional key parameters are presented in Table 1 below for the three models. Final module conditions are presented 12.5 hours after initiation. NuScale Nonproprietary
Table 1. Model nodalization sensitivity for 100% discharge line break Min. Time of Final Max. Final Collapsed Time of Final ECCS Clad T. Pressurizer Case Description Riser Level Min. Level Core Inlet Actuation (Avg. Core) Pressure above TAF (hour) T. (°F) (hour) (°F) (psia) (ft) base LOCA model 0.4 4.73 3.88 159.0 175.8 3.2 coarser LOCA model 0.3 4.89 3.91 159.4 179.3 3.2 LTC model 0.3 4.91 3.95 167.3 181.5 3.4 The primary difference between the LTC model and the evaluated LOCA models is in long term core inlet temperature, which is shown to be higher in the updated LTC model (see updated Figure 4-25 in TR-0916-51299). The temperature difference results from the single-channel versus three-channel core model. The three-channel LOCA EM core model exhibits non-physical flow oscillation between the bypass and the fuel channels long term. This flow oscillation promotes additional cooling, resulting in a lower core inlet temperature than what would be seen if the flow circulated all the way up the riser. Thus the combined channel LTC model is believed to be more physically realistic for long term predictions. Another difference is seen specifically for minimum collapsed level during an injection line break, where minimum level is lower for the LTC model relative to the LOCA models (see updated Figure 4-27 in TR-0916-51299). The LTC model conservatively models the injection line break at a lower elevation and as a direct connection between the riser and containment. This promotes additional liquid loss from the riser to containment during the characteristic level depression seen during LTC, which is conservative for evaluating collapsed level above the top of active fuel (TAF). The injection line break is the limiting event for collapsed level in LTC analysis. Additional changes were made to TR-0916-51299 as indicated at the end of this response which address the responses to RAI 9470, question 15.06.05-10 (NuScale letter RAIO-1218-63933 dated December 21, 2018); RAI 9479, question 15.06.05-5 (NuScale letter RAIO-1218-63936, dated December 21, 2018); RAI 9523, question 15-14 (NuScale letter RAIO-1218-63922, dated December 21, 2018), RAI 9522, question 15-13 (NuScale letter RAIO-1218-63938, dated December 21, 2018); and RAI 9516, questions 15-23, 15-24, and 15-25, submitted concurrently with this response. In addition, technical specification 3.5.3 is revised as NuScale Nonproprietary
indicated at the end of this response to support the assumption of ultimate heat sink temperature. Impact on DCA: Technical Report TR-0916-51299, Long-Term Cooling Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary
Ultimate Heat Sink 3.5.3 3.5 PASSIVE CORE COOLING SYSTEMS (PCCS) 3.5.3 Ultimate Heat Sink LCO 3.5.3 Ultimate Heat Sink shall be maintained within the limits specified below:
- a. Level 68 ft,
- b. Bulk average temperature 65 ºF and 140110 ºF, and
- c. Bulk average boron concentration shall be maintained within the limit specified in the COLR.
APPLICABILITY: At all times. ACTIONS
NOTE-----------------------------------------------------------
LCO 3.0.3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. Ultimate Heat Sink Level A.1 Suspend module Immediately
< 68 ft and > 55 ft. movements.
AND A.2 Suspend movement of Immediately irradiated fuel assemblies in the refueling area. AND A.3 Restore Ultimate Heat Sink 30 Days Level to within limits. NuScale 3.5.3-1 Draft Revision 3.0
Ultimate Heat Sink B 3.5.3 BASES BACKGROUND (continued) will be retained to limit offsite doses from the accident to within the values reported in FSAR Chapter 15 (Ref. 12). During transients and shutdowns which are not associated with design basis events in which DHRS or ECCS is actuated, water from the RP is added to the containment vessel by the Containment Flood and Drain System (CFDS). After reaching an appropriate level in the containment, the reactor vent valves (RVVs) and reactor recirculation valves (RRVs) are opened to permit improved heat transfer from the reactor coolant system (RCS) to the containment vessel walls. During normal operations, the RP limits temperatures of the module by maintaining the containment vessel partially submerged in water. The water also provides shielding above and around the region of the core during reactor operations, limiting exposure to personnel and equipment in the area. In MODE 4, the module is transported from the operating position to the RFP area of the UHS. The UHS provides buoyancy as the module displaces pool water during the movement, thereby reducing the load on the reactor building crane. APPLICABLE During all MODES of operation and storage of irradiated fuel, the UHS SAFETY supports multiple safety functions. ANALYSIS The UHS level is assumed and credited in a number of transient analyses. The 68 ft level provides buoyancy assumed in the reactor building crane analysis and design to ensure its single-failure proof capacity during module movement in MODE 4. A UHS level of 5565 ft provides margin above the minimum level required to support DHRS and ECCS operation in response to LOCA and non-LOCA design basis events. The 65 ft level also assures the containment vessel wall temperature initial condition assumed in the peak containment pressure analysis. The UHS bulk average temperature is assumed and credited, directly or indirectly in design basis accidents including those that require DHRS and ECCS operation such as LOCA and non-LOCA design basis events. The bulk average temperature is also assumed as an initial condition of the peak containment pressure analysis, and the minimum pool temperature is an assumption used in long-term cooling analyses. Note that the UHS sensible heat needed to heat the pool to boiling is not credited in the NuScale B 3.5.3-2 Draft Revision 3.0
Ultimate Heat Sink B 3.5.3 BASES APPLICABLE SAFETY ANALYSIS (continued) UHS safety analyses for pool inventory. Additionally, the UHS bulk average temperature is assumed in the buoyancy calculation of the reactor building crane load during movement of the module. The UHS bulk average boron concentration lower limit is established to ensure adequate shutdown margin during unit shut downs that are not associated with events resulting in DHRS or ECCS actuation, when the module is filled with RP inventory using the CFDS and the RRVs are opened. It also ensures adequate shutdown margin when the module is configured with the UHS inventory in contact with the reactor core, specifically in MODE 4 when the containment vessel is disassembled for removal, and in MODE 5. The upper limit on boron concentration is established to limit the effect of moderator temperature coefficient (MTC) during localized or UHS bulk average temperature changes while the module and core are in contact with UHS water. The upper limit also provides assurance for criticality and boron dilution analyses. The ultimate heat sink level, temperature, and boron concentration parameters satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). LCO The UHS must provide an adequate heat sink to perform its UHS function. This is accomplished by providing a sufficient submersion of the module and the mass of water that can be heated, and vaporized to steam if necessary, to remove decay heat via the decay heat removal system or conduction through the containment vessel walls and heat from irradiated fuel in the pool. The UHS level limits ensure that this level of module submersion and mass of water is available. The UHS bulk average temperature is an initial assumption of safety analyses. The limits on temperature preserves the analysies assumptions and permits crediting the pool to mitigate these events. ItThey also provides margin for performance of the UHS function in that the pool must be heated before vaporization of the contents will begin. Determination of the UHS bulk average temperature is in accordance with approved procedures. The boron concentration must be within limits when the UHS contents are in communication with the RCS to preserve core reactivity assumptions and analyses. Determination of the bulk average boron concentration is in accordance with approved plant procedures. NuScale B 3.5.3-3 Draft Revision 3.0
Ultimate Heat Sink B 3.5.3 BASES ACTIONS (continued) C.1, C.2, and C.3 If the UHS bulk average temperature is < 65 °F or > 140110 °F, actions must be taken to restore the UHS bulk average temperature to within the limits. 140110 °F is the initial temperature assumed in the UHS boiling analysispeak containment pressure analysis calculations, and is consistent withconservative with respect to the RB Crane lifting capacity calculation. Additionally, tThe minimum UHS bulk average temperature is an assumption used in long-term cooling analyses. The SFPC system in conjunction with the RFP cooling system is designed to maintain a UHS bulk average temperature of 140110 °F. D.1 and D.2 If the UHS level or bulk average temperature cannot be returned to within limits within the associated Completion Time, the unit must be brought to a condition where the decay heat of the unit with the potential to be rejected to the UHS is minimized. To achieve this status, the unit must be brought to MODE 2 within 6 hours and MODE 3 within 36 hours. The allowed Completion Times are reasonable, based on operating requirements, to reach the required unit conditions from full power conditions in an orderly manner. E.1, E.2, E.3, E.4, and E.5 If the UHS bulk average boron concentration is not within limits, actions must be initiated and continued to restore the concentration immediately. Additionally, activities that could place pool inventory in communication with the reactor core must be suspended. Therefore, CFDS flow into the containment must be immediately terminated, and disassembly of the containment vessel that would open the RCS to communication with the UHS also suspended. Additionally, module movement must be suspended and the movement of irradiated fuel suspended. The suspension of module and/or fuel movement shall not preclude completion of movement to safe position. NuScale B 3.5.3-5 Draft Revision 3.0
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Long-Term Cooling Methodology Draft Revision 10 Docket: PROJ0769 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com © Copyright 20187 by NuScale Power, LLC © Copyright 20187 by NuScale Power, LLC i
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC. The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary. © Copyright 20187 by NuScale Power, LLC ii
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008820. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. © Copyright 20187 by NuScale Power, LLC iii
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 CONTENTS Abstract ....................................................................................................................................... 1 Executive Summary .................................................................................................................... 2 1.0 Introduction ..................................................................................................................... 5 1.1 Purpose ................................................................................................................. 5 1.2 Scope .................................................................................................................... 5 1.3 Abbreviations and Definitions ................................................................................ 8 2.0 Regulatory Requirements and Roadmap .................................................................... 10 2.1 Background ......................................................................................................... 10 2.2 Regulatory Requirements and Guidance ............................................................ 10 2.2.1 Regulatory Requirements .................................................................................... 10 2.2.2 Regulatory Guidance ........................................................................................... 11 2.3 Acceptance Criteria and Transient Duration ........................................................ 13 2.3.1 Acceptance Criteria ............................................................................................. 13 2.3.2 Transient Duration ............................................................................................... 14 2.4 Long-Term Cooling Evaluation Model Roadmap ................................................. 15 3.0 Phenomena Identification and Ranking Table ............................................................ 20 3.1 Phenomena Identification and Ranking Table Process ....................................... 20 3.2 Figures of Merit ................................................................................................... 20 3.3 Highly Ranked Phenomena ................................................................................. 20 3.3.1 ((
}}2(a),(c) ................................................................................................................... 21 3.3.2 ((
}}2(a),(c) ................................................................................................................... 22 3.3.3 (( }}2(a),(c) .............................. 22 3.3.4 (( }} 2(a),(c)................................ 23 3.3.5 (( }}2(a),(c) .................. 23 3.3.6 (( }}2(a),(c) ............ 24 3.3.7 (( }}2(a),(c)............... 24 3.3.8 (( }}2(a),(c) ....................................................................................... 25 3.3.9 (( }}2(a),(c) .......................................................... 25 3.3.10 (( }}2(a),(c) ..................................... 26 3.3.11 (( }}2(a),(c) ....................................................................... 27 3.3.12 (( }}2(a),(c)................................................................... 27
© Copyright 20187 by NuScale Power, LLC iv
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.13 (( }}2(a),(c)........................................ 28 3.3.14 (( }}2(a),(c) ................................................................................. 28 3.3.15 (( }}2(a),(c) .............................................. 29 3.3.16 (( }}2(a),(c)......................................................................... 29 3.3.17 (( }}2(a),(c) ......................... 30 3.3.18 (( }}2(a),(c)......................................................... 30 3.3.19 (( }}2(a),(c)....................................................... 31 3.3.20 (( }}2(a),(c) .......................................................................... 31 3.3.21 (( }}2(a),(c).................................................................... 32 3.3.22 ((
}}2(a),(c) ............................................................ 33 3.3.23 (( }}2(a),(c) ..................................... 34 3.3.24 (( }} 2(a),(c) ..................................... 34 3.3.25 (( }}2(a),(c) ........................................ 35 3.3.26 (( }}2(a),(c) .............................................. 36 3.3.27 (( }}2(a),(c) ............................................................ 37 3.3.28 (( }}2(a),(c) ........................................ 38 3.3.29 (( }}2(a),(c) ...................................................................... 39 3.3.30 (( }}2(a),(c)........................................ 40 3.3.31 (( }}2(a),(c) ................................................................. 41 3.3.32 (( }}2(a),(c)....................................................... 41 3.3.33 ((
}}2(a),(c) ................................................................................................................... 42 3.3.34 (( }}2(a),(c) ................................................... 43 3.3.35 (( }}2(a),(c)............................................................ 44 3.3.36 ((
}}2(a),(c) ................................................................................................................... 45 3.3.37 (( }} 2(a),(c) .......... 45 3.3.38 (( }}2(a),(c) .............................. 46 3.3.39 ((
}}2(a),(c)........................................................................................... 47 3.3.40 (( }}2(a),(c) .............. 48 3.3.41 ((
}} 2(a),(c) ....................................................................................................... 49 3.3.42 (( }} 2(a),(c) ........................................................................... 50
© Copyright 20187 by NuScale Power, LLC v
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 4.0 NRELAP5 Applicability to Long-Term Cooling Analysis ........................................... 51 4.1 Summary of the Long-Term Cooling Model ......................................................... 51 4.2 NRELAP5 Validation and Assessments for Long-Term Cooling .......................... 53 4.2.1 Long-Term Cooling Tests at the NIST-1 Facility .................................................. 53 4.2.2 NIST-1 Facility NRELAP5 Model ......................................................................... 54 4.2.3 Integral Assessment of NIST-1 HP-19a ............................................................... 54 4.2.4 Integral Assessment of NIST HP-19b .................................................................. 64 4.2.5 Conclusions from Integral Test Assessments ...................................................... 72 4.3 Loss-of-Coolant Accident / Long-Term Cooling Consistency Evaluation ............. 72 5.0 Long-Term Cooling Methodology and Evaluation ..................................................... 86 5.1 Long-Term Cooling Heat Removal Methodology ................................................. 86 5.2 Events Evaluated for Long Term Cooling ............................................................ 88 5.3 Long Term Cooling Analysis Assumptions ........................................................... 89 5.3.1 Electric Power Availability .................................................................................... 89 5.3.2 Single Failure Evaluation ..................................................................................... 89 5.3.3 Multi-module Consideration ................................................................................. 90 5.3.4 Long Term Cooling Evaluation Period ................................................................. 90 5.4 Initial Conditions and Biases ............................................................................... 90 5.5 Sensitivity Considerations ................................................................................... 91 5.6 Demonstration of Limiting Results ....................................................................... 92 5.6.1 Maximum Temperature ........................................................................................ 96 5.6.2 Minimum Temperature ....................................................................................... 107 5.6.3 Minimum Level .................................................................................................. 118 5.6.4 Steam Generator Tube Failure with Minimum Level Conditions ....................... 128 5.6.5 State-point Evaluation at 72 Hours .................................................................... 137 5.6.6 Summary and Conclusions ............................................................................... 138 6.0 Boron Precipitation Methodology and Analysis Results ........................................ 147 6.1 General Approach and Acceptance Criteria ...................................................... 147 6.2 Methodology ...................................................................................................... 148 6.2.1 Calculate Total Boron Mass ............................................................................... 148 6.2.2 Calculate Mass of Fluid in Mixing Volume ......................................................... 149 6.2.3 Calculate Boron Concentration in Mixing Volume ............................................. 149 6.2.4 Assess Margin to Boron Precipitation ............................................................... 150 © Copyright 20187 by NuScale Power, LLC vi
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 6.3 Results .............................................................................................................. 150 6.4 Conclusions ....................................................................................................... 152 7.0 Summary and Conclusions ........................................................................................ 153 8.0 References ................................................................................................................... 154 8.1 Source Documents ............................................................................................ 154 8.2 Referenced Documents ..................................................................................... 154 © Copyright 20187 by NuScale Power, LLC vii
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 TABLES Table 1-1 Abbreviations ......................................................................................................... 8 Table 1-2 Definitions .............................................................................................................. 9 Table 2-1 Evaluation model development and assessment process steps and associated application in the long-term cooling evaluation model ......................................... 17 Table 5-1 Default scenario initial conditions and biases ...................................................... 90 Table 5-2 Results of state-point analysis at 72 hours ........................................................ 137 FIGURES Figure 1-1 Illustration of the scope and analyses covered by long-term cooling methodology .......................................................................................................... 7 Figure 2-1 Evaluation model development and assessment process................................... 16 Figure 4-1 HP19a transient long-term cooling containment vessel level comparison .......... 56 Figure 4-2 HP19a transient long-term cooling reactor pressure vessel level comparison .... 57 Figure 4-3 HP19a transient long-term cooling containment vessel pressure comparison .... 58 Figure 4-4 HP19a transient long-term cooling reactor pressure vessel pressure comparison .......................................................................................................... 59 Figure 4-5 HP19a transient long-term cooling pool level comparison .................................. 60 Figure 4-6 HP19a transient long-term cooling lower pool temperature ................................ 61 Figure 4-7 HP19a transient long-term cooling pool middle temperature comparison ........... 62 Figure 4-8 HP19a transient long-term cooling pool upper temperature comparison ............ 63 Figure 4-9 HP19b Transient long-term cooling containment vessel level comparison ......... 66 Figure 4-10 HP19b transient long-term cooling reactor pressure vessel level comparison .... 66 Figure 4-11 HP19b transient long-term cooling containment vessel pressure comparison .... 67 Figure 4-12 HP19b transient long-term cooling reactor pressure vessel pressure comparison .......................................................................................................... 68 Figure 4-13 HP19b transient long-term cooling pool level comparison .................................. 69 Figure 4-14 HP19b transient long-term cooling pool lower temperature comparison ............. 70 Figure 4-15 HP19b transient long-term cooling pool middle temperature comparison ........... 71 Figure 4-16 HP19b transient long-term cooling pool upper temperature comparison ............ 72 Figure 4-17 Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer pressure through 1 hour ................................................................... 75 Figure 4-18 Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer pressure ............................................................................................ 76 Figure 4-19 Loss-of-coolant accident evaluation model nodalization consistency comparison: continment pressure through 1 hour ................................................................... 77 Figure 4-20 Loss-of-coolant accident evaluation model nodalization consistency comparison: containment pressure .......................................................................................... 78 Figure 4-21 Loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuel through 1 hour ......... 79 Figure 4-22 Loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuel .................................. 80 Figure 4-23 Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer level .................................................................................................. 81 Figure 4-24 Loss-of-coolant accident evaluation model nodalization consistency comparison: containment collapsed liquid level relative to the top of active fuel ..................... 82 Figure 4-25 Loss-of-coolant accident evaluation model nodalization consistency comparison: core inlet temperature ......................................................................................... 83 © Copyright 20187 by NuScale Power, LLC viii
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 4-26 Loss-of-coolant accident evaluation model nodalization consistency comparison: core outlet temperature ....................................................................................... 84 Figure 4-27 Loss-of-coolant accident evaluation model nodalization consistency comparison - injection line break: riser collapsed liquid level relative to the top of active fuel .. 85 Figure 5-1 Maximum temperature injection line break: core inlet temperature .................... 98 Figure 5-2 Maximum temperature injection line break: maximum clading temperature ...... 99 Figure 5-3 Maximum temperature injection line break: riser collapsed liquid level above top of active fuel ...................................................................................................... 100 Figure 5-4 Maximum temperature injection line break: containment liquid level above bottom ............................................................................................................... 101 Figure 5-5 Maximum temperature injection line break: RCS pressure at RVV ................... 102 Figure 5-6 Maximum temperature injection line break: RCS pressure at RVV after 4 hours ................................................................................................................. 103 Figure 5-7 Maximum temperature injection line break: containment pressure at RVV ...... 104 Figure 5-8 Maximum temperature injection line break: containment pressure at RVV after 4 hours .............................................................................................................. 105 Figure 5-9 Maximum temperature injection line break: RVV2 flow ..................................... 106 Figure 5-10 Minimum temperature injection line break: core inlet temperature ................... 108 Figure 5-11 Minimum temperature injection line break: maximum clading temperature ..... 109 Figure 5-12 Minimum temperature injection line break: riser collapsed liquid level above top of active fuel ................................................................................................ 110 Figure 5-13 Minimum temperature injection line break: containment liquid level above bottom ............................................................................................................... 111 Figure 5-14 Minimum temperature injection line break: RCS pressure at RVV ................... 112 Figure 5-15 Minimum temperature injection line break: RCS pressure at RVV after 4 hours ................................................................................................................. 114 Figure 5-16 Minimum temperature injection line break: containment pressure at RVV ....... 115 Figure 5-17 Minimum temperature injection line break: containment pressure at RVV after 4 hours ................................................................................................................. 116 Figure 5-18 Minimum temperature injection line break: RVV2 flow ...................................... 117 Figure 5-19 Minimum level injection line break: core inlet temperature ............................... 119 Figure 5-20 Minimum level injection line break: maximum clading level ............................. 120 Figure 5-21 Minimum level injection line break: riser collapsed liquid level above top of active fuel ..................................................................................................................... 121 Figure 5-22 Minimum level injection line break: containment liquid level above bottom...... 122 Figure 5-23 Minimum level injection line break: RCS pressure at RVV ............................... 123 Figure 5-24 Minimum level injection line break: RCS pressure at RVV after 4 hours ......... 124 Figure 5-25 Minimum level injection line break: containment pressure at RVV ................... 125 Figure 5-26 Minimum level injection line break: containment pressure at RVV after 4 hours .............................................................................................................. 126 Figure 5-27 Minimum level injection line break: RVV2 flow .................................................. 127 Figure 5-28 Minimum level steam generator tube failure: core inlet temperature................ 128 Figure 5-29 Minimum level steam generator tube failure: maximum clading level .............. 129 Figure 5-30 Minimum level steam generator tube failure: riser collapsed liquid level above top of active fuel ................................................................................................ 130 Figure 5-31 Minimum level steam generator tube failure: containment liquid level above bottom ............................................................................................................... 131 Figure 5-32 Minimum level steam generator tube failure: RCS pressure at RVV ................ 132 © Copyright 20187 by NuScale Power, LLC ix
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-33 Minimum level steam generator tube failure: RCS pressure at RVV after 4 hours .............................................................................................................. 133 Figure 5-34 Minimum level steam generator tube failure: containment pressure at RVV .... 134 Figure 5-35 Minimum level steam generator tube failure: containment pressure at RVV after 4 hours ...................................................................................................... 135 Figure 5-36 Minimum level steam generator tube failure: RVV2 flow ................................... 136 Figure 6-1 Percent boric acid at solubility limit as a function of temperature ...................... 148 © Copyright 20187 by NuScale Power, LLC x
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Abstract This report presents (1) the NuScale Power, LLC, methodology used to evaluate the emergency core cooling system (ECCS) long-term cooling capability of the NuScale Power Module (NPM) after a successful initial short-term response to a design basis event, and (2) evaluation results demonstrating satisfactory ECCS performance during long-term cooling. The report includes discussion on the transition to long-term cooling for events that assume the use of the decay heat removal system (DHRS) as well as those that actuate the ECCS early in a design basis event. This report is applicable to long-term cooling capability following both loss-of-coolant (LOCA) and non-LOCA design basis events. The long-term cooling methodology is an extension of the NuScale LOCA evaluation model (EM) (Reference 8.2.1), and thus uses a graded approach to the EM development and assessment process (EMDAP) defined in Regulatory Guide 1.203. The phenomena of high importance developed in the long-term cooling phenomena identification and ranking table (PIRT) analysis performed for long-term cooling EM are discussed in this report. The long-term cooling evaluation results demonstrate ECCS conformance with the acceptance criteria in 10 CFR 50.46(b)(4) and 10 CFR 50.46(b)(5) for coolable geometry and long-term cooling for the long-term cooling phase when stable natural circulation has developed through the ECCS configuration. This report also demonstrates conformance to NuScale Principal Design Criterion 35 along with compliance with relevant Acceptance Criteria given by the Design Specific Review Standard for NuScale Small Modular Reactor Design, Sections 6.3 and 15.6.5 (Reference 8.2.3 and Reference 8.2.4, respectively). This report provides information supplementing NuScale Final Safety Analysis Report Section 6.2, Section 6.3, Section 15.0, and Section 15.6.5.
© Copyright 20187 by NuScale Power, LLC 1
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Executive Summary The NuScale Power Module (NPM) is designed to successfully cool down after experiencing an initiated event and transition to a long-term cooling condition. The purpose of this report is to define the evaluation model (EM) for evaluating long-term cooling and demonstrate that ECCS performance meets the regulatory criteria during long-term cooling in a conservative fashion. The long-term cooling (LTC) analyses are performed for 72 hours to demonstrate that the module(s) will remain in a safe, stable condition with the ECCS operating without credit for normal AC power, the nonsafety-related DC power system, or any operator action for 72 hours after event initiation. The LTC EM is developed to conservatively model the long term global heat removal capabilities of the emergency core cooling system (ECCS) and the reactor pool. This methodology ensures that the criteria of 10 CFR 50.46(b)(4) and 10 CFR 50.46(b)(5) are met. In addition, this evaluation demonstrates conformance with the ECCS Principal Design Criterion (PDC) 35, as described in the NuScale Final Safety Analysis Report (FSAR) Section 3.1. Additional regulatory guidance for the design of the ECCS is found in the Design Specific Review Standard (DSRS) for NuScale Small Modular Reactor Design, Section 6.3 relating to gravitational head providing sufficient core cooling for 72 hours, without operator actions and without nonsafety-related onsite or offsite power. The NuScale DSRS Section 15.6.5 refers to the evaluation of post-LOCA long-term cooling for decay heat removal by assuring boric acid precipitation is prevented for all break locations and sizes and asks the reviewer to verify that procedures are in place to assure boron precipitation is mitigated. DSRS Section 15.6.5 also specifies that steam generator tube failure (SGTF) be reviewed for the potential coolant inventory loss from the reactor vessel to the secondary side. The report describes the following NuScale-specific LTC acceptance criteria that were developed to assure that regulatory requirements of 10 CFR 50.46 are met: 1) collapsed liquid level in the reactor vessel remains above the top of the core, 2) boron concentrations in the core region remain below the boron solubility limit, and 3) fuel cladding temperatures predicted by NRELAP5 are maintained at an acceptable level.The report describes the following NuScale-specific acceptance criteria that were developed to assure that regulatory requirements of 10 CFR 50.46 are met: 1) collapsed liquid level in the reactor vessel remains above the top of the core, 2) cladding temperatures predicted by NRELAP5 remain acceptably low, 3) margin to the critical heat flux (CHF) predicted by NRELAP5 using a CHF correlation appropriate to the fluid conditions is maintained, 4) coolable geometry is maintained, and 5) the core remains subcritical. The third criterion is demonstrated by the first two criteria and showing that the minimum critical heat flux ratio (MCHFR) prediction for the short term LOCA is acceptable. The fifth criteria is not applicable to the LTC cooling condition since no mechanism to push a large volume of diluted water into the core inlet exists, and therefore no credible mechanism for recriticality due to boron dilution exists. The long-term phase of core cooling starts once the ECCS is actuated and the NPM is configured such that steam from the pressurizer region is released to the containment vessel (CNV) through the reactor vent valves (RVVs) and condenses on the CNV wall collecting in the bottom of the CNV, then flowing through the reactor recirculation valves (RRVs) to the core inlet. This recirculation flow loop continues as the NPM is cooled. The long-term cooling configuration is reached through both LOCA and some non-LOCA initiating events.
© Copyright 20187 by NuScale Power, LLC 2
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 The LTC EM is developed using a graded approach to the evaluation model development and assessment process (EMDAP) defined in Regulatory Guide (RG) 1.203. The approach for the long-term cooling EM utilizes the NuScale LOCA EM (Reference 8.2.1). The LTC EM focuses on the phenomena identification and ranking table (PIRT) process to identify the important parameters which are specifically addressed. An extensive PIRT was developed for LTC. Each important parameter is discussed and evaluated in this report as it relates to the LTC EM. The LTC EM uses the proprietary NRELAP5 systems analysis computer code as the computational engine, derived from the Idaho National Laboratory RELAP5-3D© computer code. The models and correlations used by the NRELAP5 code were reviewed and determined to be, where appropriate, modified for use within the long-term cooling EM. The NRELAP5 model is validated through the assessment of NIST-1 facility tests and comparison of NRELAP5 predictions to test results. Comparison of the NRELAP5 model to the NIST-1 test results demonstrate that the NRELAP5 code adequately predicts the NPM conditions both in the RPV and the CNV. The methodology for the NPM thermal-hydraulic response and boron precipitation evaluation are presented in this report. There are threetwo LTC general conditions which address the thermal-hydraulic response and boron precipitation: (1) maximum cooldown to minimize the RPV core liquid volume in the riser regioninlet region temperature for addressing boron precipitation, and the (2) minimum collapsed liquid level to minimize the volume of liquid in the riser region and above the active fuel to demonstrate core coverage and address boron precipitationabove the active fuel, and (3) minimum cooldown to maximize the fuel cladding temperature. The methodology is demonstrated in the report by presenting the limiting results of a base LOCA case for the letdowninjection line break (LDILBRK) utilizing conservative worst case conditions determined by sensitivity calculations. In addition the SGTF results are presented. Sensitivity cases performed considered the following assumptions:
- single active failure, ECCS valve failure to open is the relevant single active failure to consider in the LTC analyses
- decay heat, ranging from no decay heat to 120 percent of nominal
- heat transfer from the RPV to CNV, ranging from adiabatic to 1000 percent of nominal
- DHRS operationheat transfer from the CNV to pool, ranging from 20 percent to 1000 percent
- reactor pool temperature, ranging from 4065 degrees F to 210 degrees F
- reactor pool level, down to 455 feet (Nominal at 69 feet), 45 feet for LTC with decay heat from twelve modules
- Expansion factor used to account for compressible flow through RVVsreactor pool volume effect on calculated pool temperature heatup from initial conditions
- non-condensable gas effect
- pressurizer level, down to 20 percent level
© Copyright 20187 by NuScale Power, LLC 3
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 In all analyzed cases, the core remained covered, with greater than 2.252.8 feet of collapsed liquid level in the riser above the top of the core. Possible leakage from the CNV was found to have a negligible impact on the results. The cases identified as most limiting, maximum temperature with injection line break, minimum temperature with injection line break, minimum level with injection line break, and minimum level with steam generator tube failurethe minimum cooldown, maximum cooldown, and SGTF with decay heat removal system (DHRS), all showed consistently decreasing reactor coolant system (RCS) and cladding temperatures, supporting the conclusion that the ECCS is capable of providing adequate cooling for the 72 hour evaluation period. In order to evaluate the criterion for maintaining coolable geometry, the possibility of boron precipitation is evaluated in this report. The methodology for determining boron precipitation is conservative, as it assumes the maximum boron concentration and a minimum volume that includes the core and riser region for boron mixing and that all the RPV boron remains within this region. The maximum boron concentration is shown in this report to remain below the solubility limit for the minimum RCS temperatures reached within the 72 hour evaluation period for long-term cooling. The long-term cooling methodology, boron precipitation methodology, and analysis results presented in this report provide supplemental information designed to inform the NRCs evaluation of NuScale Final Safety Analysis Report Sections 6.2, 6.3, 15.0 and 15.6.5.
© Copyright 20187 by NuScale Power, LLC 4
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 1.0 Introduction 1.1 Purpose The purpose of this report is to present the NuScale evaluation model (EM) used to evaluate the long term module response during emergency core cooling system (ECCS) operation and to present evaluation results demonstrating satisfactory ECCS performance during long-term cooling. This report describes the ECCS long-term cooling (LTC) analysis scope, acceptance criteria, and methodology for demonstrating that the acceptance criteria are met for the NuScale Power Module (NPM). The LTC analysis scope is defined based on the applicable regulatory requirements, NuScale-specific requirements for the design, and considering relevant aspects of the NuScale design that affect the long-term transient progression. 1.2 Scope In the NPM, the ECCS is designed to operate following a loss-of-coolant accident (LOCA) or after the inadvertent opening of a valve that allows release of primary reactor coolant into containment, or if power to the ECCS valve actuators is lost and the reactor coolant system (RCS) is at sufficiently low pressure. Due to the unique NuScale ECCS design, these different scenarios are considered in the analysis of the ECCS long-term cooling. The long-term cooling phase of decay heat removal is defined as beginning when ECCS actuates to open the RVVs and RRVs, the recirculation flow is established, and the pressures and levels in containment and the RPV approach a stable condition (Reference 8.2.1, Section 4.2). This report summarizes the following:
- long term NuScale design basis event progression following ECCS actuation
- regulatory requirements and NuScale-specific design requirements applicable to LTC
- LTC acceptance criteria
- NuScale LTC phenomena identification and ranking table (PIRT)
- analysis tools, qualification of the tools, and methodology for demonstrating that the LTC acceptance criteria are met
- results of the LTC analyses.
The following LTC analysis areas are addressed in this report:
- demonstration of long-term core cooling following ECCS actuation
- evaluation for boron precipitation The following areas are outside scope of this report:
© Copyright 20187 by NuScale Power, LLC 5
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- The non-LOCA and LOCA evaluation models for the short-term time periods are covered in separate methodology reports. However the transition between short term LOCA and non-LOCA initiating events is addressed to demonstrate accurate boundary conditions are simulated upon the onset of long term cooling.However, the transition between a non-LOCA initiating event, including steam generator tube failure (SGTF) events, and ECCS long-term cooling is considered in this report.
- The effects of debris on ECCS operation that are the subject of the NRC generic safety issue 191 are outside scope of this report. The NuScale design and debris loads have been assessed to ensure that the system and its components will operate as designed under long-term ECCS operating conditions. The LTC analyses are performed assuming a clean core condition without debris.
- Assessment of the NuScale design return to power due to overcooling, assuming one control rod stuck out of the core, is outside the scope of this report. For the LTC calculations in this report, the heat source is decay heat.
- Assessment of a station blackout is outside the scope of this report and covered in separate analysis.
- Analysis of long-term decay heat removal system (DHRS) performance and decay heat removal is addressed by separate analyses.
- This EM does not assess seismic issues, which are covered in separate methodologies and assessments.
- Critical heat flux (CHF) evaluation is only of interest in the short-term response of the events analyzed in this document. Short-term LOCA CHF is addressed by the NuScale LOCA EM (Reference 8.2.1). For long-term cooling, a collapsed liquid level above the top of active fuel (TAF) and acceptably low cladding temperatures calculated by NRELAP5 are considered sufficient to demonstrate that CHF does not occur.
The ECCS long-term cooling analyses provided in this report address all design basis events that evolve to the configuration where operation of ECCS is needed for long-term cooling, as illustrated in Figure 1-1. These analyses are relevant for both LOCA and non-LOCA initiated events. © Copyright 20187 by NuScale Power, LLC 6
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 1-1 Illustration of the scope and analyses covered by long-term cooling methodology © Copyright 20187 by NuScale Power, LLC 7
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 1.3 Abbreviations and Definitions Table 1-1 Abbreviations Term Definition AC alternating current ANS American Nuclear Society ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CHF critical heat flux CNV containment vessel CVCS chemical and volume control system DCA Design Certification Application DHRS decay heat removal system DSRS Design Specific Review Standard ECCS emergency core cooling system EM evaluation model EMDAP evaluation model development and assessment process FOM figure of merit GDC Generic Design Criterion HZP hot zero power IAB inadvertent actuation block LDILBRK letdowninjection line break LOCA loss-of-coolant accident LTC long-term cooling NIST-1 NuScale Integral System Test NPM NuScale Power Module NRC Nuclear Regulatory Commission PDC principal design criteria PIRT phenomena identification and ranking table PZR pressurizer RCS reactor coolant system RG Regulatory Guide RPV reactor pressure vessel RRV reactor recirculation valve RVV reactor vent valve SAF single active failure SG steam generator SGTF steam generator tube failure TAF top of active fuel UHS ultimate heat sink © Copyright 20187 by NuScale Power, LLC 8
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Table 1-2 Definitions Term Definition Cv Flow coefficient One of the acceptance criteria defined in RG 1.203. Excellent agreement applies when the code exhibits no deficiencies in modeling a given behavior. Major and minor phenomena and trends are Excellent agreement correctly predicted. The calculated results are judged to agree closely with the data. The calculation will, with few exceptions, lay within the specified or inferred uncertainty bands of the data. The code may be used with confidence in similar applications. A parameter selected to characterize the plant long-term cooling Figure of merit response. Those postulated accidents that result in a loss of reactor coolant at a rate in excess of the capability of the reactor makeup system from Loss-of-coolant accident breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Reactor coolant system transients described in the NUREG-0800 Standard Review Plan Sections 15.1, 15.2, 15.4, and 15.5, and other comparable transients that may be unique to the NuScale system. Non-LOCA transient Other sections in the standard review plan are specific to events with reactor coolant pumps, LOCA, radiological analysis, anticipated transient without scram, or boiling water reactors, and are outside of the scope of non-LOCA transients. One of the acceptance criteria defined in RG 1.203. Reasonable agreement applies when the code exhibits minor deficiencies. Overall, the code provides an acceptable prediction. All major trends and phenomena are correctly predicted. Differences between calculation Reasonable agreement and data are greater than deemed necessary for excellent agreement. The calculation will frequently lie outside but near the specified or inferred uncertainty bands of the data. However, the correct conclusions about trends and phenomena would be reached if the code was used in similar applications. © Copyright 20187 by NuScale Power, LLC 9
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 2.0 Regulatory Requirements and Roadmap 2.1 Background In the NPM design, there are two systems that may perform the safety-related functions of decay heat and residual heat removal following an anticipated operational occurrence or accident. The DHRS provides decay and residual heat removal while RCS inventory is retained inside the reactor pressure vessel. If RCS inventory is redistributed between the reactor pressure vessel and the containment vessel (CNV), due to a pipe break LOCA or RCS valve opening event, or opening of the ECCS valves, the ECCS provides decay and residual heat removal. The scope of this EM addresses ECCS long-term cooling. As an advanced passive plant design, the NuScale plant is designed such that:
- protection against design basis events is through passive means for at least 72 hours, and
- no operator actions are required for at least 72 hours for design basis events.
Therefore, in the NuScale design, after initial operation of the ECCS, the safety-related systems continue to provide decay and residual heat removal, without operator actions, for at least 72 hours for design basis events. In the NPM, the ECCS is designed to operate following a LOCA or after the inadvertent opening of a valve that allows release of primary reactor coolant into containment, or if power to the ECCS valve actuators is lost and the RCS is at sufficiently low pressure. Due to the unique NuScale ECCS design, these different scenarios are considered in the analysis of the ECCS long-term cooling. Ultimately these scenarios will converge towards a similar long-term cooling transient. For the design basis safety analyses, reactor trip and actuation of the passive safety systems to mitigate the event will generally occur early in the transient progression. Analysis of the short-term design basis event progression is performed following the appropriate methodology. This report addresses the interface with the short-term analyses, the acceptance criteria applicable to the longer term transient progression to LTC with ECCS, and how these acceptance criteria are met for the NuScale design. 2.2 Regulatory Requirements and Guidance The NRC regulations and regulatory guidance applicable to the LTC methodology are described in this section. The elements of the LTC methodology that address each of these regulations and guidance documents are discussed. 2.2.1 Regulatory Requirements 10 CFR 50.46 (a) provides two options for an acceptable NuScale LOCA EM. Paragraph 50.46(a)(i) allows for a best-estimate approach to be followed and Paragraph 50.46.(a)(ii) allows for the conservative deterministic approach detailed in 10 CFR 50 Appendix K. As the LTC EM is an extension of the NuScale LOCA EM (Reference 8.2.1), the disposition of the 10 CFR 50 Appendix K requirements that apply to the long- © Copyright 20187 by NuScale Power, LLC 10
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 term cooling phase are applied in the same manner as for the LOCA EM. Since the NuScale LOCA EM (Reference 8.2.1) and the LTC EM are equivalent with regard to all Appendix K requirements, no further exemptions to the Appendix K requirements are required for the LTC EM beyond those identified in Reference 8.2.1. The NuScale Principal Design Criterion (PDC) 35, based on General Design Criterion 35, establishes the required safety function of the ECCS, as described in FSAR Section 3.1 of the NuScale DCA. The portion of the PDC of interest to the LTC methodology is identical to 10 CFR 50, Appendix A, General Design Criterion 35, and states: A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single failure. 10 CFR 50.46(b) implements GDC 35, and thus NuScale PDC 35, by establishing specific acceptance criteria for ECCS cooling performance. The applicable regulatory criteria from 10 CFR 50.46(b) regarding long-term ECCS performance (Reference 8.2.2) include the following: (4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling. (5) Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. 10 CFR 50.46 applies to ECCS performance following a LOCA. For the NPM, the long-term core cooling ECCS requirements following a LOCA are fulfilled through the actuation of the passive ECCS. While 10 CFR 50.46 does not address ECCS performance associated with non-LOCA events for long-term core cooling, the ECCS removes residual and core decay heat whenever the NPM transitions to the ECCS configuration. 2.2.2 Regulatory Guidance NRC review guidance regarding the ECCS requirements in DSRS Section 6.3 (Reference 8.2.3) includes the following from page 6.3-2. © Copyright 20187 by NuScale Power, LLC 11
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 For advanced passive reactors that rely on gravitational head to provide ECCS injection to the reactor coolant system (RCS), the RCS should be designed such that the available gravitational head is sufficient to provide adequate core cooling when depressurized. For advanced reactors which rely on passive safety-related systems and equipment to automatically establish and maintain safe-shutdown conditions for the plant, these passive safety systems must be designed with sufficient capability to maintain safe shutdown conditions for 72 hours, without operator actions and without nonsafety-related onsite or offsite power. The following review guidance from DSRS Section 15.6.5 (Reference 8.2.4) refers to the evaluation of post-LOCA long-term cooling for decay heat removal, and for assessment of boric acid precipitation. An evaluation of post-LOCA long-term cooling should also be performed to identify the operator actions to successfully control and prevent boric acid precipitation. Analyses of small break LOCAs should be performed to identify the timing for boric acid precipitation. A spectrum of small breaks should also be analyzed to identify other means to control boric acid precipitation when RCS pressure remains too high to enable flushing of the core. All equipment and operator action times should also be clearly identified in the analyses. From the DSRS page 15.6.5-4, the reactor systems review of this section includes the following. F. The results of the post-LOCA long-term cooling analyses to assure that an acceptable model has been employed to identify the timing of boric acid precipitation for all break locations and sizes. The review will also verify that an adequate procedure has been devised to control boric acid precipitation for all breaks to assure long-term cooling. and, Steam generator tube rupture events shall also be reviewed as part of the LOCA break spectrum analysis. The reviewer shall review the potential coolant inventory loss from reactor vessel to the secondary side. The transition of an event such as an SGTF or small pipe break outside of containment to cooling by the ECCS with reduced reactor coolant inventory is dispositioned in this report from the perspective of ensuring those event progressions meet all LTC acceptance criteria. In the NuScale design, with normal AC power available an SGTF event will result in the actuation of the DHRS; the inventory reduction from the primary to the secondary is detected and isolated before the ECCS is actuated. The short-term event progression of an SGTF is analyzed using the non-LOCA analysis methodology described in Reference 8.2.5. Similarly, in the NuScale design, with normal AC power available a break in a small pipe outside of containment will result in the actuation of the DHRS; the inventory reduction from the primary to the secondary is detected and isolated by closing the containment isolation valves before the ECCS is actuated. The short-term event progression of a small pipe break outside of containment is analyzed using the non-LOCA analysis methodology © Copyright 20187 by NuScale Power, LLC 12
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 described in Reference 8.2.5. If normal power is assumed to be lost, an SGTF or a small pipe break outside of containment will transition to cooling by the ECCS. 2.3 Acceptance Criteria and Transient Duration 2.3.1 Acceptance Criteria The NuScale-specific acceptance criteria for the LTC analysis and the transient duration for which the acceptance criteria are demonstrated are defined in this section. The NuScale-specific acceptance criteria for the ECCS long-term cooling analyses are:
- 1. core cooling is provided to remove decay and residual heat from the core.
This acceptance criterion is demonstrated in thermal-hydraulic calculations with NRELAP5 by the following:
- a. collapsed liquid level in the reactor vessel remains above the top of the core.
- b. cladding temperatures predicted by NRELAP5 remain acceptably low.
- c. margin to the CHF predicted by NRELAP5 using a CHF correlation appropriate to the fluid conditions is maintained.
- DSRS 15.6.5-10 states If core uncovery is not expected during the entire period of a LOCA, the staff should ensure that a significant number of fuel rods will not be damaged because of local dryout conditions. This may be demonstrated by showing that the limiting fuel rod heat flux remains below the critical heat flux (CHF) at a given pressure after depressurization has taken place. If, however, the heat flux exceeds the CHF, further analyses should be performed to estimate the amount of fuel damage expected from burn-out while the bulk of the core remains covered with water during the LOCA. Fuel damage and potential for radioactivity release to the environment must be consistent with 10 CFR Part 100.
- The NuScale LOCA EM addresses the short-term CHF response to a primary system pipe beak and ECCS actuation. No explicit CHF response is evaluated as part of the LTC calculations; maintaining a collapsed liquid level in the riser above the core, along with demonstrating that cladding temperatures remain acceptably low, are considered sufficient conditions to show MCHFR limits are not challenged. In addition, meeting the criteria that the core remain covered by collapsed liquid level in the riser and that cladding temperatures remain acceptably low assure that the PDC 35 criterion that clad metal-water reaction is limited to negligible amounts is met.
- 2. coolable geometry is maintained.
This acceptance criterion is demonstrated by the boron precipitation analysis that demonstrates that the boron concentration in the core region remains below the solubility limit. © Copyright 20187 by NuScale Power, LLC 13
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- 3. the core remains subcritical.
In the long-term cooling analyses, it is demonstrated that decay heat is removed and the core remains cooled. Temperature changes or changes in core boron concentration can affect reactivity. Reactivity effects due to cooldown, assuming the worst control rod stuck out of the core, are outside scope of this report. During the long-term cooling phase, boiling in the core region is expected to concentrate boron in the liquid in the core and riser region. After ECCS valves open and recirculation is established, liquid from containment enters the reactor pressure vessel through the reactor recirculation valves, circulates into the core region, and vapor is vented into containment through the reactor vent valves where it condenses on the containment wall. The boron concentration of liquid in containment may be lower than the boron concentration of liquid in the core/riser region. However, since flow rates from containment into the reactor pressure vessel through the recirculation valves are low and the boron concentration in the core region will tend to increase due to boiling in the core region, no credible means of introducing a large slug of deborated water unmixed into the core region has been identified for the NuScale design. Therefore, for the long-term cooling analyses, the core heat source is decay heat and not any additional heat due to a possible recriticality once the NPM has begun heat removal in the long-term cooling phase. 2.3.2 Transient Duration The ECCS cooling evaluation can be broken into three stages: (1) blowdown, (2) ECCS depressurization, and (3) LTC. Consistent with the NuScale LOCA EM topical report (Reference 8.2.1), the transition from LOCA to LTC occurs once natural circulation between the RPV and containment has been established and the pressure and liquid levels in the CNV and the RPV approach a stable equilibrium condition. This natural circulation pattern consists of coolant upflow through the core producing steam, steam leaving the RPV through the reactor vent valves (RVVs) and condensing on the cool containment shell, and the condensate being returned from the containment pool to the RPV through the reactor recirculation valves (RRVs). This is a natural transition point into LTC as all LOCA events will evolve to this condition. The assessment of LTC then covers the progression of the event from this point forward. The LTC analyses are performed for 72 hours to demonstrate that the module(s) will remain in a safe, stable condition with the ECCS operating without credit for normal AC power, the nonsafety-related DC power system, or any operator action for 72 hours. The ECCS long-term cooling analyses address the following scenarios:
- 1. ECCS cooling that begins during the short-term event progression. LTC begins where the NuScale LOCA EM analysis ends when ECCS recirculation flow (RCS steam is released to the CNV through the RVVs, condensed on the CNV walls, and condensed liquid re-enters the RPV through the RRVs) and pressures and levels in the RPV and CNV approach a stable equilibrium condition.
© Copyright 20187 by NuScale Power, LLC 14
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- 2. DHRS cooling scenarios that transition to ECCS cooling weretransition into ECCS cooling is considered. These scenarios include:
- a. ECCS cooling that begins after a period of initial DHRS cooling, and
- b. DHRS cooling cases that transition into ECCS cooling.
Long-term cooling analyses demonstrate that if DHRS cooling is provided until either the inadvertent actuation block (IAB) setpoint is reached or 24 hours is reached such that the ECCS timer expires, and decay heat removal transitions to ECCS cooling, then the module(s) will remain in a safe, stable condition for up to 72 hours following the event. DHRS transition cases will also include consideration for SGTF to address inventory loss prior to isolation of the SG. 2.4 Long-Term Cooling Evaluation Model Roadmap Analyses are performed to demonstrate that a nuclear power plant can meet applicable NRC regulatory acceptance criteria for a limiting set of anticipated operational occurrences, infrequent events, and accidents. The EMDAP as defined in RG 1.203 (Reference 8.2.6) provides a structured process to establish the adequacy of a methodology for evaluating complex events that are postulated to occur in nuclear power plant systems. The EM described in this report has been developed for simulating the long-term cooling capability of the NPM during long-term ECCS operation. NRELAP5 is the thermal-hydraulics code used to assess the ECCS performance of the NPM during LTC. The NuScale LOCA evaluation model (Reference 8.2.1) was developed following the EMDAP guidelines of RG 1.203 (Reference 8.2.6). Phenomena identified as high-ranked for ECCS long-term cooling were evaluated with respect to the high-ranked phenomena identified as part of the NuScale LOCA EM development. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the LTC calculations (see Section 3.0 and Section 5.0), a graded approach to the EMDAP is applied for development of the LTC evaluation model. Figure 2-1 shows various elements of EMDAP as defined in RG 1.203 (Reference 8.2.6). The elements of the EMDAP and sections of this report that relate to the elements and steps of the EMDAP are summarized in Table 2-1. © Copyright 20187 by NuScale Power, LLC 15
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Element 1 Establish Requirements for Evaluation Model Capability
- 1. Specify analysis purpose, transient class and power plant class
- 2. Specify figures of merit
- 3. Identify systems, components, phases geometries, fields and processes that should be modeled
- 4. Identify and Rank phenomena and processes Element 2 Element 3 Develop Assessment Base Develop Evaluation Model
- 5. Specify objectives for assessment base
- 6. Perform scaling analyses and identify 10. Establish EM development plan similarity criteria 11. Establish EM structure
- 7. Identify existing data and/or perform IETs 12. Develop or incorporate closure models and SETs to complete the data/base
- 8. Evaluate effects of IET distortions and SET scaleup capability
- 9. Determine experimental uncertainties Element 4 Assess Evaluation Model Adequacy Closure Relations (Bottom-up) Integrated EM (Top-down)
- 13. Determine model pedigree and 16. Determine capability of field equations and applicability to simulate physical numeric solutions to represent processes processes and phenomena
- 14. Prepare input and perform calculations 17. Determine applicability of EM to simulate to assess model fidelity and/or accuracy system components
- 15. Assess scalability of models 18. Prepare input and perform calculations to assess system interactions and global capability
- 19. Assess scalability of integrated calculations and data for distortions.
- 20. Determine EM bases and uncertainties No Yes Return to appropriate Perform plant Adequacy Decision elements, make and event analyses Does code meet adequacy standard?
assess corrections. Figure 2-1 Evaluation model development and assessment process © Copyright 20187 by NuScale Power, LLC 16
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Table 2-1 Evaluation model development and assessment process steps and associated application in the long-term cooling evaluation model EMDAP Step Description EM Section Element 1, Establish Requirements for Evaluation Model Capability The purpose of the LTC methodology is described in Section 1.1. The NuScale LOCA topical report (Reference 8.2.1) Specify analysis provides an overview of the NPM and a description of the 1 purpose, transient class plant operation. This includes the safety systems, the and power plant class. system logic, and operational phases which could occur in the NPM. The regulatory requirements that the methodology is designed to comply with are described in Section 2.2. The NuScale-specific acceptance criteria for LTC are Specify figures of merit identified in Section 2.3. Section 3.0 describes the NPM long-2 (FOMs). term cooling PIRT, including FOMs that are used to develop the PIRT. Identify systems, components, phases, Systems, components, phases and processes are identified 3 geometries, fields, and as a part of the LTC PIRT discussed in Section 3.0. processes that should be modeled. Identify and rank Section 3.0 describes the long-term cooling PIRT. 4 phenomena and processes. Element 2, Develop Assessment Base Section 3.0 describes the high ranked phenomena identified from the PIRT process and how the phenomena are addressed by NRELAP5 assessment or other approach. Many of the high ranked phenomena were Specify objectives for assessed against experimental data as part of the NuScale 5 assessment base. LOCA EM development; additional assessments against NuScale Integral Systems Test-1 (NIST-1) test data were performed as described in Section 4.0. Other parameters are bounded or treated by a conservative methodology in the LTC analyses. A scaling analysis of the LOCA and ECCS has been performed for the NPM based on the NIST-1 facility. The results of the scaling analysis are discussed in the NuScale Perform scaling analysis LOCA topical report (Reference 8.2.1). 6 and identify similarity criteria. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the LTC plant transient calculations, these assessments are adequate for the LTC EM. © Copyright 20187 by NuScale Power, LLC 17
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 EMDAP Step Description EM Section Identify existing data and The NuScale LOCA topical report (Reference 8.2.1) and perform integral effects Section 4.0 of this report provide the results of the tests (IETs) and NRELAP5 validation against the SETs and IETs. 7 separate effects tests (SETs) to complete database. In the NuScale LOCA topical report (Reference 8.2.1), a bottom-up assessment of NRELAP5 closure models and correlations essential to simulate high-ranked PIRT phenomena for LOCA events is presented; this assessment addresses the fidelity of the models and correlations to the Evaluate effects of IET appropriate fundamental or SET data. In Reference 8.2.1, a 8 distortions and SET top-down assessment of the NRELAP5 governing equations scaleup capability. and numerics is presented. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the LTC plant transient calculations, these assessments are adequate for the LTC evaluation model. The NuScale LOCA topical report (Reference 8.2.1) and Determine experimental 9 Section 4.0 of this report address experimental uncertainties uncertainties. for NRELAP5 assessments against the SETs and IETs. Element 3, Develop Evaluation Model The NRELAP5 development plan includes programming Establish EM standards and procedures, quality assurance procedures, 10 development plan. and configuration control, which are summarized in Reference 8.2.1. The NuScale LOCA topical report (Reference 8.2.1) provides a summary of NRELAP5 models and correlations. 11 Establish EM structure. For LTC analysis, the plant model is described in Section 4.0. The LTC methodology for thermal-hydraulic calculations is described in Section 5.0 and the methodology for boron precipitation analysis is described in Section 6.0. The NuScale LOCA topical report (Reference 8.2.1) provides a summary of NRELAP5 models and correlations. Develop or incorporate 12 A full description of the closure models and the associated closure models. equations used in the LTC evaluation model is provided in the NRELAP5 theory and users manuals (Reference 8.2.8). Element 4, Assess Evaluation Model Adequacy Closure Relations (Bottom-up) The NuScale LOCA topical report (Reference 8.2.1) includes a bottom-up assessment of important NRELAP5 models/correlations essential to simulate high-ranked PIRT Determine model phenomena for LOCA events, including discussion of model pedigree and 13 pedigree and applicability. Considering the overlap in high-applicability to simulate ranked phenomena and conservatism applied to input and physical processes. boundary conditions in the LTC plant transient calculations, these assessments are adequate for the LTC evaluation model. © Copyright 20187 by NuScale Power, LLC 18
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 EMDAP Step Description EM Section Prepare input and Reference 8.2.1 and Section 4.0 summarize the results of perform calculations to comparison of NRELAP5 against the selected SETs and 14 assess model fidelity IETs including evaluation of code fidelity and accuracy. and/or accuracy. The NuScale LOCA topical report (Reference 8.2.1) includes discussion on scalability of NRELAP5 models and correlations that are essential to simulate high-ranked PIRT Assess scalability of phenomena for LOCA events. Considering the overlap in 15 models. high-ranked phenomena and conservatism applied to input and boundary conditions in the LTC plant transient calculations, these assessments are adequate for the LTC EM. Element 4, Assess Evaluation Model Adequacy Integrated EM (Top-down) NRELAP5 field equations and the numeric solution scheme Determine capability of are discussed in Reference 8.2.1 and evaluated for their field equations and applicability to NPM LOCA phenomena. Considering the 16 numeric solutions to overlap in high-ranked phenomena and conservatism represent processes and applied to input and boundary conditions in the LTC plant phenomena. transient calculations, these assessments are adequate for the LTC EM. The applicability of the NuScale LOCA EM to simulate the Determine applicability NPM system and components is demonstrated by 17 of EM to simulate assessment of NRELAP5 against NuScale design-specific system components. SETs and IETs as summarized in Reference 8.2.1 and Section 4.0. Prepare input and The NuScale LOCA topical report (Reference 8.2.1) and perform calculations to Section 4.0 summarize the results of assessment of 18 assess system NRELAP5 against NIST-1 IET data. interactions and global capability. The NuScale LOCA topical report (Reference 8.2.1) provides an evaluation of scaling distortions between the NIST-1 LOCA IET data and the NPM design. The scalability Assess scalability of of the EM to represent NPM LOCA phenomena and 19 integrated calculations processes is presented therein. Considering the overlap in and data for distortions. high-ranked phenomena and conservatism applied to input and boundary conditions in the LTC plant transient calculations, these NuScale LOCA EM assessments are adequate for the LTC EM. For the LTC system transient analyses, suitably Determine EM biases conservative input is specified in the plant calculations as 20 and uncertainties. described in Section 5.0 and Section 6.0, considering the effects on the appropriate acceptance criteria. © Copyright 20187 by NuScale Power, LLC 19
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.0 Phenomena Identification and Ranking Table 3.1 Phenomena Identification and Ranking Table Process The purpose of the NuScale LTC PIRT is to provide an assessment of the relative importance of phenomena and processes that may occur in the NuScale module during LTC in relation to specified FOMs. This assessment is part of the process prescribed by Regulatory Guide 1.203 (Reference 8.2.6). The current NuScale LTC PIRT has been developed by a panel of experts for the NPM and is built upon the state-of-knowledge at the time of its development. A comprehensive, integrated PIRT was performed for LTC based on the full event progression. The PIRT panel considered the NPM design to identify systems, components, and subcomponents of the design for which phenomena were assessed. The panel then followed the PIRT process. Phenomena were identified and ranked considering their level of importance relative to identified figures-of-merit (FOM) for LTC. The panel established a knowledge ranking for each of the phenomena. The knowledge level is on a 1 to 4 scale; 4 represents well-known and easily modeled phenomena, while 1 represents a parameter that is not understood and can be difficult to sufficiently model. 3.2 Figures of Merit During post-LOCA long-term cooling, there are three identified FOMs to which the identified phenomena are compared.
- CHFR: The ratio of the heat flux needed to cause CHF phenomena to the actual local heat flux of a fuel rod. Since the core remains covered with water throughout the event, clad does not significantly heat up. As discussed in Section 2.3, during long-term cooling, maintaining a collapsed liquid level in the riser above the core and demonstrating cladding temperatures remain acceptably low indicate that minimum CHFR is not challenged.
- Coolant collapsed level: The coolant level that results if all voids in the vapor-phase coolant are collapsed. If the core remains covered, significant clad heatup is avoided and it is evident that 10 CFR 50.46 criteria of adequate LTC is established.
- Subcriticality: The condition of a nuclear reactor system, in which nuclear fuel no longer sustains a fission chain reaction (that is, the reaction fails to initiate its own repetition, as it would in a reactor's normal operating condition). A reactor becomes subcritical when its fission events fail to release a sufficient number of neutrons to sustain an ongoing series of reactions, possibly as a result of increased neutron leakage or poisons. Note that for long-term cooling, the core heat source is decay heat. This FOM is outside the scope of this report.
3.3 Highly Ranked Phenomena The following sub-sections summarize the phenomena that were ranked high importance by the PIRT panel for the NuScale LTC assessment. The knowledge level assigned by the © Copyright 20187 by NuScale Power, LLC 20
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 PIRT panel and the systems/components where the phenomena were ranked as high importance is also included. The LTC PIRT was a comprehensive, integrated PIRT for LOCA long-term cooling phase of the event progression. The NPM systems and components, and the relevant phenomena were considered in detail. As discussed in the LOCA evaluation model, NRELAP5 is NuScales system thermal-hydraulics code used to calculate the NPM system response during the LOCA long-term cooling event progression. The NRELAP5 code has been assessed against several separate effects and integral effects tests as part of the code development and development of the NuScale LOCA evaluation model to demonstrate the capability to simulate the NPM response to LOCA events (Reference 8.2.1). How the highly ranked phenomena are addressed in the LTC evaluation model is discussed. 3.3.1 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 21
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.2 (( }}2(a),(c) ((
}}2(a),(c) 3.3.3 (( }}2(a),(c)
((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 22
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.4 (( }} 2(a),(c) ((
}}2(a),(c) 3.3.5 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 23
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.6 (( }}2(a),(c) ((
}}2(a),(c) 3.3.7 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 24
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.8 (( }}2(a),(c) ((
}}2(a),(c) 3.3.9 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 25
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.10 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 26
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.11 (( }}2(a),(c) ((
}}2(a),(c) 3.3.12 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 27
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.13 (( }}2(a),(c) ((
}}2(a),(c) 3.3.14 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 28
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.15 (( }}2(a),(c) ((
}}2(a),(c) 3.3.16 (( }}2(a),(c)
((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 29
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.17 (( }}2(a),(c) ((
}}2(a),(c) 3.3.18 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 30
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.19 (( }}2(a),(c) ((
}}2(a),(c) 3.3.20 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 31
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.21 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 32
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.22 ((
}}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 33
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.23 (( }}2(a),(c) ((
}}2(a),(c) 3.3.24 (( }} 2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 34
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.25 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 35
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.26 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 36
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.27 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 37
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.28 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 38
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.29 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 39
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.30 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 40
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.31 (( }}2(a),(c) ((
}}2(a),(c) 3.3.32 (( }}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 41
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.33 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 42
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.34 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 43
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.35 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 44
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.36 (( }}2(a),(c) ((
}}2(a),(c) 3.3.37 (( }} 2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 45
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.38 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 46
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.39 ((
}}2(a),(c)
((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 47
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.40 (( }}2(a),(c) ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 48
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.41 ((
}}2(a),(c)
((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 49
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 3.3.42 (( }} 2(a),(c) ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 50
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 4.0 NRELAP5 Applicability to Long-Term Cooling Analysis The LTC EM uses an earlier version of the NuScale LOCA EM than the EM described in Reference 8.2.1.The LTC EM is developed from the NuScale LOCA EM described in Reference 8.2.1. Specifically, the LTC model is derived from the coarser nodalized LOCA model described in Section 9.6.1 of Reference 8.2.1. The coarser LOCA model is selected to improve calculation performance over long term, quasi-steady state conditions where the fidelity of finer model nodalization is not required. This section describes the LTC model, how the LTC EM was developed and the differences between the LTC EM and the NuScale LOCA EM described in Reference 8.2.1. This section also validates the LTC EM for use in LTC assessments by benchmarking to the NIST-1 facility test results. 4.1 Summary of the Long-Term Cooling Model The NRELAP5 LTC model input file is developed from engineering drawings, calculations, and reference documents. These sources of information provide the numerical information necessary to develop a complete thermal-hydraulic simulation model of the NPM. The types of required information fall into the following NRELAP5 input categories:
- thermal-hydraulic fluid volumes and connecting heat structures
- reactor vessel primary loop lower plenum core riser pressurizer SG primary side downcomer
- reactor kinetics
- reactor vessel secondary system SG secondary steam lines feedwater lines
- CNV
- reactor pool
- DHRS
- ECCS
- chemical and volume control system (CVCS) piping for RCS injection, discharge, and pressurizer spray lines
- material properties
- control systems
- simplified control systems for initialization pressurizer pressure pressurizer level vessel average temperature steam pressure turbine load
- reactor protection system
- engineered safety feature controls
© Copyright 20187 by NuScale Power, LLC 51
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(c)
In the LTC analysis, for limiting calculations ((
}}2(a)(c)((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 52
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c) 4.2 NRELAP5 Validation and Assessments for Long-Term Cooling
((
}}2(a),(b),(c) 4.2.1 Long-Term Cooling Tests at the NIST-1 Facility The description of the NIST-1 facility is provided within the NuScale LOCA topical report (Reference 8.2.1).
((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 53
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 4.2.2 NIST-1 Facility NRELAP5 Model ((
}}2(a),(b),(c) 4.2.3 Integral Assessment of NIST-1 HP-19a 4.2.3.1 Purpose of Assessment The HP-19a test results provide a better understanding of phenomena related to an ECCS reactor vent valve spurious opening (without DHRS). The focus in this report is on LTC period.
4.2.3.2 HP-19a Test Progression The test consists of the following:
* ((
}}2(a),(b),(c) 4.2.3.3 NRELAP5 Prediction of HP-19a
((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 54
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 55
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-1 HP19a transient long-term cooling containment vessel level comparison © Copyright 20187 by NuScale Power, LLC 56
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-2 HP19a transient long-term cooling reactor pressure vessel level comparison © Copyright 20187 by NuScale Power, LLC 57
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-3 HP19a transient long-term cooling containment vessel pressure comparison © Copyright 20187 by NuScale Power, LLC 58
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-4 HP19a transient long-term cooling reactor pressure vessel pressure comparison © Copyright 20187 by NuScale Power, LLC 59
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-5 HP19a transient long-term cooling pool level comparison © Copyright 20187 by NuScale Power, LLC 60
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-6 HP19a transient long-term cooling lower pool temperature © Copyright 20187 by NuScale Power, LLC 61
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-7 HP19a transient long-term cooling pool middle temperature comparison © Copyright 20187 by NuScale Power, LLC 62
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-8 HP19a transient long-term cooling pool upper temperature comparison © Copyright 20187 by NuScale Power, LLC 63
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 4.2.4 Integral Assessment of NIST HP-19b 4.2.4.1 Purpose of Assessment The HP-19b test results provide a better understanding of phenomena related to an ECCS reactor vent valve spurious opening (without DHRS), with the presence of non-condensible gas. The focus of this report is on the LTC period. 4.2.4.2 HP-19b Test Progression The test consists of the following:
* ((
}}2(a),(b),(c) 4.2.4.3 NRELAP5 Prediction of HP-19b
((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 64
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 65
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-9 HP19b Transient long-term cooling containment vessel level comparison
((
}}2(a),(b),(c),ECI Figure 4-10 HP19b transient long-term cooling reactor pressure vessel level comparison © Copyright 20187 by NuScale Power, LLC 66
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-11 HP19b transient long-term cooling containment vessel pressure comparison © Copyright 20187 by NuScale Power, LLC 67
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-12 HP19b transient long-term cooling reactor pressure vessel pressure comparison © Copyright 20187 by NuScale Power, LLC 68
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-13 HP19b transient long-term cooling pool level comparison © Copyright 20187 by NuScale Power, LLC 69
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-14 HP19b transient long-term cooling pool lower temperature comparison © Copyright 20187 by NuScale Power, LLC 70
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-15 HP19b transient long-term cooling pool middle temperature comparison © Copyright 20187 by NuScale Power, LLC 71
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c),ECI Figure 4-16 HP19b transient long-term cooling pool upper temperature comparison 4.2.5 Conclusions from Integral Test Assessments
((
}}2(a),(b),(c)
Considering the validation presented in Reference 8.2.1, and this assessment, NRELAP5 is capable of adequately predicting the key parameters of RPV and CNV pressure and level during the LTC timeframe. 4.3 Loss-of-Coolant Accident / Long-Term Cooling Consistency Evaluation ((
}}2(a),(b),(c) © Copyright 20187 by NuScale Power, LLC 72
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 73
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
© Copyright 20187 by NuScale Power, LLC 74
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-17 Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer pressure through 1 hourLong-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer pressure
© Copyright 20187 by NuScale Power, LLC 75
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-18 Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer pressureLong-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: containment pressure
© Copyright 20187 by NuScale Power, LLC 76
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-19 Loss-of-coolant accident evaluation model nodalization consistency comparison: containment pressure through 1 hourLong-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuel
© Copyright 20187 by NuScale Power, LLC 77
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-20 Long-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuelLoss-of-coolant accident evaluation model nodalization consistency comparison: containment pressure
© Copyright 20187 by NuScale Power, LLC 78
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-21 Long-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer level Loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuel through 1 hour
© Copyright 20187 by NuScale Power, LLC 79
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-22 Long-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: containment collapsed liquid level relative to the top of active fuel Loss-of-coolant accident evaluation model nodalization consistency comparison: riser collapsed liquid level relative to the top of active fuel
© Copyright 20187 by NuScale Power, LLC 80
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-23 Long-term cooling, loss-of-coolant accident evaluation model nodalization consistency comparison: core inlet temperature Loss-of-coolant accident evaluation model nodalization consistency comparison: pressurizer level
© Copyright 20187 by NuScale Power, LLC 81
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-24 Loss-of-coolant accident evaluation model nodalization consistency comparison: containment collapsed liquid level relative to the top of active fuel © Copyright 20187 by NuScale Power, LLC 82
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-25 Loss-of-coolant accident evaluation model nodalization consistency comparison: core inlet temperature © Copyright 20187 by NuScale Power, LLC 83
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-26 Loss-of-coolant accident evaluation model nodalization consistency comparison: core outlet temperature
© Copyright 20187 by NuScale Power, LLC 84
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(b),(c)
Figure 4-27 Loss-of-coolant accident evaluation model nodalization consistency comparison - injection line break: riser collapsed liquid level relative to the top of active fuel © Copyright 20187 by NuScale Power, LLC 85
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.0 Long-Term Cooling Methodology and Evaluation Section 3.0 describes the important phenomena and parameters to evaluate the FOM, which are tied to the acceptance criteria delineated in Section 2.0. This section establishes the LTC decay heat removal methodology. The methodology to address maintaining a coolable geometry by precluding boron precipitation is described in Section 6.0. The results presented in Section 5.6 for the minimum RCS temperatures will be used considered in the Section 6.0 analyses for limiting boron solubility conditions. 5.1 Long-Term Cooling Heat Removal Methodology (( For the LTC phase of heat removal the acceptance criteria addressed are: (1) collapsed liquid level is maintained above the active fuel, and (2) fuel cladding temperature is maintained at an acceptable level. These criteria are demonstrated with the basic established conditions below:
- Maximum temperature: Achieved by minimum cooldown and demonstrates that the fuel cladding temperature is maintained at an acceptable level.Minimum cooldown demonstrates that the fuel cladding temperature is maintained at an acceptable level.
- Minimum temperature: Achieved by maximum cooldown and demonstrates that the collapsed liquid level is maintained above the active fuel and that the minimum temperature supports the criteria that no boron precipitation occurs during the LTC evaluation period.Maximum cooldown demonstrates that the collapsed liquid level is maintained above the active fuel.
- Minimum level: Achieved by maximum cooldown with minimum initial RCS inventory and maximum inventory loss to the CNV, and demonstrates that the collapsed liquid level is maintained above the active fuel. Due to the conservative assumption that all boron is concentrated in the core and riser regions, these conditions also support the criterion that no boron precipitation occurs during the LTC evaluation period.Maximum cooldown with minimum initial RCS inventory and maximum CNV inventory loss demonstrates that the collapsed liquid level is maintained above the active fuel. }}2(a),(c)
Section 2.3 establishes that the ECCS long-term cooling analyses address the following scenarios:
- ECCS cooling begins during the short-term event progression. Long-term cooling begins where the NuScale LOCA EM analysis ends when ECCS recirculation flow (RCS steam is released to the CNV through the RVVs, condensed on the CNV walls, and condensed liquid re-enters the RPV through the RRVs), pressures and levels in the RPV and CNV approach a stable equilibrium condition.
- Transition from DHRS cooling to ECCS cooling is considered. LTC analyses demonstrate that if DHRS provides passive decay heat removal until either the IAB
© Copyright 20187 by NuScale Power, LLC 86
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 setpoint is reached or 24 hours is reached such that the ECCS timer expires, and then decay heat removal transitions to ECCS, the module(s) will remain in a safe, stable condition for up to 72 hours following the event. Decay Heat Removal System transition cases also include consideration for SGTF to address inventory loss prior to isolation of the SG. In order to demonstrate sufficient capability to maintain safe shutdown conditions for 72 hours without operator actions and without nonsafety-related power, two events are considered to ensure that the ECCS can adequately provide long-term core cooling. The two events are:
- LOCA in conjunction with a loss of normal AC power scenario: The CVCS letdown line break at the RPV is analyzed. It is modeled as a double-ended guillotine break of the CVCS discharge pipe at the RPV. At the time of the letdown break initiation, the normal letdown flow path is closed and the break paths are instantly opened.
Based on the results of the LOCA analysis of a spectrum of various break sizes and conditions, the long term trend of all break sizes and locations converge. Therefore, use of only the CVCS letdown line break in the sensitivity calculations is appropriate and satisfactory.
- SGTF, leading to short-term DHRS cooling which then transitions to cooling by ECCS, either on IAB release differential pressure or 24 hours when the ECCS timer expires.
Analysis of LTC only considerscredits long term decay heat removal viathrough the ECCS for the purpose of demonstrating that the top of active fuel remains covered. LTC conditions are also evaluated with DHRS enabled to demonstrate minimum temperature requirements are met for boron precipitation concerns. The sequence of events leading to long-term ECCS cooling are described below.
- 1. ECCS valves open. This may occur after short term DHRS cooling in the event of a non-LOCA transient in conjunction with a loss of normal AC power.
- 2. RCS level begins to drop while CNV level rises.
- 3. Minimum level in the RCS occurs. The minimum level reached occurs during the short term LOCA phase.
- 4. Condensation in the CNV increases CNV level.
- 5. Recirculation flow from the CNV to the RPV through the RRV is established.
- 6. Long term levels stabilize, assuming the pool boundary condition is constant. The stabilization of the long term levels begins the long-term cooling phase of the event.
Analysis of the sensitivities is limited to three days. This timeframe is considered acceptable because: (1) the most severe conditions will have been captured within the 72 hour window analyzed, and any conditions that could reasonably be expected to occur beyond this time period are thus bounded by the 72 hour calculation, and (2) after 72 © Copyright 20187 by NuScale Power, LLC 87
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 hours, operator actions can be credited. When realistic initial operating levels, temperatures, and decay heat are considered, the scope of the LTC analyses bound the expected level decrease in the reactor pool ultimate heat sink over 30 days, for up to twelve modules transferring decay and residual heat to the reactor pool. In the NuScale plant design, up to twelve modules may be operating. The safety systems credited for mitigation of the design basis events are module-specific except for the shared reactor pool portion of the UHS. The LTC analyses consider a range of reactor pool boundary conditions to sufficiently address the effects of one or more modules, up to all twelve modules, transferring decay heat into the reactor pool. These calculations demonstrate the trend in both the RCS temperatures (including fuel cladding temperatures) and collapsed liquid level in the riser. For reference in the figures, (( }}2(a),(c),ECI The single sensitivity cases show collapsed liquid level in the riser relative to the inside bottom of the RPV; the remaining cases show the level relative to TAF. These parameters are used to confirm both the capability of the ECCS to provide long-term cooling as well as show the conditions do not reach those that would induce precipitation of boric acid. For the limiting cases, additional plots are shown for RCS core inlet temperature, RCS pressure, CNV pressure, CNV level, cladding temperature, and flow through one of the RVVs. 5.2 Events Evaluated for Long Term Cooling The ECCS is designed to operate following a LOCA event, after the inadvertent opening of an RPV valve (IORV), or if power to the ECCS valve actuators is lost and the system has depressurized to the IAB release pressure. Therefore, a series of both LOCA and non-LOCA events are identified to evaluate long term ECCS cooling acceptability. For non-LOCA events, emphasis is placed on events which reduce reactor inventory, such as a small line break outside containment or a steam generator tube failure. A DHRS cooldown event and loss of feedwater event were also evaluated to confirm that events which reduce primary inventory are limiting. The following events were evaluated as part of the LTC evaluation:
- LOCA Spectrum - The full LOCA break spectrum was evaluated. This includes break locations at the discharge line, injection line, high point vent line, and pressurizer spray line, (( }}2(a),(c)
The most limiting collapsed riser levels occur for small LOCA breaks immediately after ECCS actuation. Since this time range is covered by LOCA methodology, minimum level during this time is not considered limiting for LTC analysis. Instead, limiting minimum level is determined in the hours following ECCS actuation during the characteristic level depression seen in this time range and consistent with the definition of the NPM conditions for LTC.
- IORV - The inadvertent opening of an RVV or RRV was considered in LTC analysis.
The RSV was not evaluated as the valve size is bounded between the RVV and LOCA steam space breaks. © Copyright 20187 by NuScale Power, LLC 88
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- Steam Generator Tube Failure - The SGTF transient was included in LTC analysis to evaluate the impact of RCS inventory lost to the secondary system. The break was modeled at the top of the steam generator in order to minimize return flow into the RPV.
- DHRS Cooldown - These events generically evaluated the transition from DHRS to ECCS cooling either if DC power is lost and IAB release pressure is reached, or 24 hours after losing AC power. These transients were initiated by a loss of AC and DC power at time zero. For the 24 hour transition cases, the ECCS logic was modified to actuate after 24 hours rather than on IAB release pressure.
A subset of these cases is also performed at an initial PZR level of 20% to provide a bounding evaluation of small line breaks outside containment. This level corresponds to containment system isolation on the low-low pressurizer level signal, which would isolate the break and prevent further inventory loss.
- Loss of Feedwater - The LOFW non-LOCA event was selected to demonstrate that the module temperature and pressure response prior to reactor trip has little influence on long term conditions. This event was modeled by setting feedwater flow to zero at event initiation. A loss of AC power is assumed at reactor trip which allows ECCS actuation after 24 hours.
5.3 Long Term Cooling Analysis Assumptions 5.3.1 Electric Power Availability For LTC analysis, availability of electric power is considered for its impact on ECCS actuation timing. For non-LOCA events, ECCS actuation can only occur at the IAB release pressure if DC power to the valve actuators is lost, or 24 hours after losing AC power. The following scenarios are considered:
- Loss of AC and DC power at time zero was evaluated for all cases unless specified otherwise. Losing DC power at time zero causes ECCS actuation once the IAB release pressure is reached. This timing is earlier relative to actuation on the RPV or CNV level signals for LOCA or the 24 hour timer for non-LOCA, and earlier actuation is limiting for minimum collapsed level.
- Loss of AC power was evaluated at time zero and at reactor trip for some DHRS cooldown and LOFW cases to confirm they are not limiting. These cases actuate ECCS 24 hours after losing AC power.
5.3.2 Single Failure Evaluation The failure of one ECCS division (i.e., one RVV and one RRV) was considered when evaluating sensitivity to minimum ECCS capacity. When maximum ECCS capacity was evaluated, all ECCS valves were assumed to open. Single failures in the secondary system were not considered as these have little influence on the long term results. © Copyright 20187 by NuScale Power, LLC 89
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.3.3 Multi-module Consideration In the NuScale plant design, up to twelve modules may be operating. The safety systems credited for mitigation of the design basis events are module-specific except for the shared reactor pool portion of the UHS. Long term cooling analysis evaluated a single module response to demonstrate that the acceptance criteria are met. The LTC analyses considered a range of reactor pool boundary conditions to sufficiently address the effects of one or more modules, up to all twelve modules, transferring decay heat into the reactor pool. 5.3.4 Long Term Cooling Evaluation Period LTC analysis is limited to three days. This timeframe is considered acceptable because: (1) the most severe conditions will have been captured within the 72 hour window analyzed, and any conditions that could reasonably be expected to occur beyond this time period are thus bounded by the 72 hour calculation, and (2) after 72 hours, operator actions can be credited. When realistic initial operating levels, temperatures, and decay heat are considered, the scope of the LTC analyses bound the expected level decrease in the reactor pool ultimate heat sink over 30 days, for up to twelve modules transferring decay and residual heat to the reactor pool. 5.4 Initial Conditions and Biases As stated in Section 5.1, three scenarios are defined to evaluate LTC acceptance criteria: minimum level, minimum temperature, and maximum temperature. Specific key conditions for these scenarios are defined in Table 5-1. Some cases feature minor variations from these conditions (defined on a case specific basis in the following sections) to evaluate parameter sensitivity. Table 5-1 Default scenario initial conditions and biases ECCS Capacity Single Failures Reactor Power DHRS Enabled RCS P. (psia) PZR Level (%) Pool Level (ft) Non-Decay Heat RCS Avg. T. Pool T. (°F) Expansion Scenario condensable (%)(1) for LOCA(2) (multiplier) (°F) Factor (Y) Gas (lbm) (Area and Cv) 1.2 (LOCA) RVV/ false Minimum Level 102 555 1780 52 65 69 0 minimum 0.7 1.0 (nonLOCA) RRV Minimum true 102 0.8 535 1780 68 65 69 0 maximum 1.0 none Temperature Maximum 1.2 (LOCA) RVV/ false 102 555 1920 52 210 55 ~131 minimum 0.7 Temperature 1.0 (nonLOCA) RRV (1) Lower power, down to 13% initial power, is considered as a separate sensitivity for the minimum temperature cases. (2) DHRS is always enabled for all non-LOCA events. © Copyright 20187 by NuScale Power, LLC 90
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.25.5 Sensitivity Considerations The parameters considered as part of the sensitivity analysis are based on the findings in the PIRT from Section 3.0 and are conservatively applied in Section 5.6 depending on the requirements of the specific scenario. These parameters are as follows:
- decay heat, ranging from no decay heat to 120 percent of nominal
- reactor pool temperature, ranging from 65 degrees F to 210 degrees F
- reactor pool level, ranging from 55 feet to 69 feet
- sensitivity to 45 feet is evaluated to address the possible boil-off of pool liquid due to long-term cooling from all twelve modules providing decay heat to the pool
- non-condensable gas effect
- pressurizer level, down to 20 percent of nominal
- Expansion factor used to account for compressible flow through RVVs
- DHRS operation
- reactor pool temperature is modeled as constant through assuming a very large pool volume The base case to demonstrate LTC acceptability is a letdown line break LOCA case assuming nominal conditions, ((
}}2(a),(c) Non-LOCA DHRS to ECCS transition cases were also evaluated. The limiting non-LOCA event is the SGTF event.
The parameters that are evaluated as part of the sensitivity analysis are based on the findings in the PIRT from Section 3.0. These parameters are as follows:
- single active failure, ECCS valve failure to open is the relevant single active failure to consider in the LTC analyses
- decay heat, ranging from no decay heat to 120 percent of nominal
- heat transfer from the RPV to CNV, ranging from adiabatic to 1000 percent of nominal
- heat transfer from the CNV to reactor pool, ranging from 20 percent to 1000 percent
- reactor pool temperature, ranging from 40 degrees F to 210 degrees F
- reactor pool level, down to 45 feet
- reactor pool volume effect on calculated pool temperature heatup from initial conditions
© Copyright 20187 by NuScale Power, LLC 91
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- non-condensable gas effect
- pressurizer level, down to 20 percent of nominal In addition, inventory loss through possible containment leakage is also considered and inventory loss due to containment leakage iswas found to be insignificant. With conservative assumptions of saturated vapor and an inlet pressure of 1000 psia, the calculated leakage resulted in a decrease in riser level of 0.41 inches per 24 hours.
Over 72 hours, the resultant loss of 1.23 inches of collapsed liquid level in the riser region has no impact on the conclusions drawn from the analysis. 5.35.6 Demonstration of Limiting Results Three scenarios were established to determine the limiting conditions that could develop during the LTC phase. Results are demonstrated for the LOCA injection line break and the SGTF event as these were found to be limiting for collapsed liquid level. The IORV, DHRS cooldown, and LOFW events were non-limiting for collapsed level and core temperatures, and detailed results are not discussed. The following cases are presented:
- maximum temperature with injection line break
- minimum temperature with injection line break
- minimum level with injection line break
- minimum level with SGTF A more detailed description of each case is provided in Section 5.6.1 through Section 5.6.4. The transient response is simulated for 12.5 hours following event initiation. This time range is sufficient to evaluate the influence of each initiating event on the mid-term LTC response. During this time, minimum collapsed level occurs coincident to a local peak in differential pressure between the RPV and CNV. After passing this peak, the differential pressure between the vessels follows a continually decreasing trend as energy is removed from the module. As the differential pressure decreases, the required static head in containment to drive recirculation flow through the RRVs is also reduced, allowing inventory to accumulate back inside the RPV. Riser level continues to recover toward long-term equilibrium which is a function of decay heat, heat transfer from containment to the UHS, total ECCS capacity, and RRV elevation.
Decay heat continually decreases overtime, reducing the pressure differential between the CNV and RPV over the long-term. Accumulation of non-condensible gases overtime is not modeled. Instead, depending on the conservative direction for the specific evaluation, the LTC analyses assume either zero non-condensible gas is present or assume all non-condensible gas dissolved in the RCS, in the pressurizer vapor space, in containment, in the control rod drive mechanisms, and in the RCS degasification line are instantly transported to the containment vessel at transient initiation. Additionally, the reactor pool temperature is fixed during the transient calculation. Therefore, heat transfer from containment to the UHS is only a function of temperature inside containment. Finally, total ECCS capacity and RRV elevation remain fixed during the transient calculation. © Copyright 20187 by NuScale Power, LLC 92
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 It is concluded that once minimum level is reached and level recovery begins, there is no evolving mechanism which would cause the increasing trend in collapsed level to reverse over the long term. System pressures and temperatures also follow a continually decreasing trend with decay heat over time. Since transient minimum collapsed level has been captured and there are no mechanisms to change the cooldown trajectory, explicit transient calculations past 12.5 hours are not required. Instead, a following state-point analysis is performed to save calculation time. This is done by taking module conditions at the end of 12.5 hours, setting core power to a constant value corresponding to decay heat levels at 72 hours, and then allowing system conditions to converge to equilibrium. The state-point analysis results provide final module conditions without needing to explicitly model the quasi-equilibrium, long term response as decay heat slowly decreases to the 72 hour value. The primary purpose of these calculations is to find the limiting minimum core inlet temperature which occurs at 72 hours for boron precipitation analysis. Long term maximum cladding temperature and collapsed level are also evaluated to confirm that acceptance criteria remain satisfied. Final module conditions at 72 hours are discussed in Section 5.6.5. These calculations demonstrate the trend in both the RCS temperatures (including fuel cladding temperatures) and collapsed liquid level in the riser. Combined effect cases were evaluated to determine the limiting conditions that could develop during the LTC phase. These cases are as follows:
- minimum cooldown with letdown line break (LDBRK)
- ((
}}2(a),(c)
- maximum cooldown with LDBRK
- ((
}}2(a),(c)
- maximum cooldown with SGTF
- includes 20 percent pressurizer level
© Copyright 20187 by NuScale Power, LLC 93
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- 100 percent of the ANS decay heat standard, including the actinide contribution, is the conservatively high decay heat assumed for this scenario.
Table 5-1 describes the cases that were evaluated utilizing the criteria described in this list. A more detailed description of each case is provided in Sections 5.3.1 through 5.3.4. © Copyright 20187 by NuScale Power, LLC 94
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Table 5-1 Descriptions of long-term cooling single parameter cases Sensitivity Case Name Description Group (( LDBRK.mincool.RRV.RVV.45ft.hiNCG3
}}2(a),(c)
LDBRK Min Cooldown Sens. (( LDBRK.mincool.RRV.RVV.55ft.hiNCG3
}}2(a),(c)
(( LDBRK Max Cooldown Sensitivities: LDBRK.maxcool.DK0.8.RRV.RVV.new.rvv.fix.att2.full Nom PZR Level, 0.8x Decay Heat
}}2(a),(c)
(( LDBRK Max Cooldown Sensitivities: LDBRK.maxcool.DK1.2.RRV.RVV.new.rvv.fix.att2.full Nom PZR Level, 1.2x Decay Heat
}}2(a),(c)
(( LDBRK Max Cooldown Sensitivities: LDBRK.maxcool.DK1.2.RRV.RVV.new.rvv.fix.lvl20 20% PZR Level, 1.2x Decay Heat
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 95
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Sensitivity Case Name Description Group (( SGTF Max Cooldown SGTF.LtMxT.ecc24.LTC.poolTmin.compr.lvl20 Sensitivities: Min PZR Level
}}2(a),(c) 5.3.15.6.1 Maximum TemperatureMinimum Cooldown Rate The maximum temperature scenario evaluates the combined impact of all conditions that result in a slower cooldown rate in order to maximize cladding temperature. In general, conditions are biased to maximize RCS energy and minimize heat transfer to the UHS.
Specific conditions include:
- minimum ECCS capacity, including ((
}}2(a),(c), and the single failure of one RVV and RRV to open
- DHRS operation disabled
- decay heat with 1.2 multiplier
- RCS conditions biased to maximize initial energy and minimize initial inventory
- non-condensable gas modeled inside containment
- a maximum reactor pool temperature of 210 °F and minimum reactor pool level of 55 feet Results are presented for the LOCA 100% and 5% injection line breaks. Figure 5-1 shows continually decreasing core inlet temperature post-ECCS actuation. Figure 5-2 shows that the maximum temperature scenario does not challenge cladding temperature. The long-term maximum cladding temperature is seen to decrease to a level well below those seen in the short term. The cladding temperature follows a decreasing trend with saturation temperature. Figure 5-3 and Figure 5-4 demonstrate that the core remains covered at all times, and that long term collapsed level is established well above the top of active fuel.
The 100% IL break case shows a long term level which is lower than the 5% IL break. Due to the conservative modeling of the IL break at the same axial elevation of the RRV, the 100% break is sufficiently large to establish a small liquid recirculation path from the riser to containment in the long term. This results in a conservative prediction for long term collapsed riser level and core inlet temperature for the 100% IL break relative to the 5% IL break. Figure 5-5 through Figure 5-8 show that long term pressure follows a decreasing trend, with system pressures converging to the same value for the 100% IL break and 5% © Copyright 20187 by NuScale Power, LLC 96
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 IL break cases. Figure 5-9 demonstrates that stable ECCS cooling is established through the LTC phase. The results demonstrate that cladding temperature follows a long term decreasing trend and that the core remains covered at all times, and all acceptance criteria are satisfied for the maximum temperature scenario. State point conditions at 72 hours are presented in Section 5.6.5 for the 100% IL break case. To evaluate the combined impact of all conditions that result in a slower cooldown rate, the base case (LDBRK) was re-performed with the following changes that were based on the results of the single effect sensitivities described in the previous sections:
* ((
}}2(a),(c)
Figure 5-1 through Figure 5-14 show that the minimum cooldown rate does not result in a larger RCS temperature and pressure, or a larger CNV pressure compared to the base case. In addition, the long-term collapsed liquid level in the riser is not significantly different with the slower cooldown rate, indicating a long-term level difference of less than a foot. The minimum collapsed liquid level for this condition was found to be nonlimiting, however the fuel cladding temperature was found to be highest at 72 hours compared to the other limiting cases. In Figure 5-1 through Figure 5-14, the long-term maximum cladding temperature is seen to decrease to a level well below those seen in the short term. The cladding temperature follows the saturation temperature. This indicates CHF does not occur during LTC with minimum cooldown rate. Although Figure 5-1, Figure 5-2, Figure 5-3, and Figure 5-6 indicate that the system pressures and therefore saturation temperatures are calculated to increase at the end of the 72 hour period, the results of the limiting case assuming an initial pool level of 45 feet demonstrate that even assuming a lower reactor pool level, © Copyright 20187 by NuScale Power, LLC 97
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 sufficient decay heat is removed during the 72 hour time period (Figure 5-8 through Figure 5-10, and Figure 5-13). Figure 5-1 Maximum temperature injection line break: core inlet temperatureReactor coolant system core inlet temperature for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 98
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-2 Maximum temperature injection line break: maximum cladding temperatureReactor coolant system pressure for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 99
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-3 Maximum temperature injection line break: riser collapsed liquid level above top of active fuelContainment vessel pressure for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 100
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-4 Maximum temperature injection line break: containment liquid level above bottomReactor coolant system collapsed liquid level above top of active fuel for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 101
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-5 Maximum temperature injection line break: RCS pressure at RVVContainment vessel level for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 102
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-6 Maximum temperature injection line break: RCS pressure at RVV after 4 hoursPeak cladding temperature for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 103
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-7 Maximum temperature injection line break: containment pressure at RVVFlow through single reactor vent valve for minimum cooldown rate, 55 feet level © Copyright 20187 by NuScale Power, LLC 104
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-8 Maximum temperature injection line break: containment pressure at RVV after 4 hoursReactor coolant system core inlet temperature for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 105
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-9 Maximum temperature injection line break: RVV2 flowReactor coolant system pressure for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 106
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.6.2 Minimum Temperature The minimum temperature scenario evaluates the combined impact of all conditions that result in a faster cooldown rate in order to minimize core liquid temperature and demonstrate that boron precipitation is precluded. In general, conditions are biased to minimize RCS energy and maximize heat transfer to the UHS. Specific conditions include:
- maximum ECCS capacity, including maximum valve area and flow coefficients, no expansion factor penalty applied to RVVs, and no single failure
- DHRS operation is enabled for entire transient duration
- decay heat with 0.8 multiplier RCS conditions are biased to minimize initial energy
- RCS inventory is maximized which increases total boron mass in the system
- zero non-condensable gas is modeled
- a minimum reactor pool temperature of 65 °F and maximum reactor pool level of 69 feet Results are presented for the LOCA 100% and 5% injection line breaks. Figure 5-10 and Figure 5-11 show that temperatures rapidly drop over the first few hours, then continue on a gradually decreasing trend. Figure 5-12 and Figure 5-13 show that after an initial decrease, long term level is quickly established after approximately four hours. Figure 5-14 through Figure 5-17 show that the RCS and CNV pressures become sub-atmospheric over the long term. Figure 5-18 demonstrates that stable ECCS cooling is established through the LTC phase. System temperatures, pressures, level, and ECCS flow are shown to converge toward the same value regardless of initiating break size.
Collapsed level and core inlet temperature remain sufficiently high to preclude boron precipitation, and all acceptance criteria are satisfied for the minimum temperature scenario. State point conditions at 72 hours are presented in Section 5.6.5 for the 100% IL break case. © Copyright 20187 by NuScale Power, LLC 107
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-10 Minimum temperature injection line break: core inlet temperatureContainment vessel pressure for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 108
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-11 Minimum temperature injection line break: maximum cladding temperatureReactor coolant system collapsed liquid level above top of active fuel for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 109
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-12 Minimum temperature injection line break: riser collapsed liquid level above top of active fuelContainment vessel level for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 110
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-13 Minimum temperature injection line break: containment liquid level above bottomPeak cladding temperature for minimum cooldown rate, 45 feet level © Copyright 20187 by NuScale Power, LLC 111
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-14 Minimum temperature injection line break: RCS pressure at RVVFlow through single reactor vent valve for minimum cooldown rate, 45 feet level 5.3.2 Maximum Cooldown Rate To evaluate the combined impact of all conditions that result in a faster cooldown rate, the base case letdown break was re-performed with the following changes that were based on the results of the single effect sensitivities described in the previous sections:
* ((
}}2(a),(c)
© Copyright 20187 by NuScale Power, LLC 112
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
* ((
}}2(a),(c)
In addition to the changes listed above, for some of the runs it was necessary to isolate heat transfer to the secondary side (SG tubes and DHRS) in order to allow for code convergence. Although this is not consistent with biasing for maximum cooldown, the effect was found to be negligible based on sensitivity calculations. Based on the results presented in this section, the maximum cooldown presents the limiting conditions in terms of collapsed level and minimum RCS temperature. The RCS and CNV pressures become sub-atmospheric, further reducing the temperature by significantly decreasing the saturation temperature. However, for all of the cases presented here, adequate core cooling is maintained as the collapsed level does not fall below the TAF, and the RCS temperature remains stable and acceptably low. The minimum collapsed liquid level was found to be 2.271 feet above the TAF for the 1.2 multiplier decay heat case. Figure 5-15 through Figure 5-35 demonstrate that the long-term maximum cladding temperature decreases to a level well below those seen in the short-term LOCA results. This indicates CHF does not occur during LTC with maximum cooldown rate conditions. © Copyright 20187 by NuScale Power, LLC 113
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-15 Minimum temperature injection line break: RCS pressure at RVV after 4 hoursReactor coolant system core inlet temperature for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 114
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-16 Minimum temperature injection line break: containment pressure at RVVReactor coolant system pressure for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 115
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-17 Minimum temperature injection line break: containment pressure at RVV after 4 hoursContainment vessel pressure for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 116
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-18 Minimum temperature injection line break: RVV2 flowReactor coolant system collapsed liquid level above top of active fuel for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 117
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.6.3 Minimum Level The minimum level scenario evaluates the combined impact of all conditions that reduce collapsed liquid level above the top of active fuel. Level is minimized when the long term differential pressure between the RPV and CNV is maximized. Under this condition, sufficient inventory must accumulate inside containment to create enough static head to allow coolant recirculation back into the RPV through the RRVs. As containment inventory increases, RPV inventory coincidently decreases. In general, this scenario is achieved by maximizing heat transfer from containment to the UHS, minimizing ECCS capacity, and maximizing RCS energy. Specific conditions include:
- minimum ECCS capacity, including minimum valve area and flow coefficients, expansion factor of Y=0.7 applied to RVVs, and the single failure of one RVV and RRV to open
- DHRS operation is disabled (DHRS would provide an additional means of RPV pressure relief which is non-conservative for minimizing level)
- decay heat with 1.2 multiplier
- initial RCS temperature is maximized
- initial RCS inventory is minimized
- zero non-condensable gas is modeled
- a minimum reactor pool temperature of 65 °F and maximum reactor pool level of 69 feet Results are presented for the LOCA 100% IL break, which is the limiting collapsed level case for LTC analysis, and the 5% IL break. Figure 5-21 shows the minimum collapsed level for the 100% IL break is 2.8 feet above TAF and occurs approximately 3.6 hours after ECCS actuation. It is noted that Figure 5-22 shows the 5% IL break has an overall lower minimum level than the 100% IL break, however the rapid drop in level occurs coincident with ECCS actuation and is characteristic of small line LOCA breaks. As this sudden reduction in level occurs before recirculation flow is established through the RRVs, the minimum level for the 5% IL break is covered by the LOCA EM and is not considered limiting for LTC analysis. Both cases show similar trends for long term level recovery. System temperatures and pressures and fuel temperatures are shown to converge toward the same value regardless of initiating break size (Figure 5-19 and Figure 5-20 and Figure 5-23 through Figure 5-26). Figure 5-27 demonstrates that stable ECCS cooling is established through the LTC phase.
The results demonstrate that the core remains covered at all times, and collapsed level and core inlet temperature remain sufficiently high to preclude boron precipitation at all times during the LTC phase. All acceptance criteria are satisfied for the minimum level scenario. State point conditions at 72 hours are presented in Section 5.6.5 for the 100% IL break case. © Copyright 20187 by NuScale Power, LLC 118
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-19 Minimum level injection line break: core inlet temperatureContainment vessel level for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 119
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-20 Minimum level injection line break: maximum cladding temperaturePeak cladding temperature for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 120
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-21 Minimum level injection line break: riser collapsed liquid level above top of active fuelFlow through single reactor vent valve for maximum cooldown, 0.8 decay heat © Copyright 20187 by NuScale Power, LLC 121
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-22 Minimum level injection line break: containment liquid level above bottomReactor coolant system core inlet temperature for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 122
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-23 Minimum level injection line break: RCS pressure at RVVReactor coolant system pressure for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 123
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-24 Minimum level injection line break: RCS pressure at RVV after 4 hoursContainment vessel pressure for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 124
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-25 Minimum level injection line break: containment pressure at RVVReactor coolant system collapsed liquid level above top of active fuel for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 125
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-26 Minimum level injection line break: containment pressure at RVV after 4 hoursContainment vessel level for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 126
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-27 Minimum level injection line break: RVV2 flowPeak cladding temperature for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 127
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.6.4 Steam Generator Tube Failure with Minimum Level Conditions The SGTF event was identified as the limiting non-LOCA case in terms of minimum collapsed riser level. This scenario was run with the minimum level conditions identified in Section 5.6.3, except that a decay multiplier of 1.0 was applied, and the DHRS is not disabled for non-LOCA events. A loss of normal AC and DC power is assumed at event initiation, causing ECCS actuation once the system has depressurized to the IAB release setpoint. Except for ECCS actuation occurring later in time, Figure 5-28 through Figure 5-36 demonstrate similar trends in long term conditions as seen for the LOCA IL break presented in Section 5.6.3. Adequate core cooling is maintained even with the additional inventory loss of the SGTF. The minimum collapsed liquid level for the SGTF event was non-limiting compared to the results presented in Section 5.6.3. Figure 5-28 Minimum level steam generator tube failure: core inlet temperatureFlow through single reactor vent valve for maximum cooldown, 1.2 decay heat © Copyright 20187 by NuScale Power, LLC 128
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-29 Minimum level steam generator tube failure: maximum cladding temperatureReactor coolant system core inlet temperature for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 129
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-30 Minimum level steam generator tube failure: riser collapsed liquid level above top of active fuelReactor coolant system pressure for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 130
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-31 Minimum level steam generator tube failure: containment liquid level above bottomContainment vessel pressure for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 131
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-32 Minimum level steam generator tube failure: RCS pressure at RVVReactor coolant system collapsed liquid level above top of active fuel for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 132
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-33 Minimum level steam generator tube failure: RCS pressure at RVV after 4 hoursContainment vessel level for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 133
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-34 Minimum level steam generator tube failure: containment pressure at RVVPeak cladding temperature for maximum cooldown, 1.2 decay heat, 20% pressurizer level © Copyright 20187 by NuScale Power, LLC 134
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-35 Minimum level steam generator tube failure: containment pressure at RVV after 4 hoursFlow through single reactor vent valve for maximum cooldown, 1.2 decay heat, 20% pressurizer 5.3.3 Decay Heat Removal System to Emergency Core Cooling System Transition with Maximum Cooldown Conditions for Steam Generator Tube Failure In addition to the letdown break scenario, non-LOCA initiating events with DHRS active to remove decay heat until ECCS actuation (i.e. at the IAB release pressure or at 24 hours) were considered. Figure 5-36 through Figure 5-42 present results from the SGTF initiating event with loss of normal AC power and DHRS active to remove decay heat until ECCS valve opening at 24 hours. This scenario was run with max cooldown rate conditions and 20 percent initial pressurizer level, since the previous sensitivities demonstrate this condition to be most limiting with respect to adequacy of ECCS in maintaining core cooling. As illustrated by the results presented in this section, and sensitivities where ECCS valves opened at the IAB release pressure, the effects of SGTF and DHRS with the maximum cooldown case does not significantly affect the previous maximum cooldown conclusions. Adequate core cooling is maintained even with the additional inventory loss of the SGTF. The minimum collapsed liquid level for the SGTF maximum cooldown cases were non-limiting compared to the results presented in Section 5.3.2.
© Copyright 20187 by NuScale Power, LLC 135
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-36 Minimum level steam generator tube failure: RVV2 flowReactor coolant system core inlet temperature for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 136
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.6.5 State-point Evaluation at 72 Hours State-point module conditions after 72 hours are presented in Table 5-2. Results are included for the following events: the IL break cases presented in Sections 5.6.1, 5.6.2, and 5.6.3, the maximum temperature IL break case with reactor pool level set to 45 feet, and three minimum temperature cases which assume an initial reactor power of 13%. The following conclusions are drawn:
- While a reactor pool level of 45 feet did increase core inlet temperature by 10°F relative to the 55 feet case, this increase is not significant enough to impact the conclusion that cladding temperature is not challenged during LTC.
- Collapsed level above TAF was maintained for all cases after 72 hours. Limiting minimum level occurred earlier in the LTC phase with level recovering to equilibrium by 72 hours.
- Margin to boron precipitation was demonstrated for all cases. The low power initial condition results in lower core inlet temperature after 72 hours.
Table 5-2 Results of state-point analysis at 72 hours Collapsed Riser Level Core Inlet Temperature Boron Precipitation Margin above TAF (°F) (°F) Case (ft) Transient Transient Transient State-point State-point State-point at 12.5 at 12.5 at 12.5 at 72 Hours at 72 Hours at 72 Hours Hours Hours Hours Maximum Temperature 292.8 270.4 8.9 9.1 208.9 187.8 IL break Minimum Temperature 152.8 140.4 10.0 10.4 73.1 62.3 IL break Minimum Level IL break 165.3 154.5 7.3(1) 8.0 76.2 69.0 Maximum Temperature IL break, 45 feet reactor - 280.3 - 9.2 - 197.6 pool level (2) Minimum Temperature IL break, 13% initial power - 94.3 - 10.4 - 16.6 (2) Minimum Temperature
- 112.1 - 10.1 - 33.3 SGTF, 13% initial power (2)
Minimum Temperature DHRS cooldown, 13% - 116.8 - 10.4 - 39.2 initial power (2) (1) Minimum collapsed riser level was 2.8 feet and occurred approximately 3.6 hours after ECCS actuation. (2) A 12 hour transient simulation for these cases was not performed. Limiting conditions are only important at the end of the LTC phase at 72 hours.
© Copyright 20187 by NuScale Power, LLC 137
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 5.6.6 Summary and Conclusions Evaluation of LTC conditions following a variety of LOCA and non-LOCA initiating events was performed for three scenarios, maximum temperature, minimum temperature, and minimum level, which are challenging to different acceptance criteria. Detailed results from a subset of these events are provided in Section 5.6.1 through Section 5.6.4. The LTC acceptance criteria for cladding temperature, level above TAF, and margin to boron precipitation are satisfied for all cases. The following conclusions are drawn from the LTC analysis results.
- All maximum temperature cases showed decreasing cladding temperatures over the long term, with final cladding temperature remaining well below operating temperature at full power.
- Generally, the LOCA spectrum cases result in higher cladding temperatures than the non-LOCA cases by 12.5 hours.
- The maximum temperature cases are not challenging for collapsed level and boron precipitation.
- Sensitivity results indicate that including non_condensable gases inside containment is limiting for clad temperature.
- Sensitivity to reactor pool level of 45 feet was evaluated. While final clad temperature was 24 °F higher than at 12.5 hours, all acceptance criteria remained satisfied and overall module conditions were not significantly affected. The 55 feet pool initial level was generally applied in the LTC analysis since this is the credible condition at transient initiation.
- Section 5.6.5 demonstrates acceptable cladding temperatures after 72 hours for the maximum temperature scenario, including both the 45 feet and 55 feet reactor pool level cases.
- All minimum temperature cases showed margin to boron precipitation over the long term as core inlet temperature decreased.
- Generally, all LOCA and non-LOCA events converged towards the same temperatures and pressures after 12.5 hours.
- Collapsed level by 12.5 hours was most challenged by non-LOCA events in which RCS inventory is lost to the secondary system. Such cases were also limiting for boron precipitation after 12.5 hours due to higher boric acid concentration. All cases demonstrated margin to the collapsed level and boron precipitation criteria.
- Margin to boron precipitation was demonstrated for the limiting minimum temperature case after 72 hours, which included an initial reactor power of 13%
(Section 5.6.5). © Copyright 20187 by NuScale Power, LLC 138
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10
- All minimum level cases showed the core remained covered at all times during the transient and that boron precipitation driven by increased boric acid concentration with decreasing level is precluded.
- The LOCA injection line breaks were limiting for minimum level. The most limiting case is the 100% IL break with a minimum level of 2.8 feet above the TAF occurring approximately 3.6 hours after ECCS actuation.
- All cases show long term level recovery, where final level after 12.5 hours is higher than the transient minimum level.
- The timing of minimum boron precipitation margin generally occurred shortly after the timing of minimum level. Margin is reduced as boric acid concentration in the mixing volume increases.
- The non-LOCA SGTF cases were limiting for boron precipitation due the worst case combination of collapsed level and core inlet temperature.
- Section 5.6.5 demonstrates margin to collapsed level and boron precipitation after 72 hours for the minimum level scenario.
© Copyright 20187 by NuScale Power, LLC 139
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-37 Reactor coolant system pressure for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 140
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-38 Containment vessel pressure for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 141
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-39 Reactor coolant system collapsed liquid level above top of active fuel for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 142
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-40 Containment vessel level for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 143
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-41 Peak cladding temperature for maximum cooldown with steam generator tube failure © Copyright 20187 by NuScale Power, LLC 144
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 5-42 Flow through single reactor vent valve for maximum cooldown with steam generator tube failure 5.3.4 Summary and Conclusions Decay heat removal via ECCS, regardless of the short term initiating conditions and events, is acceptable compared to the regulatory requirements outlined in Section 2.2 and the acceptance criteria outlined in Section 2.3. The reduction of the long-term peak cladding temperature for all limiting conditions (minimum cooldown, maximum cooldown, pressurizer level, DHRS availability, and SGTF) compared to the initial temperature provides justification that CHF does not occur. Stable long-term collapsed liquid levels in the riser in conjunction with maintaining positive flow through the RVV provide assurance that core flow is continuous and positive during LTC. The limiting case for the minimum collapsed liquid level above the active fuel is produced by the maximum cooldown case with 1.2 times decay heat with the initial pressurizer level at a minimum of 20 percent. The minimum collapsed liquid level was found to be 2.271 feet above the TAF. The limiting case for maximum fuel cladding temperature is produced by the minimum cooldown case with the initial reactor pool level at 45 feet. © Copyright 20187 by NuScale Power, LLC 145
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 The boron precipitation methodology presented in Section 6.0 uses the RCS inlet core temperatures and the minimum RCS liquid level above the TAF to evaluate acceptance criteria for determining that a coolable geometry is maintained. © Copyright 20187 by NuScale Power, LLC 146
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 6.0 Boron Precipitation Methodology and Analysis Results The NPM uses boron for core reactivity control during normal operation. During the long-term cooling phase of ECCS operation, boiling in the core region is expected to concentrate boron in the liquid in the core and riser region. After ECCS valves open and recirculation is established, liquid from containment enters the RPV through the RRVs, circulates into the core region, and vapor is vented into containment through the RVVs where it condenses on the containment wall. Over time, the vapor venting from the RPV into containment will result in increased boron concentration in the RPV and decreased boron concentration in the fluid in containment. In the NPM design, the collapsed liquid level remains above the top of the core during the long-term cooling phase. Therefore, the concentration of the boron in the reactor vessel core and riser region is analyzed to demonstrate that boron precipitation does not occur and coolable geometry is maintained. 6.1 General Approach and Acceptance Criteria A simplified, conservative mixing volume approach is used to demonstrate that following an event that transitions to long-term ECCS cooling, the boron concentration of the liquid in the core and riser region remains below the solubility limit and therefore boron precipitation does not occur and coolable geometry is maintained. The mixing volume credited in the boron precipitation analysis is the liquid volume in the core and riser region, based on the collapsed liquid levelmass above the bottom of the core calculated by NRELAP5 in the long-term cooling calculations. The core inlet temperature predicted by NRELAP5 is compared to the precipitationsolubility temperature for boric acid as a function of concentration in the mixing volumethe calculated core and riser region collapsed liquid level. The maximum allowable boron concentration during operation is conservatively assumed. A simplified, conservative analysis is performed where it is assumed that the mass of boron initially in the RCS is locatedcompletely concentrated in the liquid in the core and riser region; liquid in containment, the downcomer, and the lower plenum are assumed to be entirely diluted. In reality, after the initial blowdown of liquid and vapor into containment, it would take time for the boron concentration in the core and riser region to increase due to vapor venting. This time-dependent transport of boron from liquid in the downcomer and containment is conservatively neglected. The boron solubility curve that specifies the acceptance criterion of allowable concentration of boric acid as a function of temperature is shown in Figure 6-1. © Copyright 20187 by NuScale Power, LLC 147
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 Figure 6-1 Percent boric acid at solubility limit as a function of temperature 6.2 Methodology The determination of the precipitation temperature for a given mixing volumethe liquid mass in the core and riser regions starts with the calculation of the entire mass of boron in the reactor coolant system (RCS). Then, a corresponding boric acid concentration is calculated for the mixing volume. Next, the precipitation temperature is obtained for the mixing volume concentration using the boron precipitation curve. Finally, core inlet temperature is compared to the precipitation temperature to determine margin to boron precipitation. First, tThese calculations are performed by control variables built into the NRELAP5 LTC model. This allows total boron mass and boric acid concentration in the mixing volume to be calculated based on case-specific conditions.for various mixing volumes corresponding to various elevations of liquid levels above the core. Finally, the level above the TAF, and the core inlet temperature, from the NRELAP5 long-term cooling calculation are used with the calculated solubility temperature to demonstrate that boron precipitation does not occur. 6.2.1 Calculate Total Boron Mass The maximum allowable boron concentration in the RCS during operation is conservatively assumed. The initial mass of water in the RCS is calculated on a case-specific basis.determined assuming the maximum pressurizer level. The initial RCS density is taken as the density at the vessel average temperature. This implicitly assumes that the RCS fluid volume at Tcold conditions is equal to the fluid volume at Thot conditions. This approximation is acceptable because, in the NuScale design, the limiting maximum allowable boron concentration is at hot zero power (HZP) conditions. At HZP conditions, © Copyright 20187 by NuScale Power, LLC 148
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 there is minimal temperature rise across the core, and therefore a uniform density is appropriate. Given an initial boron concentration, Crv, the mass of boron in the reactor vessel is
=
10 where,
= ,
= ,
= ,
6.2.2 Calculate Mass of Fluid in Mixing Volume The total liquid mass inside the core and riser regions is captured dynamically by the NRELAP5 LTC model.With the collapsed liquid level in the core and riser region from the inside bottom of vessel, the mixing volume is calculated based on the RCS geometry. The density of the collapsed liquid is 53.351 lbm/ft3 at HZP conditions. With the density (l,mv) and volumes (Vmv) known, the mass of the mixing volume can be solved using the following equation.
= ,
The mass of the mixing volume changes with the height of the collapsed liquid level. 6.2.3 Calculate Boron Concentration in Mixing Volume The mixing volume boron and boric acid concentrations are expressed in ppm as 10
=
+
=
© Copyright 20187 by NuScale Power, LLC 149
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 where and are the molecular weights of boric acid and boron, respectively. and are the concentration of boron and boric acid, respectively. The molecular weights are 61.8 g/mol for boric acid and 10.8 g/mol for boron. The boric acid concentration in weight percent (wt%) is expressed as
% = 10 =
10 10 6.2.4 Assess Margin to Boron Precipitation Finally, the level above the TAF, and the core inlet temperature, from the NRELAP5 long-term cooling calculation are used with the calculated solubility temperature to demonstrate that boron precipitation does not occur. The precipitation temperature as a function of boric acid concentration is compared to the core inlet temperature. Margin is demonstrated by ensuring core inlet temperature remains greater than the precipitation temperature. The NRELAP5 calculations demonstrate that the minimum collapsed level in the core and riser region occurs relatively early in the transient following ECCS valve opening (within approximately 1-2 hours in the letdown break calculations). Longer term, the RCS and containment levels equilibrate and the core and riser level increases from the minimum. At the time of minimum core and riser level, the core inlet temperature remains fairly high, and decreases to a quasi-steady condition at the end of the calculation. Therefore, two points are considered in the boron precipitation analysis: the point of minimum level, and the calculation end point where minimum temperature occurs. 6.3 Results Table 6-1 shows the temperature at which boron will precipitate for mixing volumes encompassing the core and the riser fluid corresponding to the given fluid levels above TAF. These assume the HZP conditions of 1800 ppm boron concentration, 420 degrees F RCS average temperature, and 1850 psia RCS pressure. Based on the results shown in Section 5.0, collapsed liquid levels below 1foot above TAF are not reached, and have not been presented below. Table 6-1 Boron precipitation temperatures for various collapsed liquid levels above top of active fuel ((
}}2(a),(c) © Copyright 20187 by NuScale Power, LLC 150
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 ((
}}2(a),(c)
From tThe results for the LTC analysis cases considered in the boron precipitation analysis are given in Table 5-2 in Section 5.6.5., Using the input from the boron solubility curve in Figure 6-1, NRELAP5 predicted that the case which produced the minimum margin of 16.6 degrees F to boron precipitation was a 13% power IL break case biased for minimum temperature at 72 hours.the lowest level reached was 2.271 feet above TAF in a LDBRK case biased for maximum cooldown, 1.2 multiplier on the decay heat, failure of one RRV and one RVV, and minimum initial pressurizer level of 20 percent. Full power cases with lower decay heat were considered in the boron precipitation analysis and were found to
© Copyright 20187 by NuScale Power, LLC 151
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 be non-limiting due to minimal impact on core inlet temperatures offset by higher long-term level. The core conditions for this case showed a core inlet and core outlet temperature of approximately 185 degrees F at about 2.3 hours after break initiation. Using the Table 6-1 entry at 2.25 feet shows the critical boron solubility temperature is 178 degrees F. From this, boron precipitation will not occur during the level-limiting LTC event. Margin to the solubility temperature at the end of the cooling calculation was demonstrated for all cases considered in the boron precipitation analysis. For example, the core inlet temperature at 72 hours is 157 degrees F for a LDBRK case biased for maximum cooldown, 1.2 multiplier on decay heat, failure of one RRV and one RVV. At this temperature, the given case showed a collapsed liquid level of 8.50 feet above the TAF. Using , a temperature of 107 degrees F would initiate boron precipitation if the collapsed liquid level were 8.5 feet above TAF. Therefore, adequate margin to the solubility temperature is demonstrated. Cases with lower decay heat were also considered in the boron precipitation analysis and were not limiting due to minimal impact on core inlet temperatures and higher long-term level. 6.4 Conclusions Based on the results shown above, boron precipitation will not occur during any postulated condition in a long-term cooling scenario. Sufficient margin exists both to cover the low levels reached immediately following ECCS activation as well as the low temperatures reached during extended ECCS operation assuming conservative plant conditions. © Copyright 20187 by NuScale Power, LLC 152
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 7.0 Summary and Conclusions This report documents the analytical methodology for long-term ECCS operation, either as an extension to a LOCA, or as a result of ECCS activation following a non-LOCA event when loss of normal AC power is assumed. The applicable regulatory requirements from 10CFR50.46, NuScale PDC 35, and the regulatory guidance from the NuScale DSRS have been addressed by the long-term cooling methodology. This methodology utilizes the NuScale LOCA EM described in Reference 8.2.1, and was developed in accordance with to RG 1.203. The LTC methodology is informed by comprehensive work with NRELAP5 parametric calculations, exploring an extensive set of sensitivities for effect on the FOMs as defined in the PIRT. These sensitivities used appropriate ranges of controlling parameters. The LTC methodology includes the evaluation of margin to boron precipitation. Bounding evaluations were performed with a limiting set of assumptions and initial conditions based on sensitivity results. The cases identified as most limiting, the minimum cooldown, maximum cooldown, and SGTF with DHRS, demonstrated that the collapsed liquid level remains above the TAF with acceptably low RCS and cladding temperatures, showing that the ECCS capability to provide core cooling for an extended period is adequate. In addition, boron precipitation was evaluated and it was demonstrated not to occur for the range of conditions evaluated for long-term cooling, thereby demonstrating that the core remains in a coolable geometry. © Copyright 20187 by NuScale Power, LLC 153
Long-Term Cooling Methodology TR-0916-51299-NP Draft Rev. 10 8.0 References 8.1 Source Documents 8.1.1 American Society of Mechanical Engineers, Quality Assurance Program Requirements for Nuclear Facility Applications, ASME NQA-1-2008, ASME NQA-1a-2009 Addenda, as endorsed by Regulatory Guide 1.28, Revision 4. 8.1.2 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Title 10, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, (10 CFR 50 Appendix B). 8.1.3 NuScale Topical Report, NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, NP-TR-1010-859-NP-A, Revision 3. 8.2 Referenced Documents 8.2.1 NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422, Revision 0. 8.2.2 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities. Part 50, Title 10, Section 50.46, Acceptance Criteria for Emergency Core Cooling System for Light-Water Nuclear Power Reactors," (10 CFR 50.46). 8.2.3 U.S. Nuclear Regulatory Commission, Emergency Core Cooling System, Design-Specific Review Standard for NuScale SMR Design, Section 6.3, Revision 0, June 2016. 8.2.4 U.S. Nuclear Regulatory Commission, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, Design-Specific Review Standard for NuScale SMR Design, Section 15.6.5, Revision 0, June 2016. 8.2.5 NuScale Topical Report, Non-LOCA Methodologies, TR-0516-49416, Revision 0. 8.2.6 U.S. Nuclear Regulatory Commission, Transient and Accident Analysis Methods, Regulatory Guide 1.203, Revision 0, December 2005. 8.2.7 ISA, ANSI/ISA 75.01.01-2007, Flow Equations for Sizing Control Valves. 8.2.8 SwUM-0304-17023, Revision 46, NRELAP5 Version 1.34 Theory Manual. © Copyright 20187 by NuScale Power, LLC 154
RAIO-1218-63931 : Affidavit of Zackary W. Rad, AF-1218-63932 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 long term cooling 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-1218-63932
- 4. The information sought to be withheld is in the enclosed response to NRC Request for Additional Information No. 483, eRAI No. 9516. 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 December 21, 2018. Zackary W. Rad AF-1218-63932}}