ML112081967: Difference between revisions

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| document type = E-Mail, Request for Additional Information (RAI)
| document type = E-Mail, Request for Additional Information (RAI)
| page count = 2
| page count = 2
| project = TAC:ME2610, TAC:ME2609
| project = TAC:ME2609, TAC:ME2610
| stage = RAI
| stage = RAI
}}
}}
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==Subject:==
==Subject:==
Prairie Island - Request for Clarification of June 22, 2011 Request for Additional Information (RAI) Response Following is the consolidated request for clarification of Prairie Island's June 22, 2011 RAI response (ADAMS Accession No. ML111740145) concerning the license amendment request to adopt the Alternative Source Term methodology. This request modifies/supersedes the draft requests for clarification dated June 24, 2011 (ADAMS Accession No. ML111822631) and June 29, 2011 (ADAMS Accession No. ML111822669), following discussions with the NRC staff.
Prairie Island - Request for Clarification of June 22, 2011 Request for Additional Information (RAI) Response Following is the consolidated request for clarification of Prairie Island's June 22, 2011 RAI response (ADAMS Accession No. ML111740145) concerning the license amendment request to adopt the Alternative Source Term methodology. This request modifies/supersedes the draft requests for clarification dated June 24, 2011 (ADAMS Accession No. ML111822631) and June 29, 2011 (ADAMS Accession No. ML111822669), following discussions with the NRC staff.
Question #4 has been modified based on the discussions. In addition, some additional editing has been performed to enhance clarity.
Question #4 has been modified based on the discussions. In addition, some additional editing has been performed to enhance clarity.  
MODIFIED REQUEST FOR ADDITIONAL INFORMATION  (1) Table 2 lists the safety-related (SR), non-SR (NSR) and augmented quality (AQ) systems, components and instruments (SCIs) available for steam generator tube rupture (SGTR) mitigation.
Please provide a definition of the AQ SCIs, identify the NSR and AQ SCIs in Table 2 that are credited in the margin-to-overfill (MTO) or mass release analyses, discuss the functions of each identified SCIs, address the acceptability of each of the NSR and AQ SCIs for consequences mitigation assumed in the analyses for supporting the licensing applications, and demonstrate that the use of NSR or AQ SCIs meets the intent of use of the SR SCIs for design-basis analysis.  (2) The operator action delay times for reactor coolant system (RCS) depressurization are 4 minutes used in the MTO analysis and 7 minutes in the mass release analysis listed in Tables 5, and 6, respectively. Provide bases for the different operator action times used for RCS depressurization, while the operator action times for RCS cooldown initiation (19 minutes following reactor trip), safety injection (SI) termination (2 minutes) and charging flow termination (15 minutes) remain the same in the MTO and mass release analysis. Discuss the acceptance criteria for the results of the simulator exercises in support of the operator action times (4, 2, 19, 15 minutes above) credited in the analysis.
  (3) Figures 11 and 16 show the RCS and steam generator (SG) pressures, and steam releases for the mass release analysis. 1. Figure 11 indicates that at about 2850 seconds, and 3100 seconds, there are sudden decreases in the intact SG pressure and RCS pressure, respectively. Explain the thermal hydraulic phenomena, mitigating systems, or operator actions that contribute to the decreases. 2. Figure 16 shows that for the ruptured SG, the steam releases suddenly increase and decrease between 100 to 200 seconds into the transient. At 800 seconds, the releases suddenly increase and then decrease gradually, following a deep gradual decrease at 1300 second until 1500 seconds. After 1500 seconds,  the releases become a small, constant rate. Explain the changes identified above for the steam releases through the ruptured SG. For the intact SG, the first significant steam releases occur between 1200 - 1900 seconds. This is due to the operator action that opens the SG power operated relief valve (PORV) for RCS cooldown. Explain the causes for the second significant release to occur after 2800 seconds until 4000 seconds when the presented results end.    (4) Observing the time-temperature plots (Figures 5 and 6) provided in your RAI response dated June 22, 2011, the transient associated with the SGTR MTO analysis shows what could be a pressurized thermal shock (PTS) event with respect to the reactor pressure vessel. Please explain, based on your understanding of the technical basis for the NRC's PTS Rule, 10 CFR 50.61, and the condition (RTPTS values) of the limiting materials in the Prairie Island, Unit 1 and 2 reactor pressure vessels, why you understand these reactor pressure vessels to be adequately protected from failure due to the transients shown in your RAI response for the operating lifetime of the facility.  (5)  Table 1 indicates that the reactor trip occurs at 49 seconds following the event initiation. Specify the signal that actuates the reactor trip, discuss the trip point assumed in the analysis and show that it is within the technical specification (TS) value with inclusion of instrumentation uncertainties.        (6)  The first paragraph on page 47 discusses the basis to support the conclusion that the original  hand-calculated releases are the bounding result and remain valid. Please provide data of the dose releases through the intact and rupture SGs for the pre-trip and post-trip, and other applicable conditions to support the stated conclusion. Tom Wengert Project Manager - Prairie Island USNRC NRR/DORL/LPLIII-1 (301) 415-4037 


}}
MODIFIED REQUEST FOR ADDITIONAL INFORMATION (1) Table 2 lists the safety-related (SR), non-SR (NSR) and augmented quality (AQ) systems, components and instruments (SCIs) available for steam generator tube rupture (SGTR) mitigation.   
 
Please provide a definition of the AQ SCIs, identify the NSR and AQ SCIs in Table 2 that are credited in the margin-to-overfill (MTO) or mass release analyses, discuss the functions of each identified SCIs, address the acceptability of each of the NSR and AQ SCIs for consequences mitigation assumed in the analyses for supporting the licensing applications, and demonstrate that the use of NSR or AQ SCIs meets the intent of use of the SR SCIs for design-basis analysis.
(2) The operator action delay times for reactor coolant system (RCS) depressurization are 4 minutes used in the MTO analysis and 7 minutes in the mass release analysis listed in Tables 5, and 6, respectively.
Provide bases for the different operator action times used for RCS depressurization, while the operator action times for RCS cooldown initiation (19 minutes following reactor trip), safety injection (SI) termination (2 minutes) and charging flow termination (15 minutes) remain the same in the MTO and mass release analysis. Discuss the acceptance criteria for the results of the simulator exercises in support of the operator action times (4, 2, 19, 15 minutes above) credited in the analysis.
 
  (3) Figures 11 and 16 show the RCS and steam generator (SG) pressures, and steam releases for the mass release analysis.
: 1. Figure 11 indicates that at about 2850 seconds, and 3100 seconds, there are sudden decreases in the intact SG pressure and RCS pressure, respectively. Explain the thermal hydraulic phenomena, mitigating systems, or operator actions that contribute to the decreases.
: 2. Figure 16 shows that for the ruptured SG, the steam releases suddenly increase and decrease between 100 to 200 seconds into the transient. At 800 seconds, the releases suddenly increase and then decrease gradually, following a deep gradual decrease at 1300 second until 1500 seconds. After 1500 seconds,  the releases become a small, constant rate. Explain the changes identified above for the steam releases through the ruptured SG. For the intact SG, the first significant steam releases occur between 1200 - 1900 seconds. This is due to the operator action that opens the SG power operated relief valve (PORV) for RCS cooldown. Explain the causes for the second significant release to occur after 2800 seconds until 4000 seconds when the presented results end. 
(4) Observing the time-temperature plots (Figures 5 and 6) provided in your RAI response dated June 22, 2011, the transient associated with the SGTR MTO analysis shows what could be a pressurized thermal shock (PTS) event with respect to the reactor pressure vessel. Please explain, based on your understanding of the technical basis for the NRC's PTS Rule, 10 CFR 50.61, and the condition (RT PTS values) of the limiting materials in the Prairie Island, Unit 1 and 2 reactor pressure vessels, why you understand these reactor pressure vessels to be adequately protected from failure due to the transients shown in your RAI response for the operating lifetime of the facility.
(5)  Table 1 indicates that the reactor trip occurs at 49 seconds following the event initiation.
Specify the signal that actuates the reactor trip, discuss the trip point assumed in the analysis and show that it is within the technical specification (TS) value with inclusion of instrumentation uncertainties.       
(6)  The first paragraph on page 47 discusses the basis to support the conclusion that the original  hand-calculated releases are the bounding result and remain valid.
Please provide data of the dose releases through the intact and rupture SGs for the pre-trip and post-trip, and other applicable conditions to support the stated conclusion.
Tom Wengert Project Manager - Prairie Island USNRC NRR/DORL/LPLIII-1 (301) 415-4037}}

Revision as of 14:09, 4 August 2018

E-mail Prairie Island - Request for Clarification of June 22, 2011 Request for Additional Information (RAI) Response
ML112081967
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 07/20/2011
From: Wengert T J
Plant Licensing Branch III
To: Eckholt G F, Myers G R
Northern States Power Co
Wengert, Thomas
References
TAC ME2609, TAC ME2610
Download: ML112081967 (2)
v • d • e


Text

From: Wengert, Thomas Sent: Wednesday, July 20, 2011 3:56 PM To: Eckholt, Gene F.; Myers, Gregory R. Cc: Sun, Summer; Ulses, Anthony; Pascarelli, Robert

Subject:

Prairie Island - Request for Clarification of June 22, 2011 Request for Additional Information (RAI) Response Following is the consolidated request for clarification of Prairie Island's June 22, 2011 RAI response (ADAMS Accession No. ML111740145) concerning the license amendment request to adopt the Alternative Source Term methodology. This request modifies/supersedes the draft requests for clarification dated June 24, 2011 (ADAMS Accession No. ML111822631) and June 29, 2011 (ADAMS Accession No. ML111822669), following discussions with the NRC staff.

Question #4 has been modified based on the discussions. In addition, some additional editing has been performed to enhance clarity.

MODIFIED REQUEST FOR ADDITIONAL INFORMATION (1) Table 2 lists the safety-related (SR), non-SR (NSR) and augmented quality (AQ) systems, components and instruments (SCIs) available for steam generator tube rupture (SGTR) mitigation.

Please provide a definition of the AQ SCIs, identify the NSR and AQ SCIs in Table 2 that are credited in the margin-to-overfill (MTO) or mass release analyses, discuss the functions of each identified SCIs, address the acceptability of each of the NSR and AQ SCIs for consequences mitigation assumed in the analyses for supporting the licensing applications, and demonstrate that the use of NSR or AQ SCIs meets the intent of use of the SR SCIs for design-basis analysis.

(2) The operator action delay times for reactor coolant system (RCS) depressurization are 4 minutes used in the MTO analysis and 7 minutes in the mass release analysis listed in Tables 5, and 6, respectively.

Provide bases for the different operator action times used for RCS depressurization, while the operator action times for RCS cooldown initiation (19 minutes following reactor trip), safety injection (SI) termination (2 minutes) and charging flow termination (15 minutes) remain the same in the MTO and mass release analysis. Discuss the acceptance criteria for the results of the simulator exercises in support of the operator action times (4, 2, 19, 15 minutes above) credited in the analysis.

(3) Figures 11 and 16 show the RCS and steam generator (SG) pressures, and steam releases for the mass release analysis.

1. Figure 11 indicates that at about 2850 seconds, and 3100 seconds, there are sudden decreases in the intact SG pressure and RCS pressure, respectively. Explain the thermal hydraulic phenomena, mitigating systems, or operator actions that contribute to the decreases.
2. Figure 16 shows that for the ruptured SG, the steam releases suddenly increase and decrease between 100 to 200 seconds into the transient. At 800 seconds, the releases suddenly increase and then decrease gradually, following a deep gradual decrease at 1300 second until 1500 seconds. After 1500 seconds, the releases become a small, constant rate. Explain the changes identified above for the steam releases through the ruptured SG. For the intact SG, the first significant steam releases occur between 1200 - 1900 seconds. This is due to the operator action that opens the SG power operated relief valve (PORV) for RCS cooldown. Explain the causes for the second significant release to occur after 2800 seconds until 4000 seconds when the presented results end.

(4) Observing the time-temperature plots (Figures 5 and 6) provided in your RAI response dated June 22, 2011, the transient associated with the SGTR MTO analysis shows what could be a pressurized thermal shock (PTS) event with respect to the reactor pressure vessel. Please explain, based on your understanding of the technical basis for the NRC's PTS Rule, 10 CFR 50.61, and the condition (RT PTS values) of the limiting materials in the Prairie Island, Unit 1 and 2 reactor pressure vessels, why you understand these reactor pressure vessels to be adequately protected from failure due to the transients shown in your RAI response for the operating lifetime of the facility.

(5) Table 1 indicates that the reactor trip occurs at 49 seconds following the event initiation.

Specify the signal that actuates the reactor trip, discuss the trip point assumed in the analysis and show that it is within the technical specification (TS) value with inclusion of instrumentation uncertainties.

(6) The first paragraph on page 47 discusses the basis to support the conclusion that the original hand-calculated releases are the bounding result and remain valid.

Please provide data of the dose releases through the intact and rupture SGs for the pre-trip and post-trip, and other applicable conditions to support the stated conclusion.

Tom Wengert Project Manager - Prairie Island USNRC NRR/DORL/LPLIII-1 (301) 415-4037

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