ML18199A193
| ML18199A193 | |
| Person / Time | |
|---|---|
| Site: | PROJ0769 |
| Issue date: | 07/18/2018 |
| From: | NRC |
| To: | NRC/NRO/DLSE/LB1 |
| References | |
| Download: ML18199A193 (4) | |
Text
NuScaleTRRaisPEm Resource From: Cranston, Gregory Sent: Wednesday, July 18, 2018 8:25 AM To: Request for Additional Information Cc: Lee, Samuel; Karas, Rebecca; Thurston, Carl; Franovich, Rani; Chowdhury, Prosanta; NuScaleTRRaisPEm Resource
Subject:
Request for Additional Information Letter No. 9492 (eRAI No. 9492) Topical Report LOCA 15.6.5, SRSB Attachments: Request for Additional Information No. 9492 (eRAI No. 9492).pdf Attached please find NRC staffs request for additional information (RAI) concerning review of the NuScale Topical Report.
Please submit your technically correct and complete response within 60 days of the date of this RAI to the NRC Document Control Desk.
If you have any questions, please contact me.
Thank you.
1
Hearing Identifier: NuScale_SMR_DC_TR_Public Email Number: 99 Mail Envelope Properties (BN1PR09MB0258496CD0F229B7BD2C22C190530)
Subject:
Request for Additional Information Letter No. 9492 (eRAI No. 9492) Topical Report LOCA 15.6.5, SRSB Sent Date: 7/18/2018 8:24:47 AM Received Date: 7/18/2018 8:24:52 AM From: Cranston, Gregory Created By: Gregory.Cranston@nrc.gov Recipients:
"Lee, Samuel" <Samuel.Lee@nrc.gov>
Tracking Status: None "Karas, Rebecca" <Rebecca.Karas@nrc.gov>
Tracking Status: None "Thurston, Carl" <Carl.Thurston@nrc.gov>
Tracking Status: None "Franovich, Rani" <Rani.Franovich@nrc.gov>
Tracking Status: None "Chowdhury, Prosanta" <Prosanta.Chowdhury@nrc.gov>
Tracking Status: None "NuScaleTRRaisPEm Resource" <NuScaleTRRaisPEm.Resource@nrc.gov>
Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>
Tracking Status: None Post Office: BN1PR09MB0258.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 345 7/18/2018 8:24:52 AM Request for Additional Information No. 9492 (eRAI No. 9492).pdf 46397 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
Request for Additional Information No. 9492 (eRAI 9492)
Issue Date: 07/18/2018 Application
Title:
NuScale Topical Report Operating Company: NuScale Docket No. PROJ0769 Review Section: 15.06.05 - Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Application Section: 15.6.5 QUESTIONS 15.06.05-21 Title 10 of the Code of Federal Regulations (10 CFR) Part 50.46 (a)(1)(ii) states, "the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded." Regulatory Guide 1.203 describes a process that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for use in developing and assessing evaluation models (EMs) that may be used to analyze transient and accident behavior that is within the design basis of a nuclear power plant.
As stated in RG 1.203, an EM is the calculational framework for evaluating the behavior of the reactor system during a postulated transient or design-basis accident. As such, the EM may include one or more computer programs, special models, and all other information needed to apply the calculational framework to a specific event, as illustrated by the following examples: (1) Procedures for treating the input and output information (particularly the code input arising from the plant geometry and the assumed plant state at transient initiation), (2)
Specification of those portions of the analysis not included in the computer programs for which alternative approaches are used, and (3) All other information needed to specify the calculational procedure.
The entirety of an EM ultimately determines whether the results are in compliance with applicable regulations.
Therefore, the development, assessment, and review processes must consider the entire EM.
The validation of the loss of coolant accident (LOCA) methodology depends heavily upon benchmarks of experimental data to confirm the performance of the codes and methods over the applicable range of reactor system operating conditions. Integral and separate effects data for the NuScale Power Module are based on the NIST-1 facility data presented in Section 7 of the LOCA topical report (LTR) (TR-0516-49422-P, Rev. 0). However, NRC staff has determined that comparisons of addition data are necessary to find that the test results and NRELAP5 benchmarks validate the applicant's EM.
The applicant is requested to provide assessment plots and discussions from the following tests:
HP-02 (EC-T080-3822-R1):
Initial and boundary condition plots for RUN3, similar to the plots for RUN 1 (Figures 4-1 to 4-4, Figures 4-20 to 4-22 in the report). HTP temperature (TW-5XXXX) plot at various elevations for RUN 3. The TW-5XXX measurements capture heat transfer effects through the heat transfer plate and are important to assess and understand the overall heat transfer phenomena. RUN 3 is chosen for its highest CNV operating pressure that is closest to NIST-1 integral tests.
HP-05 plots along with NRELAP5 results:
DP-1101C (core DP), DP-1501C (SG DP) ,FCM-2201 (FW rate) , PT-2501a (Steam pressure). and core_pwr (core power). These are important to gauge the performance of NRELAP pressure drop calculation.
HP-06b (EC-T080-4872-R0):
Figure 4-23 RPV pressure, mid-term.
Figure 4-25, RPV level, mid-term.
Figure 4-26 secondary side pressure, mid-term.
Figure 4-27 CNV pressure, mid-term.
Figure 4-28, CNV level, mid-term.
Figure 4-29, 4-31, and 4-33 CNV fluid temp, mid-term.
Figure 4-35, 4-37, and 4-39 CNV wall temp, mid-term.
Figure 4-40, 4-42, and 4-44 (HTP temp), mid-term.
Figure 4-46, 4-47, and 4-48 (CPV fluid temp), mid-term.
(Staff noted most plots for the CVCS line break in the LTR are from HP-06. More mid-term HP-06b plots are needed to supplement the CVCS line break results, since it reflects more realistic decay power history.)
HP-43: EC-T080-5045_R0:
Figure 4-1 HP-43 pre-break CNV wall temperature TW-4X01~04 Figure 4-2 HP-43 pre-break HTP wall temperature TW-50XY to TW-59XY Figure 4-14 PZR pressure PT-1401a, short-term.
Figure 4-15 PZR level Ldp1401Cal, short-term.
Figure 4-16 RPV level, short-term.
Figure 4-19 PT-2326, 2334, 2501 secondary side pressure, short-term.
Figure 4-20 FVM-3221 RVV orifice mass flow 200s, short-term.
Figure 4-21 DP-3221, short-term.
Figure 4-22 CNV pressure, short-term.
Figure 4-23 CNV level, short-term.
Figure 4-24 PZR pressure, mid-term.
Figure 4-26 RPV level, mid-term.
Figure 4-28 CNV pressure, mid-term.
Figure 4-29 CNV level, mid-term.
Figure 4-67 "Cd Cases" on FVM-3221, 200s, short-term.
(Since HP-43 is the only NIST-1 test simulating the 3-RVV configuration.)