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{{#Wiki_filter:Farley Nuclear Plant Residual Heat Removal Autoclosure Interlock Deletion April 23, 2014 1
{{#Wiki_filter:Farley Nuclear Plant Residual Heat Removal Autoclosure Interlock Deletion April 23, 2014         1
Farley Nuclear Plant RHR ACI Deletion Agenda Introductions Purpose of Meeting
 
Farley Nuclear Plant RHR ACI Deletion Agenda
* Introductions
* Purpose of Meeting
* Background
* Issue for Discussion
* Proposed Approach
* Summary and Conclusions 2


===Background===
Issue for Discussion Proposed Approach Summary and Conclusions 2
Farley Nuclear Plant RHR ACI Deletion Purpose of Meeting Discuss the approach and obtain NRC feedback and expectations on the technical justification for the elimination of the Residual Heat Removal (RHR) System Autoclosure Interlock (ACI).
Farley Nuclear Plant RHR ACI Deletion Purpose of Meeting Discuss the approach and obtain NRC feedback and expectations on the technical justification for the elimination of the Residual Heat Removal (RHR) System Autoclosure Interlock (ACI).
3 Farley Nuclear Plant RHR ACI Deletion
3
 
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
RHR Autoclosure Interlock oEnsures a double isolation valve barrier between the RCS and RHR system when the plant is at normal operating conditions oThe double isolation valve barrier provides protection against overpressurizing the low pressure RHR system by the high pressure RCS oBoth RHR suction isolation valves close automatically if the pressure increases above the bistable setpoint oHelps to prevent interfacing system LOCAs
* RHR Autoclosure Interlock o Ensures a double isolation valve barrier between the RCS and RHR system when the plant is at normal operating conditions o  The double isolation valve barrier provides protection against overpressurizing the low pressure RHR system by the high pressure RCS o Both RHR suction isolation valves close automatically if the pressure increases above the bistable setpoint o Helps to prevent interfacing system LOCAs 4
 
OpsRhr051 M
FCV-602A 8864A M                                    PRT CHARGING PUMP 8708A                                            SUCTION              SAMPLE 8706A    SYS PRT                                                FIS                                                        FIS PT  602A  TE      CCW                          TE            605A PT                                      PT        600A        604A                                  606A                                      M 402  M      M                          601A LOOP A LOOP C                                                  A                                                                                                                COLD LEG RHR                                                                                          8888A HOT LEG                                                                            RHR HX A          HCV-603A 8701B  8701A                          S      PUMP M
PRT 8887A M        M REFUELING 8812A    8809A                                                                                    M M                              WATER                                                  8865 STORAGE                        FCV-605A                                                  LOOPS CONTAINMENT                                                  TANK                                                                                    A&B SUMP                                                                                                                        8889 8811A                                                                                                                      HOT LEG CVCS LETDOWN LOOP B HCV-142      LINE                    8881 M                                                                                                  COLD LEG 8864B PRT FCV-602B M
CHARGING PUMP 8708B                      M                    SUCTION            SAMPLE 8706B PRT                            8809B FIS SYS FIS PT  602B  TE      CCW                      TE              605B PT 600B        604B                              606B                        M                M PT 403  M      M                          601B                                                                                8887B                                LOOP C LOOP A                                                  B                                                                                                                COLD LEG RHR                                                                                          8888B HOT LEG                                                                            RHR HX B        HCV-603B 8702B  8702A                          S      PUMP M
8812B M                                                              FCV-605B CONTAINMENT SUMP 8811B INSIDE OUTSIDE                                                                                                              OUTSIDE INSIDE CONTAINMENT  CONTAINMENT                                                                                                    CONTAINMENT  CONTAINMENT 5


4 OpsRhr051 CHARGING PUMP CHARGING PUMP CONTAINMENT LOOP A HOT LEG 8812B 8702B 8702A CONTAINMENT INSIDE SUMP 8811B CONTAINMENT OUTSIDE M M 8708B PRT PT 403 M M S PUMP M PT 601B 8809B 600B RHR B PT SUCTION TE 604B FIS 602B FCV-602B M CONTAINMENT HOT LEG LOOP C 8812A PT SUMP 8701B 402 M 8701A M M 8811A M PRT 8708A PT 8809A S M 601A PUMP RHR A PT 600A FCV-602A SUCTION 604A TE 602A FIS M HCV-142 LINE CONTAINMENT INSIDE RHR HX B HCV-603B FCV-605B CCW 8706B M 606B TE CONTAINMENT OUTSIDE 8881 FIS 605B 8887B M 8888B M 8864B PRT COLD LEG LOOP C COLD LEG HOT LEG A & B LOOPS PRT LETDOWN CVCS FCV-605A RHR HX A HCV-603A 8706A CCW M TE 606A M 8889 M 8865 M 8887A 8888A FIS 605A PRT 8864A LOOP B COLD LEG LOOP A TANK STORAGE WATER REFUELING SAMPLE SYS SAMPLE SYS 5 Farley Nuclear Plant RHR ACI Deletion
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
Issue with RHR ACI oAEOD report "Decay Heat Removal Problems at U.S. Pressurized Water Reactors" dated December 1985, identifies 130 loss of RHR events in US PWRs between 1976 and 1983 o37 were caused by automatic closure of the RHR suction/isolation valves oClosure of the RHR valves results in a loss of cooling during shutdown (low pressure) operation oFarley Unit 1 experienced a loss of SDC in 2010 due to an inadvertent actuation of the RHR ACI during testing   6 Farley Nuclear Plant RHR ACI Deletion
* Issue with RHR ACI o AEOD report Decay Heat Removal Problems at U.S. Pressurized Water Reactors dated December 1985, identifies 130 loss of RHR events in US PWRs between 1976 and 1983 o 37 were caused by automatic closure of the RHR suction/isolation valves o Closure of the RHR valves results in a loss of cooling during shutdown (low pressure) operation o Farley Unit 1 experienced a loss of SDC in 2010 due to an inadvertent actuation of the RHR ACI during testing 6
 
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
NRC internal Memo on RHR ACI dated January 1985 stated:
* NRC internal Memo on RHR ACI dated January 1985 stated:
oA request to remove the ACI feature should be substantiated by proof that the change is a net improvement to safety 7
o A request to remove the ACI feature should be substantiated by proof that the change is a net improvement to safety 7
 
Farley Nuclear Plant RHR ACI Deletion
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
PWROG (WOG) Program  
* PWROG (WOG) Program - RHR System ACI Removal o WCAP-11736-A, Residual Heat Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners Group o Sorted Westinghouse NSSS plants into four groups o Provided assessments for four reference plants to demonstrate the change is a net improvement to safety Interfacing system LOCA analysis RHR unavailability analysis LTOP/overpressurization analysis 8
- RHR System ACI Removal oWCAP-11736-A, "Residual Heat Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners Group" oSorted Westinghouse NSSS plants into four groups oProvided assessments for four reference plants to demonstrate the change is a net improvement to safety Interfacing system LOCA analysis RHR unavailability analysis LTOP/overpressurization analysis 8
 
Farley Nuclear Plant RHR ACI Deletion
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
NRC SE on WCAP
* NRC SE on WCAP-11736-A: Staff Position o Removal of ACI for W NSSS plants can produce a net safety benefit provided five key improvements are in place Alarm on each RHR suction valve if the valve is open and pressure greater than pressure permissive Valve position indication Procedural improvements Power removed from RHR suction valves Sizing of RHR valve operators so valve cannot be opened against full system pressure o The WCAP can be referenced in licensees plant-specific submittals to show compliance with items that are not plant specific 9
-11736-A: Staff Position oRemoval of ACI for W NSSS plants can produce a net safety benefit provided five key improvements are in place Alarm on each RHR suction valve if the valve is open and pressure greater than pressure permissive Valve position indication Procedural improvements Power removed from RHR suction valves Sizing of RHR valve operators so valve cannot be opened against full system pressure oThe WCAP can be referenced in licensee's plant
 
-specific submittals to show compliance with items that are not plant specific 9
Farley Nuclear Plant RHR ACI Deletion
Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
NRC SE on WCAP
* NRC SE on WCAP-11736-A o Section 2.4 - The effects of ACI removal upon plant safety must be evaluated on a plant-by-plant basis because of numerous plant-specific differences o Section 2.6 - The licensee should do sufficient PRA and safety analysis to ensure that its plant will not show results that will invalidate the conclusions of WCAP-11736-A o Requires submitting a LAR for the Tech Spec change o  Delete SR 3.4.14.2 in Tech Spec 3.4.14, RCS PIV Leakage, Verify RHR System autoclosure interlock causes the valves to close automatically with a simulated or actual RCS pressure signal  700 psig and  750 psig.
-11736-A oSection 2.4  
10
- The effects of ACI removal upon plant safety must be evaluated on a plant
 
-by-plant basis because of numerous plant
Farley Nuclear Plant RHR ACI Deletion
-specific differences oSection 2.6  
- The licensee should do sufficient PRA and safety analysis to ensure that its plant will not show results that will invalidate the conclusions of WCAP-11736-A oRequires submitting a LAR for the Tech Spec change oDelete SR 3.4.14.2 in Tech Spec 3.4.14, "RCS PIV Leakage,"
"Verify RHR System autoclosure interlock causes the valves to close automatically with a simulated or actual ."        10 Farley Nuclear Plant RHR ACI Deletion


===Background===
===Background===
Previous Farley RHR ACI Removal Program oWCAP-11746, Rev. 1, "Residual Heat Removal System Autoclosure Interlock Removal Report for the Joseph M. Farley Nuclear Plant Units 1 and 2" (April 1996) documented the justification for RHR ACI deletion at Farley oAddressed NRC SE requirements oProvided plant specific PRA for impact of RHR ACI removal on plant safety Interfacing system LOCA analysis RHR unavailability analysis Low temperature overpressurization analysis   11 Farley Nuclear Plant RHR ACI Deletion Issue for Discussion The Farley Units 1 and 2 specific analysis documented in WCAP
* Previous Farley RHR ACI Removal Program o WCAP-11746, Rev. 1, Residual Heat Removal System Autoclosure Interlock Removal Report for the Joseph M. Farley Nuclear Plant Units 1 and 2 (April 1996) documented the justification for RHR ACI deletion at Farley o Addressed NRC SE requirements o Provided plant specific PRA for impact of RHR ACI removal on plant safety Interfacing system LOCA analysis RHR unavailability analysis Low temperature overpressurization analysis 11
-11746, Rev. 1 was completed in 1996 The PRA analysis may not meet RG 1.200 and may not be acceptable to NRC Staff reviewers Potential issues with:
 
oFault trees oData oHuman reliability analysis oEvent frequencies 12 Farley Nuclear Plant RHR ACI Deletion Proposed Approach Step 1: NRC meeting to discuss the proposed approach and obtain the Staff's feedback Step 2a: Assess the technical adequacy of the WCAP-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200 oInterfacing system LOCA analysis - metric: frequency change oRHR unavailability analysis - metric: unavailability change oLow temperature overpressurization analysis - metric: consequence category (non
Farley Nuclear Plant RHR ACI Deletion Issue for Discussion
-success end states)     13 Farley Nuclear Plant RHR ACI Deletion Proposed Approach Step 2b: Categorize the deficiencies in meeting the PRA Standards 1.Conservative aspect of the model that does not need to be addressed 2.No impact on the decision
* The Farley Units 1 and 2 specific analysis documented in WCAP-11746, Rev. 1 was completed in 1996
-making process 3.Could impact the results, but can be addressed via high level quantitative or qualitative assessment 4.Impacts a key aspect of the analysis/models 14 Farley Nuclear Plant RHR ACI Deletion Proposed Approach Step 3: Model Changes and Quantification oDeficiencies will be addressed and these may be addressed with:
* The PRA analysis may not meet RG 1.200 and may not be acceptable to NRC Staff reviewers
Model changes Qualitative assessments Sensitivity analyses oData from most recent Farley PRA model will be used oAll models will be re
* Potential issues with:
-quantified to demonstrate the acceptability of the RHR ACI deletion 15 Farley Nuclear Plant RHR ACI Deletion Summary and Conclusions Assess the technical adequacy of the WCAP
o Fault trees o Data o Human reliability analysis o Event frequencies 12
-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200 Categorize the deficiencies in meeting the PRA Standards Model important changes and complete quantification Submit a LAR supporting the Tech Spec Change Identify how each SE condition for WCAP
 
-11736-A is met for Farley Units 1 and 2 Update supporting PRA analysis previously performed in WCAP
Farley Nuclear Plant RHR ACI Deletion Proposed Approach
-11746, Rev. 1 Tentative Submittal Date: December 2014 16 NRC Feedback/Questions or Comments? 17}}
* Step 1: NRC meeting to discuss the proposed approach and obtain the Staffs feedback
* Step 2a: Assess the technical adequacy of the WCAP-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200 o Interfacing system LOCA analysis - metric:
frequency change o RHR unavailability analysis - metric: unavailability change o Low temperature overpressurization analysis -
metric: consequence category (non-success end states) 13
 
Farley Nuclear Plant RHR ACI Deletion Proposed Approach
* Step 2b: Categorize the deficiencies in meeting the PRA Standards
: 1. Conservative aspect of the model that does not need to be addressed
: 2. No impact on the decision-making process
: 3. Could impact the results, but can be addressed via high level quantitative or qualitative assessment
: 4. Impacts a key aspect of the analysis/models 14
 
Farley Nuclear Plant RHR ACI Deletion Proposed Approach
* Step 3: Model Changes and Quantification o Deficiencies will be addressed and these may be addressed with:
Model changes Qualitative assessments Sensitivity analyses o Data from most recent Farley PRA model will be used o All models will be re-quantified to demonstrate the acceptability of the RHR ACI deletion 15
 
Farley Nuclear Plant RHR ACI Deletion Summary and Conclusions
* Assess the technical adequacy of the WCAP-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200
* Categorize the deficiencies in meeting the PRA Standards
* Model important changes and complete quantification
* Submit a LAR supporting the Tech Spec Change
* Identify how each SE condition for WCAP-11736-A is met for Farley Units 1 and 2
* Update supporting PRA analysis previously performed in WCAP-11746, Rev. 1           16
* Tentative Submittal Date: December 2014
 
NRC Feedback/Questions or Comments?
17}}

Revision as of 05:14, 4 November 2019

April 23, 2014, Category 1 Public Meeting, J.M. Farley Presentation Slides, Residual Heat Removal Autoclosure Interlock Deletion.
ML14120A225
Person / Time
Site: Farley 
Issue date: 04/23/2014
From:
Southern Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation
Williams S
References
Download: ML14120A225 (17)
v • d • e


Text

Farley Nuclear Plant Residual Heat Removal Autoclosure Interlock Deletion April 23, 2014 1

Farley Nuclear Plant RHR ACI Deletion Agenda

  • Introductions
  • Purpose of Meeting
  • Background
  • Issue for Discussion
  • Proposed Approach
  • Summary and Conclusions 2

Farley Nuclear Plant RHR ACI Deletion Purpose of Meeting Discuss the approach and obtain NRC feedback and expectations on the technical justification for the elimination of the Residual Heat Removal (RHR) System Autoclosure Interlock (ACI).

3

Farley Nuclear Plant RHR ACI Deletion

Background

  • RHR Autoclosure Interlock o Ensures a double isolation valve barrier between the RCS and RHR system when the plant is at normal operating conditions o The double isolation valve barrier provides protection against overpressurizing the low pressure RHR system by the high pressure RCS o Both RHR suction isolation valves close automatically if the pressure increases above the bistable setpoint o Helps to prevent interfacing system LOCAs 4

OpsRhr051 M

FCV-602A 8864A M PRT CHARGING PUMP 8708A SUCTION SAMPLE 8706A SYS PRT FIS FIS PT 602A TE CCW TE 605A PT PT 600A 604A 606A M 402 M M 601A LOOP A LOOP C A COLD LEG RHR 8888A HOT LEG RHR HX A HCV-603A 8701B 8701A S PUMP M

PRT 8887A M M REFUELING 8812A 8809A M M WATER 8865 STORAGE FCV-605A LOOPS CONTAINMENT TANK A&B SUMP 8889 8811A HOT LEG CVCS LETDOWN LOOP B HCV-142 LINE 8881 M COLD LEG 8864B PRT FCV-602B M

CHARGING PUMP 8708B M SUCTION SAMPLE 8706B PRT 8809B FIS SYS FIS PT 602B TE CCW TE 605B PT 600B 604B 606B M M PT 403 M M 601B 8887B LOOP C LOOP A B COLD LEG RHR 8888B HOT LEG RHR HX B HCV-603B 8702B 8702A S PUMP M

8812B M FCV-605B CONTAINMENT SUMP 8811B INSIDE OUTSIDE OUTSIDE INSIDE CONTAINMENT CONTAINMENT CONTAINMENT CONTAINMENT 5

Farley Nuclear Plant RHR ACI Deletion

Background

  • Issue with RHR ACI o AEOD report Decay Heat Removal Problems at U.S. Pressurized Water Reactors dated December 1985, identifies 130 loss of RHR events in US PWRs between 1976 and 1983 o 37 were caused by automatic closure of the RHR suction/isolation valves o Closure of the RHR valves results in a loss of cooling during shutdown (low pressure) operation o Farley Unit 1 experienced a loss of SDC in 2010 due to an inadvertent actuation of the RHR ACI during testing 6

Farley Nuclear Plant RHR ACI Deletion

Background

  • NRC internal Memo on RHR ACI dated January 1985 stated:

o A request to remove the ACI feature should be substantiated by proof that the change is a net improvement to safety 7

Farley Nuclear Plant RHR ACI Deletion

Background

  • PWROG (WOG) Program - RHR System ACI Removal o WCAP-11736-A, Residual Heat Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners Group o Sorted Westinghouse NSSS plants into four groups o Provided assessments for four reference plants to demonstrate the change is a net improvement to safety Interfacing system LOCA analysis RHR unavailability analysis LTOP/overpressurization analysis 8

Farley Nuclear Plant RHR ACI Deletion

Background

  • NRC SE on WCAP-11736-A: Staff Position o Removal of ACI for W NSSS plants can produce a net safety benefit provided five key improvements are in place Alarm on each RHR suction valve if the valve is open and pressure greater than pressure permissive Valve position indication Procedural improvements Power removed from RHR suction valves Sizing of RHR valve operators so valve cannot be opened against full system pressure o The WCAP can be referenced in licensees plant-specific submittals to show compliance with items that are not plant specific 9

Farley Nuclear Plant RHR ACI Deletion

Background

  • NRC SE on WCAP-11736-A o Section 2.4 - The effects of ACI removal upon plant safety must be evaluated on a plant-by-plant basis because of numerous plant-specific differences o Section 2.6 - The licensee should do sufficient PRA and safety analysis to ensure that its plant will not show results that will invalidate the conclusions of WCAP-11736-A o Requires submitting a LAR for the Tech Spec change o Delete SR 3.4.14.2 in Tech Spec 3.4.14, RCS PIV Leakage, Verify RHR System autoclosure interlock causes the valves to close automatically with a simulated or actual RCS pressure signal 700 psig and 750 psig.

10

Farley Nuclear Plant RHR ACI Deletion

Background

  • Previous Farley RHR ACI Removal Program o WCAP-11746, Rev. 1, Residual Heat Removal System Autoclosure Interlock Removal Report for the Joseph M. Farley Nuclear Plant Units 1 and 2 (April 1996) documented the justification for RHR ACI deletion at Farley o Addressed NRC SE requirements o Provided plant specific PRA for impact of RHR ACI removal on plant safety Interfacing system LOCA analysis RHR unavailability analysis Low temperature overpressurization analysis 11

Farley Nuclear Plant RHR ACI Deletion Issue for Discussion

  • The Farley Units 1 and 2 specific analysis documented in WCAP-11746, Rev. 1 was completed in 1996
  • The PRA analysis may not meet RG 1.200 and may not be acceptable to NRC Staff reviewers
  • Potential issues with:

o Fault trees o Data o Human reliability analysis o Event frequencies 12

Farley Nuclear Plant RHR ACI Deletion Proposed Approach

  • Step 1: NRC meeting to discuss the proposed approach and obtain the Staffs feedback
  • Step 2a: Assess the technical adequacy of the WCAP-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200 o Interfacing system LOCA analysis - metric:

frequency change o RHR unavailability analysis - metric: unavailability change o Low temperature overpressurization analysis -

metric: consequence category (non-success end states) 13

Farley Nuclear Plant RHR ACI Deletion Proposed Approach

  • Step 2b: Categorize the deficiencies in meeting the PRA Standards
1. Conservative aspect of the model that does not need to be addressed
2. No impact on the decision-making process
3. Could impact the results, but can be addressed via high level quantitative or qualitative assessment
4. Impacts a key aspect of the analysis/models 14

Farley Nuclear Plant RHR ACI Deletion Proposed Approach

  • Step 3: Model Changes and Quantification o Deficiencies will be addressed and these may be addressed with:

Model changes Qualitative assessments Sensitivity analyses o Data from most recent Farley PRA model will be used o All models will be re-quantified to demonstrate the acceptability of the RHR ACI deletion 15

Farley Nuclear Plant RHR ACI Deletion Summary and Conclusions

  • Assess the technical adequacy of the WCAP-11746, Rev. 1 models/analyses against ASME/ANS Standards and RG 1.200
  • Categorize the deficiencies in meeting the PRA Standards
  • Model important changes and complete quantification
  • Submit a LAR supporting the Tech Spec Change
  • Identify how each SE condition for WCAP-11736-A is met for Farley Units 1 and 2
  • Update supporting PRA analysis previously performed in WCAP-11746, Rev. 1 16
  • Tentative Submittal Date: December 2014

NRC Feedback/Questions or Comments?

17

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