ML23292A311
| ML23292A311 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/19/2023 |
| From: | Harshaw K South Texas |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NOC-AE-23003990, EPID L-2023-LLA-0047 | |
| Download: ML23292A311 (1) | |
Text
October 19, 2023 NOC-AE-23003990 10 CFR 50.90 STI: 35519294 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 South Texas Project Unit 1 & 2 Docket Nos. STN 50-498 and STN 50-499 Response to Request for Additional Information for License Amendment Request to Revise Alternate Source Term Dose Calculation (EPID: L-2023-LLA-0047)
March 30, 2023; (NOC-AE-23003941) (ML23089A204).
(AE-NOC-23003381) (ML23264A097).
By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a license amendment request for South Texas Project Units 1 and 2 to the U.S. Nuclear Regulatory Commission (NRC). The proposed amendment would authorize revision of the alternate source term dose calculation for the main steam line break (MSLB) and the locked rotor accident (LRA). The reanalysis uses the asymmetric natural circulation cooldown thermal hydraulic analyses, various radiation transport assumptions, and the current licensing basis source term and meteorological data to evaluate the dose effects of an extended cooldown on the existing accident analyses.
By Reference 2, the NRC notified STPNOC that additional information was needed for the staff to complete its review.
The enclosure to this letter provides the STPNOC response to the NRC Request for Additional Information.
There are no commitments in this letter.
If there are any questions or if additional information is needed, please contact Christopher Warren at (361) 972-7293 or me at (361) 972-4778.
~.....
Nuclear Operating Company South Texas Project Electric Generating Station P.O Box 28!) midsworth, Texas 77483
References:
- 1) Letter; K. Harshaw (STP) to Document Control Desk (NRC); "License Amendment Request to Revise Alternative Source Term Dose Calculation;"
- 2) Email; D. Galvin (NRC) to W. Brost (STP); "South Texas Project - Request for Additional Information - License Amendment Request to Revise the Alternate Source Term Dose Calculation (L-2023-LLA-0047);" September 21, 2023;
NOC-AE-23003990 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on _________________
Kimberly A. Harshaw Executive VP and CNO
Enclosure:
Response to Request for Additional Information for License Amendment Request to Revise Alternate Source Term Dose Calculation cc:
Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-451 Dennis Galvin Project Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 4 Sean Lichvar Acting Senior Resident Inspector, South Texas Project U.S. Nuclear Regulatory Commission Leanne Flores Resident Inspector, South Texas Project U.S. Nuclear Regulatory Commission Robert Free, Texas Department of State Health Services
NOC-AE-23003990 Enclosure Page 1 of 5 Response to Request for Additional Information for License Amendment Request to Revise Alternate Source Term Dose Calculation By letter dated March 30, 2023 (ADAMS Accession No. ML23089A204), the STP Nuclear Operating Company (STPNOC, the licensee) submitted a license amendment request (LAR) for South Texas Project (STP), Units 1 and 2 to the U.S. Nuclear Regulatory Commission (NRC).
The proposed amendments would authorize revision of the alternate source term dose calculation for the main steam line break (MSLB) and the locked rotor accident (LRA). The reanalysis uses the asymmetric natural circulation cooldown thermal hydraulic analyses, various radiation transport assumptions, and the current licensing basis source term and meteorological data to evaluate the dose effects of an extended cooldown on the existing accident analyses.
Based on a review of the proposed MSLB and LRA asymmetric natural circulation cooldown (ANCC) analysis in Sections 2.3, 3.2.1, and 3.3.1 of the LAR enclosure, respectively, the NRC staff has identified the following requests for additional information (RAIs).
SNSB-RAI 1 Regulatory Basis 10 CFR 50, Appendix A, GDC 10, as it relates to the reactor coolant system (RCS) being designed with appropriate margin to ensure that the specified acceptable fuel design limits are not exceeded during normal operations including anticipated operational occurrences.
RAI Section 2.3 of the LAR enclosure states:
The MSLB and LRA events are impacted by this change because these events and the assumptions imposed by the analysis require cooling down an intact RCS without all four steam generators during a coincident LOOP [loss-of-offsite power]. Other accidents such as LOCA [loss-of-coolant accident], Steam Generator Tube Rupture, and Rod Ejection all involve some type of break in the RCS, their cooldown timeline is not impacted and they are not re-analyzed.
(a) The LAR indicates that if one or more steam generators were not available for RCS cooldown, an ANCC issue that resulted in a stagnant loop conditions would have an impact on both the thermal-hydraulic analysis and the dose analysis. Sections 3.2.1 and 3.3.1 present assumptions to be used in updated MSLB and LRA ANCC RETRAN analyses, respectively, but the LAR does not explain why these assumptions are suitably conservative or what controls will be put in place to implement these assumptions. For example, a cooldown rate of 15 oF/hr is assumed to prevent stagnation. However, the LAR does not explain why a higher or lower cool down rate or permitting stagnation would not be more conservative, or if controls are in place such that other assumptions do not need to be considered. Explain how the proposed assumptions are suitably conservative or are otherwise appropriate.
NOC-AE-23003990 Enclosure Page 2 of 5 STP Response RAI 1(a):
The RETRAN analyses in the LAR for MSLB and LRA under ANCC conditions are based on STP Calculations NC07143 and NC07123. Calculations NC07143 and NC07123 only provide the cooldown timeline and steam release input for the dose consequence analysis (NC06026 and NC06028). These two calculations supersede the previous dose input Calculation NC07124, in which steam release is calculated based on the heat balance between decay heat load and heat dissipation for an 8-hour release period.
The RETRAN analyses listed in this LAR is not intended to replace the current license basis thermal hydraulic analysis for design basis accident of MSLB and LRA. The current thermal hydraulic analysis listed in UFSAR 15.1.5 and 15.3.3 has adequately addressed the core response and DNBR concerns associated with 10 CFR 50, Appendix A, GDC 10.
The assumptions listed in LAR Sections 3.2.1 and 3.3.1 for the RETRAN analysis are intended to get a conservative cooldown timeline and steam release input for the dose analyses. The detailed bases for those assumptions are presented in Section 5 of STP Calculations NC07143 and NC07123.
As indicated in LAR Section 2.3, loop flow stagnation could occur during ANCC with higher cooldown rates for the event if one or more steam generators were not available for cooling.
To avoid RCS loop flow stagnation, the cooldown rates need to be reduced in an ANCC situation per WCAP-16632-P, which results in an extended cooldown timeline. For STP, the cooldown rate is procedurally limited to 15 °F/hr to 20 °F/hr to avoid loop flow stagnation in (NC07081) at ANCC condition and controlled by the emergency operating procedure for natural circulation.
The RETRAN analyses in the LAR for MSLB and LRA under ANCC conditions conservatively assume the plant cools down at the minimum procedurally allowed rate for the entirety of the cooldown process. With a reduced cooldown rate to 15 °F/hr, it takes approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> (MSLB and LRA with limiting single failure) to reach RHR cut-in conditions. Therefore, controls are in place such that other assumptions do not need to be considered.
(b) LAR Section 2.3 indicates that the non-LOCA events described in Updated Final Safety Analysis Report Sections 15.1 through 15.6, other than MSLB and LRA, are not impacted by the ANCC condition resulting from inactive steam generator(s) caused by the assumed limiting single failure in their analysis without providing a justification. Provide justifications that the other events are not impacted by the ANCC condition.
As indicated in Section 2.3, other accidents such as LOCA, Steam Generator Tube Rupture (SGTR), and Rod Ejection all involve some type of break in the RCS, which is the major contributor to the overall dose in dose consequence analysis.
For LOCA, the ANCC condition is not applicable since natural circulation cooldown via the event all SG's are not intact and active. This is supported by STP specific analysis
NOC-AE-23003990 Enclosure Page 3 of 5 steam generators is not established due to the break of an RCS leg. As indicated in the LOCA dose consequence analysis (NC06013), the core damage source term is simultaneously released from the RCS to the sprayed and unsprayed portions of the containment and to the containment sump then released out through the plant vent or containment leakage. The cooldown timeline is not an input for the RADTRAD model.
In the Rod Ejection dose consequence analysis (NC06014, UFSAR Section 15.4.8), there are two pathways modeled separately using RADTRAD. One pathway is released from the reactor to the containment atmosphere directly. Another path is from the Reactor to the RCS, then released to the secondary side through primary to secondary leakage. The leakage is stopped at 4500 seconds when the primary and secondary system pressures are equalized. The cooldown timeline has no impact on the releases from either pathway.
Calculation NC06034 (UFSAR Section 15.6.3) performed the dose consequence analysis for SGTR. In the RADTRAD model, there are three release pathways to the environment.
The first pathway is the break from RCS to the secondary side. A portion of the flow through the break flashes to steam and is assumed to go directly into the environment. The break flow terminates at approximately 5000 seconds when the primary and secondary system pressures are equalized. The other pathways are from RCS to intact/ruptured SGs. The total assumed leakage rate is 1 gpm. The leakage mixes into the secondary portion of the ruptured and intact steam generator. PORV iodine releases from the ruptured and intact SGs are partitioned by a factor of 100 between the liquid and steam phase. This is modeled by reducing the flows out of the SG nodes by a factor of 100. Considering that the primary to secondary leakage rate is low as well as the partition effect, the release from intact/ruptured SGs is a minor contributor to the overall dose. The break flow flashing is the major contributor to the overall dose since it is assumed to go directly into the environment. The cooldown timeline has negligible impact on SGTR dose consequence analysis since the break flow flashing is the major contributor to the overall dose.
In summary, the break in the RCS is the main contributor to dose consequences for each of these three events. For both LOCA and Rod Ejection, the ANCC condition is not applicable and the cooldown timeline is not an input on the dose consequences. For SGTR, the break flow ends early in the accident independently of the cooldown timeline.
SNSB-RAI 2 Regulatory Basis:
10 CFR 50.67, Accident source term, as it relates to the implementation of an alternate source term in current operating nuclear power plants.
RAI Sections 3.2.1 and 3.3.1 of the LAR enclosure provide assumptions used in the proposed MSLB and LRA ANCC analyses, respectively. The assumptions are further discussed in LAR, "Dose Analysis Calculation and Code Output Files," dose analysis calculation files NC07123 RO, "Locked Rotor Accident Steam Release for Dose Analysis," and NC07143
NOC-AE-23003990 Enclosure Page 4 of 5 performed by hand calculations while the proposed analyses are performed using RETRAN; however, the LAR does not provide a comparison of the assumptions between the proposed analyses and the corresponding analyses of record. Identify if there are any differences in assumptions and the analysis inputs between the proposed analyses and the corresponding analyses of record. For the differences (if any), provide justification in case the conservatism is reduced.
The steam release for dose consequence analysis in the current license bases (Calculation NC07124) is based on all the heat energy being removed via steaming in the steam generators.
It is an energy balance hand calculation and the methodology was based on the guidance of considers heat generated in the core, heat released or absorbed by thick metal in the RCS and intact steam generators, and heat released or absorbed within the fluids in the RCS and intact steam generators. The energy that cannot be stored within the RCS and intact steam generators is removed via steaming from steam generators.
In the energy balance steam release calculation, the assumption that a total of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of steam release occurs prior to placing the plant in the RHR mode of operation is imposed on the results. For the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, it is assumed the plant will have cooled down and stabilized at no-load conditions. The additional time from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is to cooldown and depressurize the plant from no-load conditions to the RHR operating conditions.
Under ANCC conditions, the 8-hour cooldown timeline to the RHR cut in condition is not valid anymore for MSLB and LRA due to the reduced cooldown rate. Therefore, the RETRAN models were developed with the dual purpose to determine the actual RHR cut-in time and steam release rates under the ANCC condition using the conservative assumptions (NC07143 and NC07123).
Differing from the heat balance calculation, the RETRAN models at the ANCC condition conservatively assume a 2-hour plant stabilization time and a 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> boration time (Calculation NC07143 and NC07123). The cooldown starts at 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. With the lowest procedural cooldown rate of 15 oF/hr under the ANCC condition, it takes a total of approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> to reach RHR cut-in conditions. Per Section I.C of SAS 22, there are no specific acceptance criteria associated with the calculation of the steam releases used as an input to the radiological dose analysis. All of the conservative assumptions made in the RETRAN models for an ANCC condition are intended to obtain a longer cooldown timeline and larger steam release, which will result in a higher dose consequence. The RETRAN models use the major inputs from SAS 22 as listed in Table 4 of Calculation NC07143 for MSLB and Table 3 of Calculation NC07123 for LRA.
Table 1 below compares the current analysis of record to the ANCC RETRAN model.
R2, "MSLB Steam Release for Dose Analysis. 11 The dose calculation files each indicate in Section 4.2, "Calculation Method, 11 that previous analyses (i.e., the analyses of record) were Safety Analysis Standard (SAS) 22, "Steam Release for Dose Evaluation." The energy balance
NOC-AE-23003990 Enclosure Page 5 of 5 Table 1: Comparison of Initial Condition Assumptions Between Current Analysis and ANCC RETRAN Analyses Assumption Current Analysis ANCC RETRAN Analysis Justification Initial Power Level 102% original NSSS Power 3817 Mw (no 1.4% Power Uprate)
Core Power +
Uncertainty Rated core power level 3853 Mw (with 1.4%
power uprate) with Power Uncertainty of 0.6%.
Current analyses made at 102% rated power are equivalent to 1.4% Power Uprate analyses made at 101.4% original rated power
+ 0.6% Power Uncertainty for RETRAN analysis. The RCP pump heat is not included in RETRAN analysis since there is no RCP operation under ANCC condition.
RCS T Avg.
593 oF 593 oF No Change Steam Temperature Nominal steam temperature 556.3 oF This is not a direct input for RETRAN analysis.
The initial nominal steam generator pressure is 1077.46 psia.
The saturation temperature based on 1077.46 psia for RETRAN analysis is 553.7 oF. The slight initial steam temperature difference has negligible impact on the long-term steam release under ANCC conditions.
RCS Pressure Nominal RCS pressure of 2250 psia Nominal RCS pressure of 2250 psia No change SG tube plugging 0%
0%
No Change SG water level This is not a direct input for current analysis. The initial SG water mass is based on the nominal water level.
Nominal water level of 71.7% NRS.
No change