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| {{#Wiki_filter: 1 NRR-PMDAPEm Resource From:Harrison Albon <awharrison@STPEGS.COM> | | {{#Wiki_filter:NRR-PMDAPEm Resource From: Harrison Albon <awharrison@STPEGS.COM> |
| Sent: Tuesday, September 06, 2016 3:52 PM To: Regner, Lisa | | Sent: Tuesday, September 06, 2016 3:52 PM To: Regner, Lisa |
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| [External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Attachments:SNPB-3-02 followup - partial 2 Bullet 4.pdf Lisa, Here is another partial for discussion in the public call on 9/14. | | [External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Attachments: SNPB-3-02 followup - partial 2 Bullet 4.pdf |
| | : Lisa, Here is another partial for discussion in the public call on 9/14. |
| | Wayne 1 |
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| | Hearing Identifier: NRR_PMDA Email Number: 3064 Mail Envelope Properties (FC3FC22E-DF52-40BF-806A-7633810123B7) |
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| Hearing Identifier: NRR_PMDA Email Number: 3064 Mail Envelope Properties (FC3FC22E-DF52-40BF-806A-7633810123B7) | |
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| ==Subject:== | | ==Subject:== |
| [External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Sent Date: 9/6/2016 3:51:58 PM Received Date: 9/6/2016 3:52:04 PM From: Harrison Albon Created By: awharrison@STPEGS.COM Recipients: "Regner, Lisa" <Lisa.Regner@nrc.gov> Tracking Status: None | | [External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Sent Date: 9/6/2016 3:51:58 PM Received Date: 9/6/2016 3:52:04 PM From: Harrison Albon Created By: awharrison@STPEGS.COM Recipients: |
| | "Regner, Lisa" <Lisa.Regner@nrc.gov> |
| | Tracking Status: None Post Office: stpegs.com Files Size Date & Time MESSAGE 88 9/6/2016 3:52:04 PM SNPB-3-02 followup - partial 2 Bullet 4.pdf 397348 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: |
| | Recipients Received: |
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| Post Office: stpegs.com Files Size Date & Time MESSAGE 88 9/6/2016 3:52:04 PM SNPB-3-02 followup - partial 2 Bullet 4.pdf 397348
| | DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16) |
| | : 1. Follow-up SNPB-3-2 Accident Scenario Progression Initial RAI: Provide a description of the accident progression of the accident scenarios being simulated using the LTCC EM. This description should start at the initiation of the break, define each phase, and the important phenomena occurring in that phase in the various locations of the RCS (e.g., core, reactor vessel, steam generators - both primary and secondary side, loops, pressurizer, pumps, containment) |
| | Though only the 16-inch hot leg break analysis was provided, STPNOC did not justify or demonstrate that the 16 inch hot leg break bounds smaller hot leg breaks. |
| | Justify that the 16-inch hot leg break analysis bounds the smaller hot leg break scenarios. |
| | STP Response The 16-inch HLB LOCA scenario progression is described as part of the response to the SNPB-3-02. The accident progression simulated with the LTCC EM can be summarized as follow: |
| | Immediately after the break event, the primary system experiences a fast depressurization where water and steam are discharged from the break. |
| | Reactor scram and ECCS actuation occur, controlled by low pressure signal. |
| | Due to the location of the break, injection of liquid water from the cold side is sustained by the ECCS injection system (HHSI, Accumulators, and LHSI). |
| | Core liquid inventory increases rapidly after the actuation of the accumulators and SI pump. Pre-Core Blockage LTCC starts when the core is fully flooded (Core Collapsed Liquid Level = 14 feet). |
| | When RWST water is depleted, injection start from the sump (SSO time) |
| | The total liquid inventory in the core and in the primary system is sufficient to supply the liquid water to the core immediately after the core blockage event. |
| | Liquid inventory in the core decreases until flow through alternative flow paths (steam generators u-tubes) is established. |
| | Liquid water flow is maintained at the top of the core through the steam generators during the Post-Core Blockage LTCC. |
| | Different important thermal-hydraulic parameters can be identified from the description provided. These parameters (listed and described below) can be used as figure of merit to confirm that the large break scenarios bound smaller hot leg breaks. |
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| Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:
| | DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16) |
| DRAFT Response to SNPB Follow-up RAI 3-(9/6/16) 1. Follow-up SNPB-3-2 Accident Scenario Progression Initial RAI: Provide a description of the accident progression of the accident scenarios being simulated using the LTCC EM. This description should start at the initiation of the break, define each phase, and the important phenomena occurring in that phase in the various locations of the RCS (e.g., core, reactor vessel, steam generators - both primary and secondary side, loops, pressurizer, pumps, containment) | | Sump Switchover time The sump switchover (SSO) time and, subsequently, the core blockage time, is one of the most important parameters defining the thermal-hydraulic initial conditions of the reactor system for the post-core blockage LTCC phase. The SSO time is defined based on the RWST volume available for injection, and the injection rate of the SI system and CS system. The rate of injection has three1 contributors: |
| Though only the 16-inch hot leg break analysis was provided, STPNOC did not justify or demonstrate that the 16 inch hot leg break bounds smaller hot leg breaks. Justify that the 16-inch hot leg break analysis bounds the smaller hot leg break scenarios.
| | The High Head Safety Injection (HHSI) pumps The Low Head Safety Injection (LHSI) pumps The Containment Spray (CS) pumps The rate of injection of the HHSI and LHSI depends on the pressure of the primary system and, based on the pump characteristics, is expected to increase at lower primary system pressures. The contribution of the CS pumps may be assumed to be constant throughout the accident progression2. The primary pressure during the accident progression is strongly dependent on the break size. |
| STP Response The 16-inch HLB LOCA scenario progression is described as part of the response to the SNPB-3-02. The accident progression simulated with the LTCC EM can be summarized as follow: Immediately after the break event, the primary system experiences a fast depressurization where water and steam are discharged from the break. Reactor scram and ECCS actuation occur, controlled by low pressure signal. Due to the location of the break, injection of liquid water from the cold side is sustained by the ECCS injection system (HHSI, Accumulators, and LHSI). Core liquid inventory increases rapidly after the actuation of the accumulators and SI pump. Pre-Core Blockage LTCC starts when the core is fully flooded (Core Collapsed Liquid Level = 14 feet). When RWST water is depleted, injection start from the sump (SSO time) The total liquid inventory in the core and in the primary system is sufficient to supply the liquid water to the core immediately after the core blockage event. Liquid inventory in the core decreases until flow through alternative flow paths (steam generators u-tubes) is established. Liquid water flow is maintained at the top of the core through the steam generators during the Post-Core Blockage LTCC.
| | In particular, larger breaks are expected to experience a faster depressurization, stabilizing the primary system at a lower pressure. Subsequently, the total ECCS injection is expected to be higher during large break scenario than smaller breaks. |
| Different important thermal-hydraulic parameters can be identified from the description provided. These parameters (listed and described below) can be used as figure of merit to confirm that the large break scenarios bound smaller hot leg breaks.
| | The SSO time for the three HLB scenarios simulated is reported in the table below. |
| DRAFT Response to SNPB Follow-up RAI 3-(9/6/16) Sump Switchover time The sump switchover (SSO) time and, subsequently, the core blockage time, is one of the most important parameters defining the thermal-hydraulic initial conditions of the reactor system for the post-core blockage LTCC phase. The SSO time is defined based on the RWST volume available for injection, and the injection rate of the SI system and CS system. The rate of injection has three 1 contributors: The High Head Safety Injection (HHSI) pumps The Low Head Safety Injection (LHSI) pumps The Containment Spray (CS) pumps The rate of injection of the HHSI and LHSI depends on the pressure of the primary system and, based on the pump characteristics, is expected to increase at lower primary system pressures. The contribution of the CS pumps may be assumed to be constant throughout the accident progression
| | Break Size 16" 6" 2" SSO Time (s) 1740 2302 3729 Decay Power at Core Blockage and Core Boil-off Rate The LTCC EM calculates the decay power using the ANS-1979 model. Based on the considerations on the SSO time provided in the previous paragraph, the decay power at time of core blockage is higher for larger breaks since the core blockage event occurs earlier in the transient. The amount of water required to compensate the boil-off rate and cool down the reactor core during at the post-core blockage LTCC is relatively larger in the 16-inch HLB scenario than smaller break scenarios. |
| : 2. The primary pressure during the accident progression is strongly dependent on the break size. In particular, larger breaks are expected to experience a faster depressurization, stabilizing the primary system at a lower pressure. Subsequently, the total ECCS injection is expected to be higher during large break scenario than smaller breaks. The SSO time for the three HLB scenarios simulated is reported in the table below. Decay Power at Core Blockage and Core Boil-off Rate The LTCC EM calculates the decay power using the ANS-1979 model. Based on the considerations on the SSO time provided in the previous paragraph, the decay power at time of core blockage is higher for larger breaks since the core blockage event occurs earlier in the transient. The amount of water required to compensate the boil-off rate and cool down the reactor core during at the post-core blockage LTCC is relatively larger in the 16-inch HLB scenario than smaller break scenarios. The core decay power at the time of core blockage for the three HLB scenarios simulated is reported in the table below.
| | The core decay power at the time of core blockage for the three HLB scenarios simulated is reported in the table below. |
| | Break Size 16" 6" 2" Decay Power @ Core Blockage (MW) 68.25 63.01 54.92 1 Accumulators do not contribute to the RWST depletion 2 The CS pumps injection volumetric flow rate rumps up at pump startup and stabilizes to a nominal value defined by the pump and injection pipeline characteristics. |
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| 1 Accumulators do not contribute to the RWST depletion 2 The CS pumps injection volumetric flow rate rumps up at pump startup and stabilizes to a nominal value defined by the pump and injection pipeline characteristics.
| | DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16) |
| DRAFT Response to SNPB Follow-up RAI 3-(9/6/16) Core Collapsed Liquid Level (CLL) In the seconds after the core blockage, the liquid water in the core reaches saturation and starts boiling. The boil-off rate depends on the core decay power (as mentioned in the previous sections). The steam leaving the core is replaced by liquid water entering the top of the core in opposite direction. Higher boil-off rates are expected right after the core blockage during larger break scenarios and, subsequently. The core mass inventory during the post-core blockage LTCC is expected to be smaller for larger break scenarios. In the following figure, the core collapsed liquid level is plotted for the cases simulated. In the figure, the core blockage time is set to zero for all the break scenarios to facilitate the comparison. Due to the higher decay power at the SSO (occurring 360 seconds before the core blockage time) in large breaks, voids are created in the core even before the core blockage time. The core collapsed liquid level decreases rapidly after the core blockage time in large breaks due to the smaller water inventory at the top of have a core fully covered at the core blockage time (core collapsed liquid level = 14 ft) and a larger inventory of water available at the top of the core. This effect, combined with a lower decay power level in the core, allows a slower decrease in the core liquid inventory, which will stabilize to a higher level. | | Core Collapsed Liquid Level (CLL) |
| | In the seconds after the core blockage, the liquid water in the core reaches saturation and starts boiling. The boil-off rate depends on the core decay power (as mentioned in the previous sections). The steam leaving the core is replaced by liquid water entering the top of the core in opposite direction. Higher boil-off rates are expected right after the core blockage during larger break scenarios and, subsequently. The core mass inventory during the post-core blockage LTCC is expected to be smaller for larger break scenarios. In the following figure, the core collapsed liquid level is plotted for the cases simulated. In the figure, the core blockage time is set to zero for all the break scenarios to facilitate the comparison. |
| | Due to the higher decay power at the SSO (occurring 360 seconds before the core blockage time) in large breaks, voids are created in the core even before the core blockage time. The core collapsed liquid level decreases rapidly after the core blockage time in large breaks due to the smaller water inventory at the top of the core. Smaller breaks (6 and 2 in the figure) would have a core fully covered at the core blockage time (core collapsed liquid level = 14 ft) and a larger inventory of water available at the top of the core. This effect, combined with a lower decay power level in the core, allows a slower decrease in the core liquid inventory, which will stabilize to a higher level. |
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| DRAFT Response to SNPB Follow-up RAI 3-(9/6/16) Other Considerations: Boundary Conditions volume available for injection, RWST water temperature, ECCS injection characteristics, and all other simulation. Of particular importance is the sump pool temperature profile which is also assumed to be the same of the one imposed as boundary condition for the large hot leg break. Considerations on the heat transfer mechanisms in the reactor containment, and simulation results of different break sizes available in literature [1] show that the sump pool temperature at the SSO time is relatively lower for smaller breaks compared to large breaks. | | DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16) |
| The claddseen: During the pre-core blockage LTCC and before the SSO time, cladding temperature are determined by the amount of flow forced through the core. The flow is lower for smaller breaks (no LHSI pumps running due to the high pressure of the primary system 3). Subsequently, cladding than larger breaks. At SSO, the increase in the injection temperature produces an increase of the cladding temperature. ECCS flow is still forced through the core from the bottom. When core and core bypass is fully blocked, cladding temperature rises. Subcooled and saturated heat transfer regimes are seen in the core for the scenarios simulated. During the post-core blockage LTCC phase, the cladding temperature stabilizes to a level slightly above the saturation temperature at the pressure of the core. Differences in the cladding temperature shown in the figure below are due to the difference in the primary pressure. | | Other Considerations: Boundary Conditions The simulations of the 6 and 2 HLB scenarios and core blockages are executed under the same boundary conditions of the large break (16) scenario. RWST volume available for injection, RWST water temperature, ECCS injection characteristics, and all other boundary conditions are imported from the 16 break simulation. Of particular importance is the sump pool temperature profile which is also assumed to be the same of the one imposed as boundary condition for the large hot leg break. Considerations on the heat transfer mechanisms in the reactor containment, and simulation results of different break sizes available in literature [1] show that the sump pool temperature at the SSO time is relatively lower for smaller breaks compared to large breaks. |
| | The cladding temperature for the three accident scenarios simulated (16, 6, and 2 HLB) is shown in the figure below. In the figure, the following features can be seen: |
| | During the pre-core blockage LTCC and before the SSO time, cladding temperature are determined by the amount of flow forced through the core. |
| | The flow is lower for smaller breaks (no LHSI pumps running due to the high pressure of the primary system 3). Subsequently, cladding temperature of small breaks (2 in the figure) is found to be higher than larger breaks. |
| | At SSO, the increase in the injection temperature produces an increase of the cladding temperature. ECCS flow is still forced through the core from the bottom. |
| | When core and core bypass is fully blocked, cladding temperature rises. |
| | Subcooled and saturated heat transfer regimes are seen in the core for the scenarios simulated. During the post-core blockage LTCC phase, the cladding temperature stabilizes to a level slightly above the saturation temperature at the pressure of the core. Differences in the cladding temperature shown in the figure below are due to the difference in the primary pressure. |
| | 3 No manual operations to cool down and depressurize the primary system are modeled in the LTCC EM. |
| | This condition generates a relatively high pressure of the primary system and a subsequent low ECCS flow rate through the core. |
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| 3 No manual operations to cool down and depressurize the primary system are modeled in the LTCC EM. This condition generates a relatively high pressure of the primary system and a subsequent low ECCS flow rate through the core.
| | DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16) |
| DRAFT Response to SNPB Follow-up RAI 3-(9/6/16) Final Remarks and Conclusions The simulations of the three HLB scenarios are execute under the same boundary conditions. The important phenomena observed in the simulations are similar. Nevertheless, the large break scenario shows bounding initial conditions for the post-core blockage LTCC simulation, specifically in regards to the status of important thermal-hydraulic parameters of the reactor system at the SSO time: Decay power level, Water inventory in the core (Core CLL) Water inventory in the primary system The arguments and discussion provided above confirm that the overall simulation results for the large HLB LTCC scenario are bounding the ones for smaller breaks. [1]. NUREG/CR-6770GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System & Containments to Selected Accident SequencesAugust 2002. (http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6770/)}} | | Final Remarks and Conclusions The simulations of the three HLB scenarios are execute under the same boundary conditions. The important phenomena observed in the simulations are similar. |
| | Nevertheless, the large break scenario shows bounding initial conditions for the post-core blockage LTCC simulation, specifically in regards to the status of important thermal-hydraulic parameters of the reactor system at the SSO time: |
| | Decay power level, Water inventory in the core (Core CLL) |
| | Water inventory in the primary system The arguments and discussion provided above confirm that the overall simulation results for the large HLB LTCC scenario are bounding the ones for smaller breaks. |
| | [1]. NUREG/CR-6770, GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System & Containments to Selected Accident Sequences. August 2002. |
| | (http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6770/)}} |
Letter Sequence Request |
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CAC:MF2400, Control Room Habitability (Approved, Closed) CAC:MF2402, Control Room Habitability (Approved, Closed) |
Administration
- Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting
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MONTHYEARML15274A5992015-10-0101 October 2015
10-01-15 Public Phone Call
Project stage: Request
ML16011A0612016-02-0202 February 2016
Summary of 10/1/15 Meeting with STP Nuclear Operating Company to Discuss Revised Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 at South Texas Project, Units 1 and 2 (CAC MF
Project stage: Meeting
ML16028A1522016-02-18018 February 2016
1/14/2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Modifications to the Licensee'S Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2
Project stage: Meeting
ML16088A2432016-04-0101 April 2016
2/18/2016, Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409)
Project stage: Meeting
ML16092A0442016-04-11011 April 2016
Summary of 3/3/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409)
Project stage: Meeting
ML16082A5072016-04-11011 April 2016
Request for Additional Information, Phased Response Requested, Exemption and License Amendment Request for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC Nos. MF2400-MF2409)
Project stage: RAI
ML16095A0102016-04-13013 April 2016
Summary of November 17-19, 2015 Thermal-Hydraulic Review at Texas A&M University; Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC MF2400-MF2409)
Project stage: Other
ML16092A0852016-04-13013 April 2016
3/17/2016 Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2401-MF2409)
Project stage: Meeting
ML16096A0652016-04-13013 April 2016
Summary of February 4 and 16, 2016, Regulatory Audit at Westinghouse in Rockville, MD, Boric Acid Precipitation, Exemption and License Amendment Request, Risk-Informed Approach to Resolve Generic Safety Issue 191
Project stage: Other
ML16111A0272016-04-18018 April 2016
STP Nuclear Operating Company Slideshow for Public Meeting on April 21, 2016 for GSI-191 Resolution (CAC No. MF2400-01)
Project stage: Meeting
ML16103A3442016-04-26026 April 2016
Summary of February 24-26, 2016 Audit, Debris Transport Review at Alion Science and Technology Corporation, Albuquerque, Nm; Pilot Submittal and Exemption Request, Risk-Informed Approach to Resolve GSI 191
Project stage: Other
ML16032A4032016-04-26026 April 2016
FRN, Draft Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400-MF2409)
Project stage: Draft Other
ML16032A3872016-04-26026 April 2016
Draft Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400-MF2409)
Project stage: Draft Other
ML16127A4002016-05-11011 May 2016
Audit Summary, Thermal-Hydraulic Review on February 23-25, 2016 at Texas A&M University; Pilot Submittal and Request for Exemptions for a Risk-Informed Approach to Resolve Generic Safety Issue 191 (CAC MF2400-MF2409)
Project stage: Other
ML16125A2902016-05-26026 May 2016
Request for Additional Information, Risk Review, Exemption and License Amendment Request for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191 (CAC Nos. MF2400-MF2409)
Project stage: RAI
ML16141B0812016-05-31031 May 2016
Audit Summary, Risk Audit on April 12-13, 2016, at Alumni Center at the University of Texas, Austin, Tx; Pilot Generic Safety Issue 191 Submittal and Exemption Request, and Draft Request for Additional Information
Project stage: Draft RAI
ML16175A1082016-06-24024 June 2016
Summary of 4/21/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2
Project stage: Meeting
ML16190A0082016-07-13013 July 2016
Summary of 6/22/2016 Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409)
Project stage: Meeting
ML16194A2342016-07-21021 July 2016
Request for Withholding Information from Public Disclosure - 5/6/16 Affidavit Executed by D. Munoz and M. Rozboril, Alion Science and Technology, Response to Follow-up RAIs 18, 38, and 44
Project stage: RAI
ML16242A0102016-08-23023 August 2016
NRR E-mail Capture - (External_Sender) FW: South Texas Project GSI-191 Draft Risk Responses to Questions for Public Meeting on August 29, 2016
Project stage: Request
ML16242A0092016-08-25025 August 2016
NRR E-mail Capture - (External_Sender) South Texas Project GSI-191 Draft thermal-hydraulic Responses to Questions for Public Meeting on August 29, 2016
Project stage: Request
ML16238A5272016-09-0101 September 2016
Summary of 7/28/16, Public Meeting with STP Nuclear Operating Company to Discuss the License Amendment and Exemption Requests to Use a Risk-Informed Approach to the Resolution of GSI-191 (CAC Nos. MF2400 - MF2409)
Project stage: Meeting
ML16258A3672016-09-0606 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution SNPB-3-2, Bullet 4 16 -inch Bounding Break Clarification for September 14, 2016 Public Meeting
Project stage: Request
ML16258A1002016-09-12012 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI APLA-3-2 Clarification for September 14, 2016 Public Meeting
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ML16258A1042016-09-12012 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI SNPB-3-13 Clarification for September 14, 2016 Public Meeting
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ML16258A0962016-09-13013 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI SSIB-3-4 Clarification for September 14, 2016 Public Meeting
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ML16258A0982016-09-13013 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 33 Clarification for September 14, 2016 Public Meeting
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ML16258A2072016-09-14014 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 37 Clarification for September 14, 2016 Public Meeting
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ML16258A2082016-09-14014 September 2016
NRR E-mail Capture - (External_Sender) South Texas Project Generic Safety Issue 191 Resolution RAI 34 Clarification for September 14, 2016 Public Meeting
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ML16246A0222016-10-0505 October 2016
Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items Related to Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400 - MF2409)
Project stage: Meeting
ML16279A3312016-10-27027 October 2016
Summary of Public Meeting with STP Nuclear Operating Company Regarding Open Items, Risk-Informed Generic Safety Issue (GSI) 191 Application and Supplements for South Texas Project, Units 1 and 2 (CAC Nos. MF2400-MF2409)
Project stage: Meeting
ML16302A4532016-12-12012 December 2016
Closeout of Request for Additional Information Questions That Are No Longer Applicable, Resolution of Generic Safety Issue (GSI) 191 (CAC Nos. MF2400-MF2409)
Project stage: RAI
ML16351A1502016-12-16016 December 2016
Correction Letter Closeout of Request for Information Questions That Are No Longer Applicable Associated with the Resolution of Generic Safety Issue 191 (CAC Nos. MF2400 - MF2409)
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ML16278A5982017-05-0202 May 2017
Letter, Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400 to MF2409)
Project stage: Other
ML16278A5992017-05-0202 May 2017
Environmental Assessment and Finding of No Significant Impact, Revise Licensing Basis as Documented in the UFSAR and Request for Exemptions, Risk-Informed Approach to Address GSI-191 (CAC MF2400 to MF2409)
Project stage: Other
ML17137A3252017-05-17017 May 2017
Safety Evaluation of LAR by South Texas Project Nuclear Operating Company to Adopt a Risk-Informed Resolution of Generic Safety Issue-191
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ML17151A8432017-06-16016 June 2017
OEDO-17-00326 - EDO Response to ACRS Chairman, Safety Evaluation of License Amendment Request by STP Nuclear Operating Company to Adopt a Risk-Informed Resolution of Generic Safety Issue-191
Project stage: Approval
NRC-2016-0092, South Texas Project, Units 1 and 2 - FRN: Exemption from the Requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix a, General Design Criteria 35, 38, 41 (CAC Nos. MF2402-MF2409)2017-07-11011 July 2017
South Texas Project, Units 1 and 2 - FRN: Exemption from the Requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix a, General Design Criteria 35, 38, 41 (CAC Nos. MF2402-MF2409)
Project stage: Approval
ML17037C8712017-07-11011 July 2017
Letter, Exemption from the Requirements of 10 CFR Part 50, Section 50.46 (CAC Nos. MF2402-MF2409)
Project stage: Approval
ML17037C8762017-07-11011 July 2017
FRN: Exemption from the Requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix a, General Design Criteria 35, 38, 41 (CAC Nos. MF2402-MF2409)
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ML17055A5002017-07-11011 July 2017
Enclosure 4 to Amendment Nos. 212 and 198, Resolution of Licensee Comments on Safety Evaluation - Risk-Informed Approach to Resolve Generic Safety Issue 191
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ML17019A0032017-07-11011 July 2017
Attachment 2 to Safety Evaluation - In-Vessel Thermal-Hydraulic Analysis, Issuance of Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191
Project stage: Approval
ML17038A2232017-07-11011 July 2017
Issuance of Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191
Project stage: Approval
ML17019A0022017-07-11011 July 2017
Safety Evaluation, Enclosure 3 to Amendment Nos. 212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191
Project stage: Approval
2016-05-31
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Category:E-Mail
MONTHYEARML24086A4212024-06-26026 June 2024
Email to Chris Warren Decision on the Exemption Request for South Texas Project
ML24086A4252024-06-24024 June 2024
Email to Chris Warren Availability of Environmental Assessment for the South Texas Project Exemption
ML24101A0082024-04-10010 April 2024
NRR E-mail Capture - South Texas Project - Relief Request - 50.55a(z)(2) - Rvh Penetration 75
ML24052A0152024-02-20020 February 2024
NRR E-mail Capture - South Texas Project - Request for Additional Information - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications (L-2023-LLE-0046)
ML24052A0112024-02-16016 February 2024
NRR E-mail Capture - South Texas Project - Draft Request for Additional Information - Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications (L-2023-LLE-0046)
ML23334A0062023-11-29029 November 2023
NRR E-mail Capture - South Texas Project - Request for Additional Information - Proposed Alternative to ASME BPV Code Requirements
ML23319A4442023-11-15015 November 2023
NRR E-mail Capture - South Texas Project - Draft Request for Additional Information - Proposed Alternative to ASME BPV Code Requirements
ML23278A1532023-10-0505 October 2023
NRR E-mail Capture - South Texas Project Request for Additional Information - Request for Indirect Transfer of Licenses and Conforming Amendments (L-2023-LLM-0004)
ML23264A0972023-09-21021 September 2023
NRR E-mail Capture - South Texas Project - Request for Additional Information - License Amendment Request to Revise the Alternate Source Term Dose Calculation (L-2023-LLA-0047)
ML23255A2862023-09-12012 September 2023
NRR E-mail Capture - South Texas Project Draft Request for Additional Information License Amendment Request to Revise the Alternate Source Term Dose Calculation (L-2023-LLA-0047)
ML23191A0102023-07-0707 July 2023
NRR E-mail Capture - South Texas Project - Acceptance of Request for Indirect Transfer of Licenses and Conforming Amendments (L-2023-LLM-0004)
ML23145A1872023-05-24024 May 2023
July 2023 Emergency Preparedness Program Inspection - Request for Information
ML23124A3272023-05-0404 May 2023
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request to Revise the Alternate Source Term Dose Calculation (L-2023-LLA-0047)
ML23115A0702023-04-24024 April 2023
NRR E-mail Capture - South Texas Project Units 1 and 2 - Audit Plan - Proposed Alternative to ASME Code, Section XI Requirements for Containment Building Inspections
ML23096A1592023-04-0606 April 2023
NRR E-mail Capture - South Texas Project - Acceptance of Requested Licensing Action - Proposed Alternative to the Requirements of the ASME Code
ML23065A0912023-03-0303 March 2023
NRR E-mail Capture - South Texas Project - Draft Supplemental Information Request - Proposed Alternative to the Requirements of the ASME Code
ML23003A8082022-11-28028 November 2022
NRR E-mail Capture - Texas Commission on Environmental Quality (TCEQ) Review/Comment Regarding the Draft Environmental Assessment for the South Texas Project Alternate Disposal Request (L-2021-LLL-0022))
ML22249A0692022-09-0606 September 2022
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request Regarding Adoption of TSTF-554, Revise Reactor Coolant Leakage Requirements (L-2002-LLA-0113)
ML22206A0142022-07-20020 July 2022
NRR E-mail Capture - South Texas Project - Request for Additional Information - 10 CFR 20.2002 Alternate Disposal Request
ML22194A0502022-07-0707 July 2022
NRR E-mail Capture - South Texas Project - Updated Draft Request for Additional Information - 10 CFR 20.2002 Alternate Disposal Request
ML22194A0492022-06-0808 June 2022
NRR E-mail Capture - South Texas Project - Draft Request for Additional Information - 10 CFR 20.2002 Alternate Disposal Request
ML22123A3272022-05-0303 May 2022
STP EP Exercise Inspection July 2022 RFI
ML22115A2632022-04-19019 April 2022
Email Response to David Lochbaum Regarding Interim Configuration of Holtec HI-STORE FW Spent Fuel Multipurpose Canister at South Texas Project Energy Generating Station
ML22105A5642022-04-0505 April 2022
Request for Information for Problem Identification and Resolution at STP Nuclear Operating Company
ML21344A0072021-12-0909 December 2021
NRR E-mail Capture - South Texas Project - Acceptance of Requested Licensing Action - Alternative Disposal Request
ML21305B8022021-11-0101 November 2021
NRR E-mail Capture - South Texas Project - Acceptance of Requested Licensing Action - Inservice Inspection Impracticality
ML21257A2262021-09-14014 September 2021
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request Regarding Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (L-2021-LLA-0151)
ML21160A1522021-06-0909 June 2021
Request for Information NRC Inspection 2021-003 Public Radiation Safety
ML21133A2372021-05-0505 May 2021
Request for Additional Information: STP EP Exercise Inspection - July 2021
ML21117A3242021-04-22022 April 2021
NRR E-mail Capture - (External_Sender) STP Questions About Eplan Amendment SE
ML21103A0072021-04-12012 April 2021
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request to Add a Note to TS 3.6.3 and to Remove the TS Index (L-2021-LLA-0041)
ML21058A0002021-02-26026 February 2021
NRR E-mail Capture - South Texas Units 1 and 2 - License Amendment Request to Revise Moderator Temperature Coefficient SRs - Audit Plan and Setup of Online Document Access (L-2021-LLA-0004)
ML21049A2652021-02-16016 February 2021
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request to Revise the Moderator Temperature Coefficient Surveillance Requirements (L-2021-LLA-0004)
ML21039A8902021-02-0808 February 2021
NRR E-mail Capture - South Texas Project - Request for Additional Information - 1RE22 Inspection Summary Report for Steam Generator Tubing
ML21029A3312021-01-29029 January 2021
NRR E-mail Capture - South Texas Project - Draft Request for Additional Information - 1RE22 Inspection Summary Report for Steam Generator Tubing
ML20244A0222020-08-28028 August 2020
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request to Adopt TSTF-374, Revision to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil. (L-2020-LLA-0172)
ML20135H1982020-05-14014 May 2020
NRR E-mail Capture - South Texas Unit 1 - License Amendment Request to Revise Safety Injection System Accumulator Pressure - Audit Plan and Setup of Online Reference Portal
ML20125A3312020-05-0404 May 2020
NRR E-mail Capture - South Texas Project - Acceptance of License Amendment Request to Revise the Emergency Plan (L-2020-LLA-0057)
ML20111A0052020-04-17017 April 2020
NRR E-mail Capture - South Texas Project - Request for Additional Information - Proposed Alternative to ASME OM Code 2012 Edition - Relief Request PRR-01
ML20050E1192020-02-19019 February 2020
NRR E-mail Capture - South Texas Project - Acceptance of Requested Licensing Actions - Proposed Alternatives to ASME OM Code Requirements (EPID: L 2020-LLR-0007 to L-2020-LLR-0010)
ML19344A0092019-12-0909 December 2019
NRR E-mail Capture - South Texas Project - Acceptance of Requested Licensing Action - Proposed Alternative to ASME Code Requirements for the Repair of Essential Cooling Water System Class 3 Buried Piping
ML19309F1412019-11-0404 November 2019
Supplemental Information Request - Proposed Alternative to ASME Code Requirements for the Repair of Essential Cooling Water System Class 3 Buried Piping (EPID L 2019-LLR-0096) (Draft)
ML19291A0332019-10-17017 October 2019
NRR E-mail Capture - Acceptance Review Results: STP License Amendment Request Regarding Request to Delete E Bar Definition
ML19165A1032019-06-14014 June 2019
NRR E-mail Capture - Draft RAI for STP TS 3.8.1.1 Sbdg SR Change
ML19162A1162019-05-22022 May 2019
NRR E-mail Capture - Acceptance Review Results: STP License Amendment Request Regarding TSTF-529
ML19095A6562019-04-0404 April 2019
NRR E-mail Capture - (External_Sender) Handout for 4/10/19 Meeting to Discuss a Proposed Request for South Texas Project, Units 1 and 2, Regarding Repair of Piping Using Carbon Fiber Repair Methods
ML19081A1512019-03-21021 March 2019
NRR E-mail Capture - Draft Round 2 RAI - Sbdg Voltage and Frequency LAR (L-2018-LLA-0078)
ML19078A3792019-03-19019 March 2019
NRR E-mail Capture - Acceptance Review Results: STP RR-ENG-3-23 (L-2018-LLR-0021)
ML18292A7272018-10-19019 October 2018
NRR E-mail Capture - Acceptance Review Results: TSTF-522 (L-2018-LLA-0271)
ML18283B9522018-10-10010 October 2018
NRR E-mail Capture - Final RAI - South Texas Standby DG TS Change (L-2018-LLA-0078)
2024-06-26
[Table view] |
Text
NRR-PMDAPEm Resource From: Harrison Albon <awharrison@STPEGS.COM>
Sent: Tuesday, September 06, 2016 3:52 PM To: Regner, Lisa
Subject:
[External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Attachments: SNPB-3-02 followup - partial 2 Bullet 4.pdf
- Lisa, Here is another partial for discussion in the public call on 9/14.
Wayne 1
Hearing Identifier: NRR_PMDA Email Number: 3064 Mail Envelope Properties (FC3FC22E-DF52-40BF-806A-7633810123B7)
Subject:
[External_Sender] STP draft response to SNPB 3-2 bullet 4 re 16" break bounding Sent Date: 9/6/2016 3:51:58 PM Received Date: 9/6/2016 3:52:04 PM From: Harrison Albon Created By: awharrison@STPEGS.COM Recipients:
"Regner, Lisa" <Lisa.Regner@nrc.gov>
Tracking Status: None Post Office: stpegs.com Files Size Date & Time MESSAGE 88 9/6/2016 3:52:04 PM SNPB-3-02 followup - partial 2 Bullet 4.pdf 397348 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16)
- 1. Follow-up SNPB-3-2 Accident Scenario Progression Initial RAI: Provide a description of the accident progression of the accident scenarios being simulated using the LTCC EM. This description should start at the initiation of the break, define each phase, and the important phenomena occurring in that phase in the various locations of the RCS (e.g., core, reactor vessel, steam generators - both primary and secondary side, loops, pressurizer, pumps, containment)
Though only the 16-inch hot leg break analysis was provided, STPNOC did not justify or demonstrate that the 16 inch hot leg break bounds smaller hot leg breaks.
Justify that the 16-inch hot leg break analysis bounds the smaller hot leg break scenarios.
STP Response The 16-inch HLB LOCA scenario progression is described as part of the response to the SNPB-3-02. The accident progression simulated with the LTCC EM can be summarized as follow:
Immediately after the break event, the primary system experiences a fast depressurization where water and steam are discharged from the break.
Reactor scram and ECCS actuation occur, controlled by low pressure signal.
Due to the location of the break, injection of liquid water from the cold side is sustained by the ECCS injection system (HHSI, Accumulators, and LHSI).
Core liquid inventory increases rapidly after the actuation of the accumulators and SI pump. Pre-Core Blockage LTCC starts when the core is fully flooded (Core Collapsed Liquid Level = 14 feet).
When RWST water is depleted, injection start from the sump (SSO time)
The total liquid inventory in the core and in the primary system is sufficient to supply the liquid water to the core immediately after the core blockage event.
Liquid inventory in the core decreases until flow through alternative flow paths (steam generators u-tubes) is established.
Liquid water flow is maintained at the top of the core through the steam generators during the Post-Core Blockage LTCC.
Different important thermal-hydraulic parameters can be identified from the description provided. These parameters (listed and described below) can be used as figure of merit to confirm that the large break scenarios bound smaller hot leg breaks.
DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16)
Sump Switchover time The sump switchover (SSO) time and, subsequently, the core blockage time, is one of the most important parameters defining the thermal-hydraulic initial conditions of the reactor system for the post-core blockage LTCC phase. The SSO time is defined based on the RWST volume available for injection, and the injection rate of the SI system and CS system. The rate of injection has three1 contributors:
The High Head Safety Injection (HHSI) pumps The Low Head Safety Injection (LHSI) pumps The Containment Spray (CS) pumps The rate of injection of the HHSI and LHSI depends on the pressure of the primary system and, based on the pump characteristics, is expected to increase at lower primary system pressures. The contribution of the CS pumps may be assumed to be constant throughout the accident progression2. The primary pressure during the accident progression is strongly dependent on the break size.
In particular, larger breaks are expected to experience a faster depressurization, stabilizing the primary system at a lower pressure. Subsequently, the total ECCS injection is expected to be higher during large break scenario than smaller breaks.
The SSO time for the three HLB scenarios simulated is reported in the table below.
Break Size 16" 6" 2" SSO Time (s) 1740 2302 3729 Decay Power at Core Blockage and Core Boil-off Rate The LTCC EM calculates the decay power using the ANS-1979 model. Based on the considerations on the SSO time provided in the previous paragraph, the decay power at time of core blockage is higher for larger breaks since the core blockage event occurs earlier in the transient. The amount of water required to compensate the boil-off rate and cool down the reactor core during at the post-core blockage LTCC is relatively larger in the 16-inch HLB scenario than smaller break scenarios.
The core decay power at the time of core blockage for the three HLB scenarios simulated is reported in the table below.
Break Size 16" 6" 2" Decay Power @ Core Blockage (MW) 68.25 63.01 54.92 1 Accumulators do not contribute to the RWST depletion 2 The CS pumps injection volumetric flow rate rumps up at pump startup and stabilizes to a nominal value defined by the pump and injection pipeline characteristics.
DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16)
Core Collapsed Liquid Level (CLL)
In the seconds after the core blockage, the liquid water in the core reaches saturation and starts boiling. The boil-off rate depends on the core decay power (as mentioned in the previous sections). The steam leaving the core is replaced by liquid water entering the top of the core in opposite direction. Higher boil-off rates are expected right after the core blockage during larger break scenarios and, subsequently. The core mass inventory during the post-core blockage LTCC is expected to be smaller for larger break scenarios. In the following figure, the core collapsed liquid level is plotted for the cases simulated. In the figure, the core blockage time is set to zero for all the break scenarios to facilitate the comparison.
Due to the higher decay power at the SSO (occurring 360 seconds before the core blockage time) in large breaks, voids are created in the core even before the core blockage time. The core collapsed liquid level decreases rapidly after the core blockage time in large breaks due to the smaller water inventory at the top of the core. Smaller breaks (6 and 2 in the figure) would have a core fully covered at the core blockage time (core collapsed liquid level = 14 ft) and a larger inventory of water available at the top of the core. This effect, combined with a lower decay power level in the core, allows a slower decrease in the core liquid inventory, which will stabilize to a higher level.
DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16)
Other Considerations: Boundary Conditions The simulations of the 6 and 2 HLB scenarios and core blockages are executed under the same boundary conditions of the large break (16) scenario. RWST volume available for injection, RWST water temperature, ECCS injection characteristics, and all other boundary conditions are imported from the 16 break simulation. Of particular importance is the sump pool temperature profile which is also assumed to be the same of the one imposed as boundary condition for the large hot leg break. Considerations on the heat transfer mechanisms in the reactor containment, and simulation results of different break sizes available in literature [1] show that the sump pool temperature at the SSO time is relatively lower for smaller breaks compared to large breaks.
The cladding temperature for the three accident scenarios simulated (16, 6, and 2 HLB) is shown in the figure below. In the figure, the following features can be seen:
During the pre-core blockage LTCC and before the SSO time, cladding temperature are determined by the amount of flow forced through the core.
The flow is lower for smaller breaks (no LHSI pumps running due to the high pressure of the primary system 3). Subsequently, cladding temperature of small breaks (2 in the figure) is found to be higher than larger breaks.
At SSO, the increase in the injection temperature produces an increase of the cladding temperature. ECCS flow is still forced through the core from the bottom.
When core and core bypass is fully blocked, cladding temperature rises.
Subcooled and saturated heat transfer regimes are seen in the core for the scenarios simulated. During the post-core blockage LTCC phase, the cladding temperature stabilizes to a level slightly above the saturation temperature at the pressure of the core. Differences in the cladding temperature shown in the figure below are due to the difference in the primary pressure.
3 No manual operations to cool down and depressurize the primary system are modeled in the LTCC EM.
This condition generates a relatively high pressure of the primary system and a subsequent low ECCS flow rate through the core.
DRAFT Response to SNPB Follow-up RAI 3-2 (partial): Bullet 4 re 16 break being bounding (9/6/16)
Final Remarks and Conclusions The simulations of the three HLB scenarios are execute under the same boundary conditions. The important phenomena observed in the simulations are similar.
Nevertheless, the large break scenario shows bounding initial conditions for the post-core blockage LTCC simulation, specifically in regards to the status of important thermal-hydraulic parameters of the reactor system at the SSO time:
Decay power level, Water inventory in the core (Core CLL)
Water inventory in the primary system The arguments and discussion provided above confirm that the overall simulation results for the large HLB LTCC scenario are bounding the ones for smaller breaks.
[1]. NUREG/CR-6770, GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System & Containments to Selected Accident Sequences. August 2002.
(http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6770/)
