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Sent throuch www.intcrmedia.com 3
Sent throuch www.intcrmedia.com 3


Victor M . McCree - NSIAC Remarks- October 12, 2016 KEY MESSAGES
Victor M . McCree - NSIAC Remarks- October 12, 2016 KEY MESSAGES Non-Responsive Record BACKGROUND Stakeholder Meeting Follow-Up
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* Non-Responsive Record Project AIM Non-Responsive Record 1
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* Non-Responsive Record
* BACKGROUND Stakeholder Meeting Follow-Up
*
* Non-Responsive Record
* Project AIM
*
* Non-Responsive Record
* 1


Licensing Action Initiatives
Licensing Action Initiatives Non-Responsive Record Byron/Braidwood Backfit Appeal Decision
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*
*
* Non-Responsive Record
* Byron/Braidwood Backfit Appeal Decision
* The EDO concluded that the NRC staffs position in the October 2015 backfit issued to Byron and Braidwood related to pressurizer valve performance was a new or modified interpretation of what constitutes compliance and did not provide a basis for a compliance backfit.
* The EDO concluded that the NRC staffs position in the October 2015 backfit issued to Byron and Braidwood related to pressurizer valve performance was a new or modified interpretation of what constitutes compliance and did not provide a basis for a compliance backfit.
* This decision was communicated on 9/15/16 in publicly available letters to Exelon and NEI, as well as in a memo to the staff tha,t also requested preparation of a plan to reevaluate the generic implications of the technical issue.
* This decision was communicated on 9/15/16 in publicly available letters to Exelon and NEI, as well as in a memo to the staff tha,t also requested preparation of a plan to reevaluate the generic implications of the technical issue.
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Subsequent License Renewal
Subsequent License Renewal
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* Non-Responsive Record Operator Licensing Non-Responsive Record Digital l&C Action Plan Non-Responsive Record 50.46c Rulemakin~
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Non-Responsive Record 3
* Non-Responsive Record
* Operator Licensing
* Non-Responsive Record Digital l&C Action Plan
*
* Non-Responsive Record
* 50.46c Rulemakin~
*
* Non-Responsive Record 3


NUREG-2180 / Very Early Warning Fire Detection Systems
NUREG-2180 / Very Early Warning Fire Detection Systems Non-Responsive Record Force on Force Efficiencies/Guidance Non-Responsive Record Cybersecurlty at Nuclear Power Plants Non-Responsive Record Advanced Reactor Preparations
* Non-Responsive Record
* Non-Responsive Record 4
* Force on Force Efficiencies/Guidance
* Non-Responsive Record
*
* Cybersecurlty at Nuclear Power Plants
*
*
* Non-Responsive Record
*
* Advanced Reactor Preparations
*
* Non-Responsive Record
* 4


Ne.ru.........,~....IL!.:~.!.:£!.in:..:.e..~A~t~iv~it~i~e~s~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Ne.ru.........,~....IL!.:~.!.:£!.in:..:.e..~A~t~iv~it~i~e~s~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
*
Non-Responsive Record 5
* Non-Responsive Record
*
* 5


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_ _ _ _....., backfit (once at the office level, then at NEXT ST!;PS                              the EDO level) claiming it was
_ _ _ _....., backfit (once at the office level, then at NEXT ST!;PS                              the EDO level) claiming it was
                                                                                                   ----  - - - - - l  inappropriate since, in their view, the staff d                  failed to identify any error or omission mber 13,            that make the previously approved analysis incorrect.
                                                                                                   ----  - - - - - l  inappropriate since, in their view, the staff d                  failed to identify any error or omission mber 13,            that make the previously approved analysis incorrect.
ector, NRR,          The NRC regulation for backfitting eptember 13,        ( 1O CFR 50.109) indicates that "the compliance exception is intended to address situations where the licensee see (Exelon)          has failed to meet known and established
ector, NRR,          The NRC regulation for backfitting eptember 13,        ( 10 CFR 50.109) indicates that "the compliance exception is intended to address situations where the licensee see (Exelon)          has failed to meet known and established
' constitutes compliance in addressing              of this decision. Date: September 13, 2016 (pm)                  standards of the Commission because of potential pressurizer safety valve failures                                                                        omission or mistake of fact.. .. new or following water discharge, and did not          OPA, in coordination with OEDO, will issue a news release          modified interpretations of what provide a basis for a compliance backfit.        (preferably) or BLOG post as an alternative, announcing          constitutes compliance would not fall this decision: September 13, 2016 (pm)                          . within the exception .... "
' constitutes compliance in addressing              of this decision. Date: September 13, 2016 (pm)                  standards of the Commission because of potential pressurizer safety valve failures                                                                        omission or mistake of fact.. .. new or following water discharge, and did not          OPA, in coordination with OEDO, will issue a news release          modified interpretations of what provide a basis for a compliance backfit.        (preferably) or BLOG post as an alternative, announcing          constitutes compliance would not fall this decision: September 13, 2016 (pm)                          . within the exception .... "
Previously (on June ~2:. 2016) the EDO tasked the Committee to Review Generic          The EDO will send a letter to NEI in response to its earlier
Previously (on June ~2:. 2016) the EDO tasked the Committee to Review Generic          The EDO will send a letter to NEI in response to its earlier
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Communications Message Map:                        EDO EXELON BACKFIT APPEAL OECISJOt,i:--                        SEPTEMBER 12, 2016 KEY MESSAGES                                        NRC ACTIONS/ACTIVITIES                                  BACKGROUND
Communications Message Map:                        EDO EXELON BACKFIT APPEAL OECISJOt,i:--                        SEPTEMBER 12, 2016 KEY MESSAGES                                        NRC ACTIONS/ACTIVITIES                                  BACKGROUND
!
*--------------------+----------------------
*--------------------+----------------------
. The staff takes its responsibility for          On June 22, 2016, the EDO established a Backfit Appeal        The NRG staff issued a compliance assuring safety very seriously; and            Review Panel (Panel) of senior staff and managers to          backfit letter to E)(elon (October 9, 2015) pursues technically sound, and legally        review the Exelon backfit appeal.                            on the issue of pressurizer overfill and well-founded backfits when it concludes                                                                      safety valve performance during they are needed to assure or enhance            On August 24, 2016, the Panel recommended, and the            Condition II events (ANS Condition II safety_                                        EDO supported, a reversal of the compliance backfit,          categorization as frequent events).
. The staff takes its responsibility for          On June 22, 2016, the EDO established a Backfit Appeal        The NRG staff issued a compliance assuring safety very seriously; and            Review Panel (Panel) of senior staff and managers to          backfit letter to E)(elon (October 9, 2015) pursues technically sound, and legally        review the Exelon backfit appeal.                            on the issue of pressurizer overfill and well-founded backfits when it concludes                                                                      safety valve performance during they are needed to assure or enhance            On August 24, 2016, the Panel recommended, and the            Condition II events (ANS Condition II safety_                                        EDO supported, a reversal of the compliance backfit,          categorization as frequent events).
                                                 . agreeing with the Exelon appeal.
                                                 . agreeing with the Exelon appeal.
On complex technical and legal matters                                                                      Exelon twice appealed the compliance there can be differing views either within the  staff, or with licensees stakeholders. The NRC used and  other its formal
On complex technical and legal matters                                                                      Exelon twice appealed the compliance there can be differing views either within the  staff, or with licensees stakeholders. The NRC used and  other its formal NEXT STEPS
                                                  . _____
NEXT STEPS
                                                                                                         ----1 Ibackfit (once at the office level, then at the EDO level) claiming it was inappropriate since, in their view, the staff backfit review process to ensure      this    The    EDO will verbally inform the Chairman and            failed to identify any error or omission issue received appropriate consideration. Commissioners of this decision. Date: September 13,                that make the previously approved 2016 (am)                                                    analysis incorrect.
                                                                                                         ----1 Ibackfit (once at the office level, then at the EDO level) claiming it was inappropriate since, in their view, the staff backfit review process to ensure      this    The    EDO will verbally inform the Chairman and            failed to identify any error or omission issue received appropriate consideration. Commissioners of this decision. Date: September 13,                that make the previously approved 2016 (am)                                                    analysis incorrect.
In this case, the EDD concluded that the NRC staff's position on valve                  The EDO will send a memorandum to the Director, NRR,          The NRC regulation for backfitting qualification, valve performance, and the      formally notifylng him of this decision. Date: September 13,    a (1 CFR 50.109) indicates that "the application of the single failure criterion    2016 (am)                                                    compliance exception is intended to in the backfit safety evaluation was a                                                                      address situations where the licensee        !
In this case, the EDD concluded that the NRC staff's position on valve                  The EDO will send a memorandum to the Director, NRR,          The NRC regulation for backfitting qualification, valve performance, and the      formally notifylng him of this decision. Date: September 13,    a (1 CFR 50.109) indicates that "the application of the single failure criterion    2016 (am)                                                    compliance exception is intended to in the backfit safety evaluation was a                                                                      address situations where the licensee        !
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From:                                    West, Steven Sent:                                    Tuesday, December 06, 2016 8:53 AM To:                                      Johnson, Michael Cc:                                      Inverso, Tara; Clark, Theresa; Holian, Brian; Bowen, Jeremy
From:                                    West, Steven Sent:                                    Tuesday, December 06, 2016 8:53 AM To:                                      Johnson, Michael Cc:                                      Inverso, Tara; Clark, Theresa; Holian, Brian; Bowen, Jeremy
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From:                                          Clark, Theresa Sent:                                          Thursday, September 15, 2016 9:54 AM To:                                            Valliere, Nanette; Ruesch, Eric; Castleman, Patrick; Frazier, Alan; Krsek, Robert Cc:                                            Lewis, Robert; Rasouli, Houman; Inverso, Tara; Bowen, Jeremy; Holahan, Gary
From:                                          Clark, Theresa Sent:                                          Thursday, September 15, 2016 9:54 AM To:                                            Valliere, Nanette; Ruesch, Eric; Castleman, Patrick; Frazier, Alan; Krsek, Robert Cc:                                            Lewis, Robert; Rasouli, Houman; Inverso, Tara; Bowen, Jeremy; Holahan, Gary


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Accession No. ML16173A311). On August 24, 2016, the Panel transmitted the results of its review to me (ADAMS Accession No. ML16236A202}. The memorandum from the Panel responding to my tasking , recommended that the 2015 compliance backfit be withdrawn, and included the Panel's report and the basis for this recommendation (ADAMS Accession No. ML16236A208}.
Accession No. ML16173A311). On August 24, 2016, the Panel transmitted the results of its review to me (ADAMS Accession No. ML16236A202}. The memorandum from the Panel responding to my tasking , recommended that the 2015 compliance backfit be withdrawn, and included the Panel's report and the basis for this recommendation (ADAMS Accession No. ML16236A208}.
I have reviewed the Panel's report, its recommendations , and its responses to the questions I posed when establishing the panel. In addition, I met with you on September 12, 2016, to discuss my decision and assure that it reflects the thorough, technically sound, and legally well-founded consideration that this matter merits. Our discussion included my response to the additional perspectives you provided to me in your email dated September 2, 2016, which is enclosed, for reference.
I have reviewed the Panel's report, its recommendations , and its responses to the questions I posed when establishing the panel. In addition, I met with you on September 12, 2016, to discuss my decision and assure that it reflects the thorough, technically sound, and legally well-founded consideration that this matter merits. Our discussion included my response to the additional perspectives you provided to me in your email dated September 2, 2016, which is enclosed, for reference.
As we discussed, the central question in the a1:31:3eal p,Eanel's review was whether an adequate basis exists for backfitting using the compliance exception in Title 10 of the Code of Federal Regulations (1 OCFR}, Section 50.109(a)4(i) to address potential pressurizer safety valve failures following water discharge. With regard to compliance, the 1985 statement of considerations for 1O CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact.. .. new or modified interpretations of what constitutes compliance would not fall within the exception ... ." In answering this question, the Panel focused on the following three related technical and regulatory positions for the pressurizer safety valves (PSVs} described in the staff's October 9, 2015, safety evaluation CONTACT:        Gary M. Holahan, OEDO (301) 415-1765
As we discussed, the central question in the a1:31:3eal p,Eanel's review was whether an adequate basis exists for backfitting using the compliance exception in Title 10 of the Code of Federal Regulations (1 OCFR}, Section 50.109(a)4(i) to address potential pressurizer safety valve failures following water discharge. With regard to compliance, the 1985 statement of considerations for 10 CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact.. .. new or modified interpretations of what constitutes compliance would not fall within the exception ... ." In answering this question, the Panel focused on the following three related technical and regulatory positions for the pressurizer safety valves (PSVs} described in the staff's October 9, 2015, safety evaluation CONTACT:        Gary M. Holahan, OEDO (301) 415-1765


W. Dean                                            imposing the back.fit (ADAMS Accession No. ML14225A871, referred to as the Backfit SE), as well as the staff's May 3, 2016, response (ADAMS Accession No. ML16095A204) to the backfit appeal by Exelon Generation Company, LLC (the licensee):
W. Dean                                            imposing the back.fit (ADAMS Accession No. ML14225A871, referred to as the Backfit SE), as well as the staff's May 3, 2016, response (ADAMS Accession No. ML16095A204) to the backfit appeal by Exelon Generation Company, LLC (the licensee):
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Response: ASME BPV Code does indeed provide certification requirements for safety and relief valves for their intended design function. However, as noted in the Panel's report, since the TMl-2 accident, the NRC has accepted qualification of safety and relief valves based on EPRI, Wyle, and vendor testing to demonstrate that these valves will not stick open during water discharge as part of TMI action items, Chapter 15 accident analyses, or other evaluations for numerous nuclear power plants. Therefore, a longstanding NRC precedent for the acceptance of water qualification of safety and relief valves through such testing has been established in those evaluations. Based on infonnation provided by NRR and a sampling review by the Panel, the NRG has not required safety or relief valves to be certified by the ASME BPV Code for water service when referenced Chapter 15 accident analyses. A requirement at this time that all safety and relief valves be certified for water service in accordance with the ASME BPV Code for every reference to safety and relief valves not sticking open upon water discharge for all TMI action items, Chapter 15 accident analyses, and other evaluations would constitute a significant regulatory action and warrant a different decision-making process
Response: ASME BPV Code does indeed provide certification requirements for safety and relief valves for their intended design function. However, as noted in the Panel's report, since the TMl-2 accident, the NRC has accepted qualification of safety and relief valves based on EPRI, Wyle, and vendor testing to demonstrate that these valves will not stick open during water discharge as part of TMI action items, Chapter 15 accident analyses, or other evaluations for numerous nuclear power plants. Therefore, a longstanding NRC precedent for the acceptance of water qualification of safety and relief valves through such testing has been established in those evaluations. Based on infonnation provided by NRR and a sampling review by the Panel, the NRG has not required safety or relief valves to be certified by the ASME BPV Code for water service when referenced Chapter 15 accident analyses. A requirement at this time that all safety and relief valves be certified for water service in accordance with the ASME BPV Code for every reference to safety and relief valves not sticking open upon water discharge for all TMI action items, Chapter 15 accident analyses, and other evaluations would constitute a significant regulatory action and warrant a different decision-making process
: f. The panel asserts in its summary that the valves in question were water qualified due to the licensee's reliance on them to pass water during feedline break events.
: f. The panel asserts in its summary that the valves in question were water qualified due to the licensee's reliance on them to pass water during feedline break events.
The panel does not appear to acknowledge that feedline breaks are Condition IV events, similar to [loss-of-coolant accidents], which are never expected to occur in the lifetime of the facilities and therefore, given their lower probability of occurrence, are permitted to have more significant consequences. The EPRI testing demonstrated acceptable performance under conditions anticipated during these Condition IV events (higher temperature fluid - 650&deg;F), while the EPRI
The panel does not appear to acknowledge that feedline breaks are Condition IV events, similar to [loss-of-coolant accidents], which are never expected to occur in the lifetime of the facilities and therefore, given their lower probability of occurrence, are permitted to have more significant consequences. The EPRI testing demonstrated acceptable performance under conditions anticipated during these Condition IV events (higher temperature fluid - 650&deg;F), while the EPRI test at the more likely Condition II inadvertent mass addition event conditions (lower temperature fluid -sso*F) was terminated early due to valve chatter on opening. The summary of the EPRI testing indicated that for subcooled water conditions valve chatter and resultant valve damage was generally observed.
 
test at the more likely Condition II inadvertent mass addition event conditions (lower temperature fluid -sso*F) was terminated early due to valve chatter on opening. The summary of the EPRI testing indicated that for subcooled water conditions valve chatter and resultant valve damage was generally observed.
Response: In evaluating information provided by NRR associated with water testing, NRR maintained that the EPRI testing did not address water discharge for the Byron and Braidwood PSVs. The current question recognizes the EPRI testing, but asserts that the water was not at an acceptable temperature, or the water relief might result in valve chatter or damage. Based on the Panel's document review and discussions, the NRC staff approval of the amendments during the 2001 and 2004 reviews evaluated the EPRI testing, and were indeed aware of the test, its results and water temperatures.
Response: In evaluating information provided by NRR associated with water testing, NRR maintained that the EPRI testing did not address water discharge for the Byron and Braidwood PSVs. The current question recognizes the EPRI testing, but asserts that the water was not at an acceptable temperature, or the water relief might result in valve chatter or damage. Based on the Panel's document review and discussions, the NRC staff approval of the amendments during the 2001 and 2004 reviews evaluated the EPRI testing, and were indeed aware of the test, its results and water temperatures.
: 3. Path Forward
: 3. Path Forward
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: b. If the ECO supports the Backfit Panel's conclusion, NRR requests that the EDO allow the staff to independently assess what path forward is appropriate given the positions documented in the panel's report and EDO's decision. In particular, NRR has concerns regarding the recommendations on page 3 of the report that need to be further considered before determining what future course of action is most appropriate.
: b. If the ECO supports the Backfit Panel's conclusion, NRR requests that the EDO allow the staff to independently assess what path forward is appropriate given the positions documented in the panel's report and EDO's decision. In particular, NRR has concerns regarding the recommendations on page 3 of the report that need to be further considered before determining what future course of action is most appropriate.
Response: I agree. The report reveals the need to assess the treatment of the underlying technical issue described in the 1993 Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013) on PSV performance after water discharge at pressurized-water reactors. In addition, given the decision communicated herein, the positions included in Regulatory Issue Summary 2005-29, as well as its proposed Revision 1, should be (re)assessed through the appropriate generic process to ensure they receive appropriate backfit consideration. The Director of NRR should inform me within 30 days of the plan to respond to these issues.
Response: I agree. The report reveals the need to assess the treatment of the underlying technical issue described in the 1993 Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013) on PSV performance after water discharge at pressurized-water reactors. In addition, given the decision communicated herein, the positions included in Regulatory Issue Summary 2005-29, as well as its proposed Revision 1, should be (re)assessed through the appropriate generic process to ensure they receive appropriate backfit consideration. The Director of NRR should inform me within 30 days of the plan to respond to these issues.
From:                              McCree, Victor Sent:                              Sunday, September 11, 2016 7:06 PM To:                                Holahan, Gary; Johnson, Michael; Tracy, Glenn; Clark, Theresa
From:                              McCree, Victor Sent:                              Sunday, September 11, 2016 7:06 PM To:                                Holahan, Gary; Johnson, Michael; Tracy, Glenn; Clark, Theresa


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Specifically, inadvertent operation of the emergency core cooling system, malfunction of the chemical and volume control system, and inadvertent opening of a pressurizer safety or relief valve.
Specifically, inadvertent operation of the emergency core cooling system, malfunction of the chemical and volume control system, and inadvertent opening of a pressurizer safety or relief valve.
6 For consistency in this report, the Panel uses the phrase "water discharge" rather than "water relief' or "liquid discharge" (except in direct quotes), as this is the phrase used in the Westinghouse documents that raised the issue addressed in this report.
6 For consistency in this report, the Panel uses the phrase "water discharge" rather than "water relief' or "liquid discharge" (except in direct quotes), as this is the phrase used in the Westinghouse documents that raised the issue addressed in this report.
7 NRC 2001 b - referred to as the Uprate SE in the remainder of the report
7 NRC 2001 b - referred to as the Uprate SE in the remainder of the report staff had twice approved the underlying analysis. 8 The approvals referenced by the licensee were an August 26, 2004, license amendment associated with pressurizer safety valve (PSV) setpoints9 and the above-referenced Uprate SE. In a letter dated May 3, 2016, the NRC responded to the licensee's appeal and reaffirmed its decision that the backfit per the compliance exception provisions of 10 CFR 50.109(a)(4){i) is appropriate. 10 On June 2, 2016, the licensee again appealed the NRG staffs decision, this time to the ED0. 11 The purpose of this report by the Backfit Appeal Review Panel is to provide information and recommendations to support the EDO's decision on the appeal.
 
staff had twice approved the underlying analysis. 8 The approvals referenced by the licensee were an August 26, 2004, license amendment associated with pressurizer safety valve (PSV) setpoints9 and the above-referenced Uprate SE. In a letter dated May 3, 2016, the NRC responded to the licensee's appeal and reaffirmed its decision that the backfit per the compliance exception provisions of 10 CFR 50.109(a)(4){i) is appropriate. 10 On June 2, 2016, the licensee again appealed the NRG staffs decision, this time to the ED0. 11 The purpose of this report by the Backfit Appeal Review Panel is to provide information and recommendations to support the EDO's decision on the appeal.
1.1 Conduct of the Panel's Review In order to establish a technically sound, well informed, and legally defensible basis for its recommendations, the Backfit Appeal Review Panel undertook a review of the relevant documents in this case. This included the licensee and NRG staff letters mentioned above; the Uprate SE and the Setpoint SE; and a June 16, 2016, letter from the Nuclear Energy Institute (NEl)12 supporting the EDO Appeal. The Panel also reviewed many other related documents, which fall into five broad categories:
1.1 Conduct of the Panel's Review In order to establish a technically sound, well informed, and legally defensible basis for its recommendations, the Backfit Appeal Review Panel undertook a review of the relevant documents in this case. This included the licensee and NRG staff letters mentioned above; the Uprate SE and the Setpoint SE; and a June 16, 2016, letter from the Nuclear Energy Institute (NEl)12 supporting the EDO Appeal. The Panel also reviewed many other related documents, which fall into five broad categories:
* The Backfit Rule ( 1O CFR 50.109), related court actions, and Commission and st~ff guidance on application of the Backfit Rule
* The Backfit Rule ( 10 CFR 50.109), related court actions, and Commission and st~ff guidance on application of the Backfit Rule
* Docketed communications for Byron and Braidwood from 1982 to the present, including license amendment requests (LARs) by the licensee, NRG-issued license amendments, NRC requests for additional information (RAls), licensee responses, meeting summaries, NRC SEs, and the licensee's Updated Final Safety Analysis Report (UFSAR}13
* Docketed communications for Byron and Braidwood from 1982 to the present, including license amendment requests (LARs) by the licensee, NRG-issued license amendments, NRC requests for additional information (RAls), licensee responses, meeting summaries, NRC SEs, and the licensee's Updated Final Safety Analysis Report (UFSAR}13
* NRG guidance relevant to the analysis of inadvertent operation of the emergency core cooling system (IOECCS) events over the period of 1981 to the present, including Standard Review Plan (SRP) Section 15.0, Sections 15.5.1 - 15.5.2, and Section 15.6.1 14
* NRG guidance relevant to the analysis of inadvertent operation of the emergency core cooling system (IOECCS) events over the period of 1981 to the present, including Standard Review Plan (SRP) Section 15.0, Sections 15.5.1 - 15.5.2, and Section 15.6.1 14
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as well as docketed communications regarding actions taken by other licensees in response to Westinghouse NSAL-93-013
as well as docketed communications regarding actions taken by other licensees in response to Westinghouse NSAL-93-013
* The history of NRC and industry activities related to power operated relief valves (PORVs), their block valves, and PSVs (including Three Mile Island (TMI) Action Plan s Exelon 2015 - referred to as the NRR Appeal in the remainder of the report 9 NRG 2004b - referred to as the Setpoint SE in the remainder of the report 10 NRG 2016d - referred to as NRR Appeal Decision in the remainder of the  report 11  Exelon 2016a - referred to as EDO Appeal in the remainder of the report 12 NEI 2016 13 Exelon 2002 and Exelon 201 4 (The Panel reviewed other revisions as well, but they are not included in Appendix Das they are not referenced in this report.)
* The history of NRC and industry activities related to power operated relief valves (PORVs), their block valves, and PSVs (including Three Mile Island (TMI) Action Plan s Exelon 2015 - referred to as the NRR Appeal in the remainder of the report 9 NRG 2004b - referred to as the Setpoint SE in the remainder of the report 10 NRG 2016d - referred to as NRR Appeal Decision in the remainder of the  report 11  Exelon 2016a - referred to as EDO Appeal in the remainder of the report 12 NEI 2016 13 Exelon 2002 and Exelon 201 4 (The Panel reviewed other revisions as well, but they are not included in Appendix Das they are not referenced in this report.)
14 NRG 1981 a, NRC 1981b, NRG 1981c, NRC 2007a, NRC 2007b, and NRC 2007c 1s Westinghouse 1993 1s Westinghouse 1994
14 NRG 1981 a, NRC 1981b, NRG 1981c, NRC 2007a, NRC 2007b, and NRC 2007c 1s Westinghouse 1993 1s Westinghouse 1994 Items 11.D.1, 11.0.3, 11.G.1, and 11.K.3 as documented in NUREG-073717, as well as Generic Letter 89-1018 and its supplements), Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREG/CR-703719)
 
Items 11.D.1, 11.0.3, 11.G.1, and 11.K.3 as documented in NUREG-073717, as well as Generic Letter 89-1018 and its supplements), Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREG/CR-703719)
In addition to the document review, the Panel had the benefit of meetings with NRR (both the Division of Safety Systems and the Division of Engineering), the Office of the General Counsel, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon (Bradley Fewell, Senior Vice President of Regulatory Affairs) and NEI (Tony Pietrangelo, Senior Vice President and Chief Nuclear Officer) declined offers for a public meeting, but indicated a willingness to provide information if the Panel identified the need. The Panel did not Identify a need for additional information from either Exelon or NEI to complete the review documented in this report.
In addition to the document review, the Panel had the benefit of meetings with NRR (both the Division of Safety Systems and the Division of Engineering), the Office of the General Counsel, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon (Bradley Fewell, Senior Vice President of Regulatory Affairs) and NEI (Tony Pietrangelo, Senior Vice President and Chief Nuclear Officer) declined offers for a public meeting, but indicated a willingness to provide information if the Panel identified the need. The Panel did not Identify a need for additional information from either Exelon or NEI to complete the review documented in this report.
At the request of the Panel, the Office of Nuclear Regulatory Research (RES) conducted risk analyses using the NRC's Standardized Plant Analysis Risk model for Byron Unit 1.20 These analyses informed the Panel's response to the question from the EDO regarding the risk significance of the relevant accident sequences.
At the request of the Panel, the Office of Nuclear Regulatory Research (RES) conducted risk analyses using the NRC's Standardized Plant Analysis Risk model for Byron Unit 1.20 These analyses informed the Panel's response to the question from the EDO regarding the risk significance of the relevant accident sequences.
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Regarding an IOPORV, the NRC staff stated in Section 3 .3 of the Backfit SE that the licensee had not provided an analysis for the IOPORV that extends long enough into the transient to demonstrate the event would not transition from an ANS Condition II event to an ANS Condition Ill event.
Regarding an IOPORV, the NRC staff stated in Section 3 .3 of the Backfit SE that the licensee had not provided an analysis for the IOPORV that extends long enough into the transient to demonstrate the event would not transition from an ANS Condition II event to an ANS Condition Ill event.
In the Backfit SE, the NRC staff referenced Millstone23 and Callaway24 license amendments as examples of licensees upgrading PORVs for water discharge; a Beaver Valley extended power uprate (EPU) license amendment25 as an example of qualifying PORVs for water discharge; and Turkey Point26 and St. Lucie Unit 227 EPU amendments as additional precedent in support of the backfit decision.
In the Backfit SE, the NRC staff referenced Millstone23 and Callaway24 license amendments as examples of licensees upgrading PORVs for water discharge; a Beaver Valley extended power uprate (EPU) license amendment25 as an example of qualifying PORVs for water discharge; and Turkey Point26 and St. Lucie Unit 227 EPU amendments as additional precedent in support of the backfit decision.
22 NRC 2005b 23 NRC 1998 24 NRC 2000 25 NRC 2006 26 NRC 2012a 27 NRC 2012b
22 NRC 2005b 23 NRC 1998 24 NRC 2000 25 NRC 2006 26 NRC 2012a 27 NRC 2012b In the NRR Appeal, Exelon asserted that the NRC had not justified invoking the compliance exception to the backfit rule. Exelon stated that the NRC approved its IOECCS analysis in both the Uprate SE and the Setpoint SE.
 
In the NRR Appeal, Exelon asserted that the NRC had not justified invoking the compliance exception to the backfit rule. Exelon stated that the NRC approved its IOECCS analysis in both the Uprate SE and the Setpoint SE.
  !In ~crsnMAS6Jthe !NRR Appeal Decisiori[SW7], the NRC staff stated that the previous NRC approvals in 2001 and 2004 were inconsistent with the Agency's general position on the known and established standard at issue- in this case, the progression of ANS Condition II events to higher level events. The NRC staff stated that the fact that the NRC staff were aware of references to EPRI reports on the ability of these non-water qualified PSVs to reseat in certain circumstances was not sufficient to support the licensee's position on the compliance backfit.
  !In ~crsnMAS6Jthe !NRR Appeal Decisiori[SW7], the NRC staff stated that the previous NRC approvals in 2001 and 2004 were inconsistent with the Agency's general position on the known and established standard at issue- in this case, the progression of ANS Condition II events to higher level events. The NRC staff stated that the fact that the NRC staff were aware of references to EPRI reports on the ability of these non-water qualified PSVs to reseat in certain circumstances was not sufficient to support the licensee's position on the compliance backfit.
In the EDO Appeal, Exelon stated that the NRC had misidentified the "known and established standard" at issue as the prohibition of ANS Condition II events progressing to ANS Condition Ill events. Exelon asserted that the standard in question concerns what is necessary to ,,qualify''
In the EDO Appeal, Exelon stated that the NRC had misidentified the "known and established standard" at issue as the prohibition of ANS Condition II events progressing to ANS Condition Ill events. Exelon asserted that the standard in question concerns what is necessary to ,,qualify''
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           ... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position Interpreting the Commission's regulations that is either new or different from a previously applicable staff position ... !.1crairMAS9HSW1o1 Unless one of three specified exceptions apply, the NRC may impose a backfit only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50. 109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection."
           ... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position Interpreting the Commission's regulations that is either new or different from a previously applicable staff position ... !.1crairMAS9HSW1o1 Unless one of three specified exceptions apply, the NRC may impose a backfit only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50. 109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection."
Section 50.109(a)(4) sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2) and (a)(3). The first exception, the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee." The second and third exceptions relate to actions necessary to ensure adequate protection or to actions that involve defining or redefining adequate protection.
Section 50.109(a)(4) sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2) and (a)(3). The first exception, the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee." The second and third exceptions relate to actions necessary to ensure adequate protection or to actions that involve defining or redefining adequate protection.
The Commission explained its intended application of the compliance exception in the Statements of Consideration (SOC) accompanying the 1985 final rule amending 10 CFR 50.109:28 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.
The Commission explained its intended application of the compliance exception in the Statements of Consideration (SOC) accompanying the 1985 final rule amending 10 CFR 50.109:28 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.
In the same SOC, the Commission acknowledged that staff interpretations of rules are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."20 By its terms, the compliance exception applies to actions necessary for compliance with rules, licenses, and orders, or for conformance with written connmitrnents. 30 Also, the Commission explicitly acknowledged the importance of staff interpretations of rules in the regulatory process.
In the same SOC, the Commission acknowledged that staff interpretations of rules are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."20 By its terms, the compliance exception applies to actions necessary for compliance with rules, licenses, and orders, or for conformance with written connmitrnents. 30 Also, the Commission explicitly acknowledged the importance of staff interpretations of rules in the regulatory process.
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, the court's concerns, but did not change the 1985 rule's compliance exception provision. Thus, the quoted statements from the 1985 rule are the applicable expression of Commission intent regarding compliance backfits.
, the court's concerns, but did not change the 1985 rule's compliance exception provision. Thus, the quoted statements from the 1985 rule are the applicable expression of Commission intent regarding compliance backfits.
30 NUREG-1409 (NRC 1990c) defines written commitments broadly to include the "final safety analysis report, licensee event reports, and docketed correspondence, including responses to NRC bulletins, generic letters, inspection reports, or notices of violation and confirmatory action letters."
30 NUREG-1409 (NRC 1990c) defines written commitments broadly to include the "final safety analysis report, licensee event reports, and docketed correspondence, including responses to NRC bulletins, generic letters, inspection reports, or notices of violation and confirmatory action letters."
1.4 A Brief History of Pressurizer Valve Issues Appendix B to this report provides a summary of the NRC and industry's testing, evaluation, and other consideration of PORVs and PSVs since the TMI Unit 2 (TMl-2) accident in 1979. This historical review provides context for discussion of valve "qualification" in the Backfit SE. It also provides the basis for the Panel's conclusions regarding the "known and established standard" for "qualification" in the context of TMI Action Plan Item 11.D.1 and subsequent activities, as well as how it should be interpreted in the Byron and Braidwood licensing basis.
1.4 A Brief History of Pressurizer Valve Issues Appendix B to this report provides a summary of the NRC and industry's testing, evaluation, and other consideration of PORVs and PSVs since the TMI Unit 2 (TMl-2) accident in 1979. This historical review provides context for discussion of valve "qualification" in the Backfit SE. It also provides the basis for the Panel's conclusions regarding the "known and established standard" for "qualification" in the context of TMI Action Plan Item 11.D.1 and subsequent activities, as well as how it should be interpreted in the Byron and Braidwood licensing basis.
In light of the NRC staff's assertion that the licensee had not applied the "single-failure assumption" as noted above, the Panel also considered the applicability of the single failure criterion to PSVs. The Panel expended considerable effort in searching for an answer to what appears to be a simple question: "Are PSVs active components subject to the single failure criterion, or are they passive components exempt from the single failure criterion?" NRR staff have taken the position that PSVs have consistently been treated as active components.
In light of the NRC staff's assertion that the licensee had not applied the "single-failure assumption" as noted above, the Panel also considered the applicability of the single failure criterion to PSVs. The Panel expended considerable effort in searching for an answer to what appears to be a simple question: "Are PSVs active components subject to the single failure criterion, or are they passive components exempt from the single failure criterion?" NRR staff have taken the position that PSVs have consistently been treated as active components.
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The introduction to the GDCs and the related footnote define the applicability of the single failure criterion in terms of electrical versus fluid systems, and active versus passive components. Neither the GDCs nor NRC guidance define which characteristics of passive components are necessary to make a component exempt from the single failure criterion. Some examples are clear: pipes are passive components and pumps and motor-operated valves that operate to perform their safety functions are active components. As discussed in Section 3.6 31 For example, SECY-77-439 (NRC 1977) states: "Examples [of passive failures in fluid systems] include the failure of a simple check valve to move to its correct position when required, the leakage of fluid from failed components, such as pipes and valves- particularly through a failed seal at a valve or pump-or line blockage. Motor-operated valves which have the source of power locked out are allowed to be treated as passive components."
The introduction to the GDCs and the related footnote define the applicability of the single failure criterion in terms of electrical versus fluid systems, and active versus passive components. Neither the GDCs nor NRC guidance define which characteristics of passive components are necessary to make a component exempt from the single failure criterion. Some examples are clear: pipes are passive components and pumps and motor-operated valves that operate to perform their safety functions are active components. As discussed in Section 3.6 31 For example, SECY-77-439 (NRC 1977) states: "Examples [of passive failures in fluid systems] include the failure of a simple check valve to move to its correct position when required, the leakage of fluid from failed components, such as pipes and valves- particularly through a failed seal at a valve or pump-or line blockage. Motor-operated valves which have the source of power locked out are allowed to be treated as passive components."
32 For example, NUREG-1800 (NRC 2001c) states that "'[p]assive' structures and components, for the purpose of the license renewal rule, are those that perform an intended function ... without moving parts or without a change in configuration or properties ... 'passive' may also be interpreted to include structures and components that do not display 'a change of state.'"
32 For example, NUREG-1800 (NRC 2001c) states that "'[p]assive' structures and components, for the purpose of the license renewal rule, are those that perform an intended function ... without moving parts or without a change in configuration or properties ... 'passive' may also be interpreted to include structures and components that do not display 'a change of state.'"
33 IAEA 2009 34 NRC 2005a
33 IAEA 2009 34 NRC 2005a below, check valves might be classified as active or passive components depending on certain specific considerations.
 
below, check valves might be classified as active or passive components depending on certain specific considerations.
With respect to PSVs, the ASME BPV Code applicable to Byron and Braidwood includes requirements for overpressure protection that relate to the single failure criterion through several specific design and construction requirements. As a result, the PSVs are conservatively sized with sufficient margin to accommodate a single failure although the single failure criterion is almost never explicitly discussed or applied in accident analyses. The Byron and Braidwood UFSAR states that "adequate overpressurization protection is provided by the three installed safety valves." Neither the UFSAR system descriptions nor the safety analyses provide detailed discussions of potential PSV failures or their consequences. The principal discussion of potential PSV failures in the accident analyses occurs in the evaluation of an inadvertent opening of a PSV in UFSAR Section 15.6.1.
With respect to PSVs, the ASME BPV Code applicable to Byron and Braidwood includes requirements for overpressure protection that relate to the single failure criterion through several specific design and construction requirements. As a result, the PSVs are conservatively sized with sufficient margin to accommodate a single failure although the single failure criterion is almost never explicitly discussed or applied in accident analyses. The Byron and Braidwood UFSAR states that "adequate overpressurization protection is provided by the three installed safety valves." Neither the UFSAR system descriptions nor the safety analyses provide detailed discussions of potential PSV failures or their consequences. The principal discussion of potential PSV failures in the accident analyses occurs in the evaluation of an inadvertent opening of a PSV in UFSAR Section 15.6.1.
Most relevant for the current issue, the Byron and Braidwood UFSAR analyses of overpressure events (e.g., loss of load, loss of feedwater) do not apply the single failure criterion to cause a PSV to stick open (i.e., fail to reseat) when opening on steam flow. In addition, the UFSAR Feedwater System Pipe Break analysis (Chapter 15.2.8) does not apply the single failure criterion to cause a PSV to stick open either during steam discharge or during water discharge.
Most relevant for the current issue, the Byron and Braidwood UFSAR analyses of overpressure events (e.g., loss of load, loss of feedwater) do not apply the single failure criterion to cause a PSV to stick open (i.e., fail to reseat) when opening on steam flow. In addition, the UFSAR Feedwater System Pipe Break analysis (Chapter 15.2.8) does not apply the single failure criterion to cause a PSV to stick open either during steam discharge or during water discharge.
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OF THE APPEAL REVIEW PANEL FINDINGS For the reasons provided in Section 3, the Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Panel also concluded that, in preparing the Uprate SE and the Setpoint SE, the NRC staff exercised reasonable and well-informed engineering judgment when the NRC staff concluded that the PSVs were unlikely to stick open.
OF THE APPEAL REVIEW PANEL FINDINGS For the reasons provided in Section 3, the Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Panel also concluded that, in preparing the Uprate SE and the Setpoint SE, the NRC staff exercised reasonable and well-informed engineering judgment when the NRC staff concluded that the PSVs were unlikely to stick open.
The non-escalation position does not establish specific standards for valve qualification, so the non-escalation position, standing alone, provides no basis for rejecting the licensee's reliance on EPRI valve testing . Moreover, the Panel found that no mistake or error occurred in the licensee's or previous staff's reliance on the EPRI testing program that included an evaluation of 35 Examples include Watts Bar (NRC 1982 and TVA 1983), North Anna (NRC 1976), and AP1000 (Westinghouse 2011 ).
The non-escalation position does not establish specific standards for valve qualification, so the non-escalation position, standing alone, provides no basis for rejecting the licensee's reliance on EPRI valve testing . Moreover, the Panel found that no mistake or error occurred in the licensee's or previous staff's reliance on the EPRI testing program that included an evaluation of 35 Examples include Watts Bar (NRC 1982 and TVA 1983), North Anna (NRC 1976), and AP1000 (Westinghouse 2011 ).
water discharge through pressurizer valves. 36 Therefore, the Panel also concluded that the NRC staff's position on valve qualification in the Backfit SE is a new or modified interpretation of what constitutes compliance.
water discharge through pressurizer valves. 36 Therefore, the Panel also concluded that the NRC staff's position on valve qualification in the Backfit SE is a new or modified interpretation of what constitutes compliance.
The Panel also concluded that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. The Panel included additional information about this finding in Section 6 and Appendices B and C below.
The Panel also concluded that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. The Panel included additional information about this finding in Section 6 and Appendices B and C below.
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The Panel's evaluation of the treatment of PSV failure potential includes an assessment of multiple relevant references, which are discussed chronologically in the sections that follow.
The Panel's evaluation of the treatment of PSV failure potential includes an assessment of multiple relevant references, which are discussed chronologically in the sections that follow.
36 "Pressurizer valves" is used in this report to refer to either PORVs or PSVs when discussing issues common to both types of valves.
36 "Pressurizer valves" is used in this report to refer to either PORVs or PSVs when discussing issues common to both types of valves.
3.1    General Design Criteria (1971)
3.1    General Design Criteria (1971)
In 1971, the Atomic Energy Commission published the GDCs, which had been under development since 1965.37 The introduction to 10 CFR Part 50, Appendix A addresses "Single Failure" in the section on Definitions and Explanations. The paragraph on single failures includes a footnote stating: "The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development" (emphasis added}.
In 1971, the Atomic Energy Commission published the GDCs, which had been under development since 1965.37 The introduction to 10 CFR Part 50, Appendix A addresses "Single Failure" in the section on Definitions and Explanations. The paragraph on single failures includes a footnote stating: "The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development" (emphasis added}.
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SECY-77-439 also stresses the use of engineering j udgment relating to the probability of component failure and does not suggest that valve "certification" or "qualification" in accordance with ASME standards should be invoked as the basis for such decisions.
SECY-77-439 also stresses the use of engineering j udgment relating to the probability of component failure and does not suggest that valve "certification" or "qualification" in accordance with ASME standards should be invoked as the basis for such decisions.
3.3    TMI Action Plan Item 11.D.1 (1980)
3.3    TMI Action Plan Item 11.D.1 (1980)
As an element of the TMI Action Plan, the NRC staff required licensees to address the capability of relief and safety valves to perform their intended functions without failure. Specifically, 37 AEC 1971 38 NRC 1977
As an element of the TMI Action Plan, the NRC staff required licensees to address the capability of relief and safety valves to perform their intended functions without failure. Specifically, 37 AEC 1971 38 NRC 1977 Item 11.0.1 states that "[pJressurized-water reactor [PWR] and boiling-water reactor [BWR]
 
Item 11.0.1 states that "[pJressurized-water reactor [PWR] and boiling-water reactor [BWR]
licensees and applicants shall conduct testing to qualify the [RCS] relief and safety valves under expected operating conditions for design-basis transients and accidents." With reference to planned EPRI testing and other generic industry test programs, NUREG-0737 specified provisions for then-operating nuclear power plants and applicants for operating licenses and holders of construction perm its to address the TM I Action Plan items, including Item 11. D .1.
licensees and applicants shall conduct testing to qualify the [RCS] relief and safety valves under expected operating conditions for design-basis transients and accidents." With reference to planned EPRI testing and other generic industry test programs, NUREG-0737 specified provisions for then-operating nuclear power plants and applicants for operating licenses and holders of construction perm its to address the TM I Action Plan items, including Item 11. D .1.
NUREG-0737 stated, for the performance testing of relief and safety valves for Item 11.D.1 , that
NUREG-0737 stated, for the performance testing of relief and safety valves for Item 11.D.1 , that
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[GDCs] are met." As discussed in Appendix B to this report, the 1988 SE described the NRC staffs evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs. Based on the NRC staff and contractor review, the 1988 SE found that the performance of the PSVs and PORVs was acceptable based on the EPRI tests.
[GDCs] are met." As discussed in Appendix B to this report, the 1988 SE described the NRC staffs evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs. Based on the NRC staff and contractor review, the 1988 SE found that the performance of the PSVs and PORVs was acceptable based on the EPRI tests.
For the specific extended high pressure injection event, the 1988 SE states that water discharge through the PSVs and PORVs could be disregarded because of the long time available for operator action. However, the SE addressed water discharge through the PSVs and PORVs as part of the feedwater line break evaluation.
For the specific extended high pressure injection event, the 1988 SE states that water discharge through the PSVs and PORVs could be disregarded because of the long time available for operator action. However, the SE addressed water discharge through the PSVs and PORVs as part of the feedwater line break evaluation.
In the cover letter for the 1988 SE, the NRC staff states that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The 1988 SE contains no reference to or suggestion of a need for certification of these valves in accordance with the ASME BPV Code for water discharge capability. In 1990, the NRC staff also found the use of the EPRI test program similarly acceptable for Braidwood.4 1 39 WOG 1982 40 N RC 1988c, referred to as the 1988 SE 41 NRG 1990a
In the cover letter for the 1988 SE, the NRC staff states that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The 1988 SE contains no reference to or suggestion of a need for certification of these valves in accordance with the ASME BPV Code for water discharge capability. In 1990, the NRC staff also found the use of the EPRI test program similarly acceptable for Braidwood.4 1 39 WOG 1982 40 N RC 1988c, referred to as the 1988 SE 41 NRG 1990a 3.5 Westinghouse NSAL-93-013 and Supplement 1 (1993-1994)
 
3.5 Westinghouse NSAL-93-013 and Supplement 1 (1993-1994)
In 1993, Westinghouse sent NSAL-93-013 to operating nuclear power plants in response to its discovery that potentially non-conservative assumptions had been used in the licensing analysis of the IOECCS event. Westinghouse recommended that licensees determine if their pressurizer safety relief valves (PSRVs)42 "are capable of closing following discharge of subcooled water.tt Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse also noted that "licensees may have qualified these valves in compliance to NUREG-0737, Item 11.D. 1." If the PSRVs were not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees reevaluate the IOECCS event with three possible options of ( 1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the accident.
In 1993, Westinghouse sent NSAL-93-013 to operating nuclear power plants in response to its discovery that potentially non-conservative assumptions had been used in the licensing analysis of the IOECCS event. Westinghouse recommended that licensees determine if their pressurizer safety relief valves (PSRVs)42 "are capable of closing following discharge of subcooled water.tt Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse also noted that "licensees may have qualified these valves in compliance to NUREG-0737, Item 11.D. 1." If the PSRVs were not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees reevaluate the IOECCS event with three possible options of ( 1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the accident.
Later, in Supplement 1 to NSAL-93-013, Westinghouse alerted licensees to potential reduced time for operator action if a positive displacement pump (a typical component of the CVCS) were in service, and to the need to qualify the PSRVs and the piping downstream of the PSRVs and PORVs if water discharge from the pressurizer is predicted.
Later, in Supplement 1 to NSAL-93-013, Westinghouse alerted licensees to potential reduced time for operator action if a positive displacement pump (a typical component of the CVCS) were in service, and to the need to qualify the PSRVs and the piping downstream of the PSRVs and PORVs if water discharge from the pressurizer is predicted.
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3.6 Commission Paper on Passive Plant Designs (1994)
3.6 Commission Paper on Passive Plant Designs (1994)
In 1994, in preparation for the design certification reviews of passive reactor designs (e.g., the Westinghouse Advanced Passive 1000 (AP1000) and the General Electric Economic Simplified Boiling-Water Reactor (ESBWR)), the NRC staff presented nine issues to the Commission for policy decisions. 49 Although PSV categorization and performance requirements were not explicitly addressed, the paper does include an issue on "Definition of Passive Failure" and an 42 Westinghouse used the term PSRVs. The specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code, and by the licensee. This difference in terminology is not significant to any of the findings or conclusions in this report.
In 1994, in preparation for the design certification reviews of passive reactor designs (e.g., the Westinghouse Advanced Passive 1000 (AP1000) and the General Electric Economic Simplified Boiling-Water Reactor (ESBWR)), the NRC staff presented nine issues to the Commission for policy decisions. 49 Although PSV categorization and performance requirements were not explicitly addressed, the paper does include an issue on "Definition of Passive Failure" and an 42 Westinghouse used the term PSRVs. The specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code, and by the licensee. This difference in terminology is not significant to any of the findings or conclusions in this report.
43 NRC 1997 44 NRC 1998 45 NRC 2000 46 NRC 2004a 47 ComEd 1998 48 ComEd 1999 4 9 NRC 1994a
43 NRC 1997 44 NRC 1998 45 NRC 2000 46 NRC 2004a 47 ComEd 1998 48 ComEd 1999 4 9 NRC 1994a extensive discussion on whether check valves are passive or active components and how they should be addressed in current plants and future passive designs.
 
extensive discussion on whether check valves are passive or active components and how they should be addressed in current plants and future passive designs.
SEeY-94-084 recognized the GDes and SEeY-77-439 as establishing long-standing requirements and guidance in this area. The paper acknowledge~(CT13JISW14J that the industry (including EPRI documents and ANSI/ANS 58.950) have been inconsistent with respect to check valve failures, sometimes considering them as "active failures" and sometimes as "passive failures." In SEeY-77-439, however, the NRe staff stated that the failure of a simple check valve to move to its correct position when required was a "passive failure." In addition, SECY-94-084 states that "[i]n licensing reviews, however, only on a long-term basis [e.g., long-term recirculation cooling following a loss of coolant accident (LOCA)] does the NRe staff consider passive failures in fluid systems as potential accident initiators in addition to initiating events."
SEeY-94-084 recognized the GDes and SEeY-77-439 as establishing long-standing requirements and guidance in this area. The paper acknowledge~(CT13JISW14J that the industry (including EPRI documents and ANSI/ANS 58.950) have been inconsistent with respect to check valve failures, sometimes considering them as "active failures" and sometimes as "passive failures." In SEeY-77-439, however, the NRe staff stated that the failure of a simple check valve to move to its correct position when required was a "passive failure." In addition, SECY-94-084 states that "[i]n licensing reviews, however, only on a long-term basis [e.g., long-term recirculation cooling following a loss of coolant accident (LOCA)] does the NRe staff consider passive failures in fluid systems as potential accident initiators in addition to initiating events."
The paper also states that "[f]or current plants, the NRe staff normally treats check valves, except for those in containment isolation systems, as passive devices during transients or design-basis accidents."
The paper also states that "[f]or current plants, the NRe staff normally treats check valves, except for those in containment isolation systems, as passive devices during transients or design-basis accidents."
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The Panel considered the opening function of check valves and PSVs to be similar in that they both open through the motion of the valve disk under differential pressure with no external signal or motive power. The Panel also recognized that the ambiguity with respect to "passive" versus "active" component definitions and nomenclature exists for safety valves. In addition, the passive or active classification of check valves or safety valves may differ based on design considerations, inservice testing, or accident analyses. For example, the PSVs and PORVs, as well as numerous check valves, are classified as active components in the Byron and Braidwood inservice testing programs. However, for purposes of applying the single failure criterion in the GDe context, the Panel concluded that it is appropriate to consider the potential failure of a PSV following water discharge as a passive failure (consistent with the treatment of check valve failures for the operating fleet), provided the licensee or applicant qualifies the performance of the PSV in an acceptable mannertsM15J. !In tcn s1the case of Byron and Braidwood, the NRe staff accepted the EPRI testing associated with TMI Action Plan Item 11.D.1 to provide this qualification.
The Panel considered the opening function of check valves and PSVs to be similar in that they both open through the motion of the valve disk under differential pressure with no external signal or motive power. The Panel also recognized that the ambiguity with respect to "passive" versus "active" component definitions and nomenclature exists for safety valves. In addition, the passive or active classification of check valves or safety valves may differ based on design considerations, inservice testing, or accident analyses. For example, the PSVs and PORVs, as well as numerous check valves, are classified as active components in the Byron and Braidwood inservice testing programs. However, for purposes of applying the single failure criterion in the GDe context, the Panel concluded that it is appropriate to consider the potential failure of a PSV following water discharge as a passive failure (consistent with the treatment of check valve failures for the operating fleet), provided the licensee or applicant qualifies the performance of the PSV in an acceptable mannertsM15J. !In tcn s1the case of Byron and Braidwood, the NRe staff accepted the EPRI testing associated with TMI Action Plan Item 11.D.1 to provide this qualification.
3.7 Draft Standard Review Plan Revision (1996)
3.7 Draft Standard Review Plan Revision (1996)
The 1996 draft revision to SRP Sections 15.5.1 - 15.5.2 on IOEees and eves malfunctions includes extensive updates to the 1981 revision, but neither version includes any discussion, 50  ANS 1981 51  NRC 1994b
The 1996 draft revision to SRP Sections 15.5.1 - 15.5.2 on IOEees and eves malfunctions includes extensive updates to the 1981 revision, but neither version includes any discussion, 50  ANS 1981 51  NRC 1994b criteria, or guidance on applying ASME Code requirements to PSVs or on applying the single failure criterion or any other failure assumption to PSVs.52 3.8    Power Uprate Reviews and License Amendments (2001-2006)
 
criteria, or guidance on applying ASME Code requirements to PSVs or on applying the single failure criterion or any other failure assumption to PSVs.52 3.8    Power Uprate Reviews and License Amendments (2001-2006)
As part of the 2001 power uprate review for Byron and Braidwood, the NRC staff approved the analysis of an IOECCS (UFSAR Section 15.5.1) that included pressurizer filling, PSV water discharge, ECCS termination, and PSV closure. In the Backfit SE, the NRC staff indicated that the 2001 license amendment was predicated on the NRC's mistaken (unsubstantiated) belief that the valves were ASME-qualified (certified). However, the Panel's review of the SE and associated RAls showed that, in 2001, the NRC staff was well aware of the nature of the EPRI testing that the licensee relied on. The Panel did not find any evidence that the licensee claimed or the NRC staff believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows*that the NRC staff thoroughly considered the testing conducted on valves of the type installed at the plants and applied well-informed and reasoned technical judgment in reaching its conclusion that the EPRI testing provided appropriate qualification.
As part of the 2001 power uprate review for Byron and Braidwood, the NRC staff approved the analysis of an IOECCS (UFSAR Section 15.5.1) that included pressurizer filling, PSV water discharge, ECCS termination, and PSV closure. In the Backfit SE, the NRC staff indicated that the 2001 license amendment was predicated on the NRC's mistaken (unsubstantiated) belief that the valves were ASME-qualified (certified). However, the Panel's review of the SE and associated RAls showed that, in 2001, the NRC staff was well aware of the nature of the EPRI testing that the licensee relied on. The Panel did not find any evidence that the licensee claimed or the NRC staff believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows*that the NRC staff thoroughly considered the testing conducted on valves of the type installed at the plants and applied well-informed and reasoned technical judgment in reaching its conclusion that the EPRI testing provided appropriate qualification.
The Panel confirmed its conclusions and understanding about the 2001 NRC staff review via discussions with the individual who was the responsible Section Chief in the Reactor Systems Branch at the time. He informed the Panel that the 2001 license amendment was based on the exercise of staff engineering judgment and that there was no discussion of ASME BPV Code certification or qualification of valves. In addition, the Panel found that the NRC approved power uprates for other nuclear power plants that included comparable staff evaluations of water discharge through PORVs or PSVs based on test information provided by individual licensees.
The Panel confirmed its conclusions and understanding about the 2001 NRC staff review via discussions with the individual who was the responsible Section Chief in the Reactor Systems Branch at the time. He informed the Panel that the 2001 license amendment was based on the exercise of staff engineering judgment and that there was no discussion of ASME BPV Code certification or qualification of valves. In addition, the Panel found that the NRC approved power uprates for other nuclear power plants that included comparable staff evaluations of water discharge through PORVs or PSVs based on test information provided by individual licensees.
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In the sample of SEs it reviewed, the Panel did not find a specific requirement for the PORVs or PSVs to be certified under the ASME BPV Code as capable of reclosing ~fter rcn71water discharge.
In the sample of SEs it reviewed, the Panel did not find a specific requirement for the PORVs or PSVs to be certified under the ASME BPV Code as capable of reclosing ~fter rcn71water discharge.
In 2004, the NRC issued license amendments for Byron and Braidwood granting an adjustment to the PSV setpoints. In an RAI, the NRC staff requested that the licensee perform a quantitative analysis regarding the number of opening cycles during which the PSV would be expected to pass water and the temperature of the water being discharged. In the Setpoint SE, the NRC staff concluded that the analysis was acceptable for assuring that the PSVs would remain operable following a spurious safety injection event.
In 2004, the NRC issued license amendments for Byron and Braidwood granting an adjustment to the PSV setpoints. In an RAI, the NRC staff requested that the licensee perform a quantitative analysis regarding the number of opening cycles during which the PSV would be expected to pass water and the temperature of the water being discharged. In the Setpoint SE, the NRC staff concluded that the analysis was acceptable for assuring that the PSVs would remain operable following a spurious safety injection event.
52 NRC 1996 53 NRC 2001d
52 NRC 1996 53 NRC 2001d 3.9    RIS 2005-29 (2005), and Proposed Draft Revision 1 to RIS 2005-29 (2015)
 
3.9    RIS 2005-29 (2005), and Proposed Draft Revision 1 to RIS 2005-29 (2015)
In 2005, the NRC staff issued RIS 2005-29 "to notify licensees of a concern identified during recent reviews of power uprate [LARs]." The RIS addressed the manner in which some licensees acted in response to NSAL-93-013. The RIS was issued at the division level in NRR and does not include a record of office-level concurrence. The RIS was not reviewed by CRGR.
In 2005, the NRC staff issued RIS 2005-29 "to notify licensees of a concern identified during recent reviews of power uprate [LARs]." The RIS addressed the manner in which some licensees acted in response to NSAL-93-013. The RIS was issued at the division level in NRR and does not include a record of office-level concurrence. The RIS was not reviewed by CRGR.
The Panel requested information on the basis for the CRGR's decision not to review the proposed RIS before it was issued, but the CRGR staff could not find any related documentation. It appears to the Panel that the CRGR may not have reviewed the RIS because of assertions in the RIS such as these:
The Panel requested information on the basis for the CRGR's decision not to review the proposed RIS before it was issued, but the CRGR staff could not find any related documentation. It appears to the Panel that the CRGR may not have reviewed the RIS because of assertions in the RIS such as these:
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The Panel also notes that neither RIS 2005-29 nor its draft Revision 1,55 which is currently under development, discuss water discharge certification requirements in accordance with the ASME BPV Code. In fact, as stated above, the NRC issued a 2006 power uprate amendment for Beaver Valley in which the SE cited RIS 2005-29 and yet relied on the EPRI testing data to address the concern.
The Panel also notes that neither RIS 2005-29 nor its draft Revision 1,55 which is currently under development, discuss water discharge certification requirements in accordance with the ASME BPV Code. In fact, as stated above, the NRC issued a 2006 power uprate amendment for Beaver Valley in which the SE cited RIS 2005-29 and yet relied on the EPRI testing data to address the concern.
3.10 SECY-05-0138 (2005)
3.10 SECY-05-0138 (2005)
SECY-05-0138 presents a comprehensive history of the application of the single failure criterion, including extensive discussion of the treatment of passive components in fluid 54 NRC 2003 55 NRC 2015a
SECY-05-0138 presents a comprehensive history of the application of the single failure criterion, including extensive discussion of the treatment of passive components in fluid 54 NRC 2003 55 NRC 2015a systems.56 The paper enclosed a July 2005 draft of an NRC staff technical report on the single failure criterion. Section 4.2.2 of this report acknowledges that "[o]ne particular issue identified in this project is the continued existence of the footnote to the definition of single failure in 10CFR
 
systems.56 The paper enclosed a July 2005 draft of an NRC staff technical report on the single failure criterion. Section 4.2.2 of this report acknowledges that "[o]ne particular issue identified in this project is the continued existence of the footnote to the definition of single failure in 1OCFR
[Part] 50 Appendix A stating that the regulatory position on considering passive failures in fluid systems is under development." In Section 2.5.3, the draft report quotes from SECY-77-439 (discussed above) and recognizes that in current practice, as in 1977, "[p]assive failures in fluid systems are generally excluded from single-failure assessments."
[Part] 50 Appendix A stating that the regulatory position on considering passive failures in fluid systems is under development." In Section 2.5.3, the draft report quotes from SECY-77-439 (discussed above) and recognizes that in current practice, as in 1977, "[p]assive failures in fluid systems are generally excluded from single-failure assessments."
SECY-05-0138 and the accompanying draft report present three alternatives for using a risk-informed and performance-based approach to address the single failure issue. The draft report clarifies that all of the alternatives "could include developing a position on single passive failures in fluid systems to replace the footnote now in 10 CFR Part 50 Appendix A definitions."
SECY-05-0138 and the accompanying draft report present three alternatives for using a risk-informed and performance-based approach to address the single failure issue. The draft report clarifies that all of the alternatives "could include developing a position on single passive failures in fluid systems to replace the footnote now in 10 CFR Part 50 Appendix A definitions."
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         ... the ASME ... orig1inal Overpressure Protection Report ... inservice test history ...
         ... the ASME ... orig1inal Overpressure Protection Report ... inservice test history ...
including both water and steam tests" (emphasis added)
including both water and steam tests" (emphasis added)
The Backfit SE contends that an IOECCS would escalate to a more severe event. Such an escalation would be contrary to the Byron and Braidwood licensing basis (i.e., contrary to the ANS non-escalation position) and could be in non-compliance with the GDCs (as included in the 56 NRC 2005a
The Backfit SE contends that an IOECCS would escalate to a more severe event. Such an escalation would be contrary to the Byron and Braidwood licensing basis (i.e., contrary to the ANS non-escalation position) and could be in non-compliance with the GDCs (as included in the 56 NRC 2005a Byron and Braidwood licensing basis) since an IOECCS with a stuck-open valve had not been analyzed and shown to meet the appropriate criteria for an AOO.
 
Byron and Braidwood licensing basis) since an IOECCS with a stuck-open valve had not been analyzed and shown to meet the appropriate criteria for an AOO.
Based on its review of all the relevant documents and discussions with the individuals (staff and managers) involved in the original review and the backfit, the Panel has developed an understanding of the regulatory requirements and practices, the potential safety issues, and backfit rule obligations. The Panel has determined that the numerous, complex, and detailed regulatory and technical issues all depend on the answers to two critical questions on valve performance:
Based on its review of all the relevant documents and discussions with the individuals (staff and managers) involved in the original review and the backfit, the Panel has developed an understanding of the regulatory requirements and practices, the potential safety issues, and backfit rule obligations. The Panel has determined that the numerous, complex, and detailed regulatory and technical issues all depend on the answers to two critical questions on valve performance:
* Must the PSVs in question be assumed to fail given liquid water discharge because of the lack of ASME BPV Code certification for water discharge?
* Must the PSVs in question be assumed to fail given liquid water discharge because of the lack of ASME BPV Code certification for water discharge?
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In interactions with the Panel, NRR staff emphasized several issues raised in the Backfit Letter.
In interactions with the Panel, NRR staff emphasized several issues raised in the Backfit Letter.
The Panel summarizes its consideration of those issues in the following subsections.
The Panel summarizes its consideration of those issues in the following subsections.
3.12.1 Non-Escalation Position and Valve Failure In the Backfit SE, the NRC staff discussed the definition of event conditions in ANS-51 .1/N18.2-1973 and the provision In this standard that events of one condition do not propagate to cause a more serious fault. This position is commonly known as the non-escalation position. In
3.12.1 Non-Escalation Position and Valve Failure In the Backfit SE, the NRC staff discussed the definition of event conditions in ANS-51 .1/N18.2-1973 and the provision In this standard that events of one condition do not propagate to cause a more serious fault. This position is commonly known as the non-escalation position. In
~nteractions JCT18JlMAs19isw201with the Panel, NRR staff provided several clarifications on this topic, summarized by the Panel as follows:
~nteractions JCT18JlMAs19isw201with the Panel, NRR staff provided several clarifications on this topic, summarized by the Panel as follows:
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3.12.2 Non-Escalation Position and Return to Service In the Backfit SE, the NRG staff makes reference to the time it would take to clean up a contaminated containment fiollowing a stuck-open pressurizer valve. In interactions with the Panel, NRR staff re-emphasized concerns that extended steam and water discharge through the pressurizer valves would result in the failure of the pressurizer relief tank rupture disk, would require repair of the damaged PSVs, and might cause an extended time period for the return to service of the nuclear power plant.
3.12.2 Non-Escalation Position and Return to Service In the Backfit SE, the NRG staff makes reference to the time it would take to clean up a contaminated containment fiollowing a stuck-open pressurizer valve. In interactions with the Panel, NRR staff re-emphasized concerns that extended steam and water discharge through the pressurizer valves would result in the failure of the pressurizer relief tank rupture disk, would require repair of the damaged PSVs, and might cause an extended time period for the return to service of the nuclear power plant.
The Panel does not consider the time period necessary for the licensee to perform radioactive clean-up activities in the containment building, to inspect and conduct any necessary repairs to the PSVs, or to prepare for plant startup, to constitute issues that support a compliance backfit imposed by the NRG. The NRG staff would verify (e.g., through inspection) that the licensee had conducted these activities appropriately to protect the public health and safety prior to plant restart. The Backfit SE states that UFSAR Section 15.5.1.3 "implie[s]" that the plant will return to operation in a "short period," but the Panel found no eases basis for a timing requirement in UFSAR Section 15.5.1 .3. Also, the Panel did not find a regulatory requirement or basis for defining or limiting the time available for the plant to return to operation.
The Panel does not consider the time period necessary for the licensee to perform radioactive clean-up activities in the containment building, to inspect and conduct any necessary repairs to the PSVs, or to prepare for plant startup, to constitute issues that support a compliance backfit imposed by the NRG. The NRG staff would verify (e.g., through inspection) that the licensee had conducted these activities appropriately to protect the public health and safety prior to plant restart. The Backfit SE states that UFSAR Section 15.5.1.3 "implie[s]" that the plant will return to operation in a "short period," but the Panel found no eases basis for a timing requirement in UFSAR Section 15.5.1 .3. Also, the Panel did not find a regulatory requirement or basis for defining or limiting the time available for the plant to return to operation.
3.12.3 TMI Action Plan Item 11.0.1 and EPRI Testing Although the Backfit Letter and NRR Appeal Decision do not speak explicitly to TMI Action Plan Item 11.0.1, in interactions with the Panel, NRR staff stated that the known and established standard in question is the TMI Action Plan Item 11.D.I standard for licensees and applicants to conduct testing to qualify the RCS relief and safety valves under expected operating conditions for design-basis transients and accidents. As discussed above and in Appendix B to this report,
3.12.3 TMI Action Plan Item 11.0.1 and EPRI Testing Although the Backfit Letter and NRR Appeal Decision do not speak explicitly to TMI Action Plan Item 11.0.1, in interactions with the Panel, NRR staff stated that the known and established standard in question is the TMI Action Plan Item 11.D.I standard for licensees and applicants to conduct testing to qualify the RCS relief and safety valves under expected operating conditions for design-basis transients and accidents. As discussed above and in Appendix B to this report, the NRC accepted the EPRI testing to satisfy TMI Action Plan Item 11.D. 1 for Byron and Braidwood in SEs forwarded by letters in 1988 and 1990. Therefore, the Panel concludes that this known and established standard referenced by the NRC staff had been met for Byron and Braidwood.
 
the NRC accepted the EPRI testing to satisfy TMI Action Plan Item 11.D. 1 for Byron and Braidwood in SEs forwarded by letters in 1988 and 1990. Therefore, the Panel concludes that this known and established standard referenced by the NRC staff had been met for Byron and Braidwood.
In interactions with the Panel, the NRR staff further stated that an omission or mistake of fact occurred when the licensee failed to acknowledge that the EPRI testing program did not evaluate water discharge from the pressurizer valves during extended high pressure safety injection for Byron and Braidwood. As discussed in Appendix B to this report, in the 1988 and 1990 SEs for the Byron and Braidwood responses to TMI Action Plan Item 11.D.1, the NRC staff evaluated the capability of the PSVs and PORVs during feedwater line break accidents, including water discharge. In these SEs, the NRC staff found that the performance of the PSVs and PORVs with water discharge was acceptable based on the EPRI tests. Therefore, the Panel also concluded that the licensee's reference to the EPRI testing program was not an omission or a mistake of fact.
In interactions with the Panel, the NRR staff further stated that an omission or mistake of fact occurred when the licensee failed to acknowledge that the EPRI testing program did not evaluate water discharge from the pressurizer valves during extended high pressure safety injection for Byron and Braidwood. As discussed in Appendix B to this report, in the 1988 and 1990 SEs for the Byron and Braidwood responses to TMI Action Plan Item 11.D.1, the NRC staff evaluated the capability of the PSVs and PORVs during feedwater line break accidents, including water discharge. In these SEs, the NRC staff found that the performance of the PSVs and PORVs with water discharge was acceptable based on the EPRI tests. Therefore, the Panel also concluded that the licensee's reference to the EPRI testing program was not an omission or a mistake of fact.
3.12.4 ASME Code Certification In the Backfit SE, the NRC staff stated that certain ASME Code information would be necessary to support water qualification of the PSVs. In interactions with the Panel, NRR staff stated that, to satisfy the standard for water discharge capability of pressurizer valves, it would be necessary to conduct flow capacity certification in accordance with the ASME BPV Code and inservice testing throughout the service life in accordance with the ASME OM Code. The NRR staff referenced certain licensing actions in which water discharge was not considered acceptable, or different actions were required. 57 As discussed in Appendix C to this report, the NRC staff required additional actions for some licensees to support reliance on the PORVs for water discharge and to avoid water discharge through the PSVs. The Panel found, however, that the NRC staff also allowed some licensees to rely only on EPRI testing without significant additional activities. The Panel did not identify instances where the NRC staff imposed certification by the ASME BPV Code and testing in accordance with the OM Code, or required alternatives to the ASME BPV or OM Codes, in the examples of NRC staff review of water discharge capability for pressurizer valves.
3.12.4 ASME Code Certification In the Backfit SE, the NRC staff stated that certain ASME Code information would be necessary to support water qualification of the PSVs. In interactions with the Panel, NRR staff stated that, to satisfy the standard for water discharge capability of pressurizer valves, it would be necessary to conduct flow capacity certification in accordance with the ASME BPV Code and inservice testing throughout the service life in accordance with the ASME OM Code. The NRR staff referenced certain licensing actions in which water discharge was not considered acceptable, or different actions were required. 57 As discussed in Appendix C to this report, the NRC staff required additional actions for some licensees to support reliance on the PORVs for water discharge and to avoid water discharge through the PSVs. The Panel found, however, that the NRC staff also allowed some licensees to rely only on EPRI testing without significant additional activities. The Panel did not identify instances where the NRC staff imposed certification by the ASME BPV Code and testing in accordance with the OM Code, or required alternatives to the ASME BPV or OM Codes, in the examples of NRC staff review of water discharge capability for pressurizer valves.
The NRR staff also identified for the Panel specific ASME Code provisions that it viewed as supporting its position that ASME Code requirements apply to qualification of pressurizer valves for water discharge. The NRR staff, however, did not provide evidence that the NRC staff has consistently interpreted these provisions as the NRC staff is now interpreting them. Given th~ I rcr21Jrsw221NRC staff's resolution of TMI Action Plan Item 11.D.1 and the 1Jariations in
The NRR staff also identified for the Panel specific ASME Code provisions that it viewed as supporting its position that ASME Code requirements apply to qualification of pressurizer valves for water discharge. The NRR staff, however, did not provide evidence that the NRC staff has consistently interpreted these provisions as the NRC staff is now interpreting them. Given th~ I rcr21Jrsw221NRC staff's resolution of TMI Action Plan Item 11.D.1 and the 1Jariations in
~MAS231JSW24Jthe NRC staffs licensing practices, the Panel concludes that the NRR staff's current application of the ASME Code is not supported by the his torical record.
~MAS231JSW24Jthe NRC staffs licensing practices, the Panel concludes that the NRR staff's current application of the ASME Code is not supported by the his torical record.
3.12.5 Conduct of 2001 and 2004 License Amendment Reviews In light of the wide range of positions taken by the NRC staff during its reviews of pressurizer valve capability since the TMl-2 accident, the Panel agrees that, in the course of preparing the 2001 Uprate SE or Setpoint SE, the NRC staff could have considered the need for the licensee for Byron and Braidwood to improve the reliability of the PSVs or PORVs for water discharge or to avoid water discharge through the PSVs by PORV improvements. The NRG staff may have s7 Salem (NRC 1997), Millstone (NRC 1998), and Callaway (NRC 2000)
3.12.5 Conduct of 2001 and 2004 License Amendment Reviews In light of the wide range of positions taken by the NRC staff during its reviews of pressurizer valve capability since the TMl-2 accident, the Panel agrees that, in the course of preparing the 2001 Uprate SE or Setpoint SE, the NRC staff could have considered the need for the licensee for Byron and Braidwood to improve the reliability of the PSVs or PORVs for water discharge or to avoid water discharge through the PSVs by PORV improvements. The NRG staff may have s7 Salem (NRC 1997), Millstone (NRC 1998), and Callaway (NRC 2000) been able to justify additional actions, but they determined that it was not necessary. Instead, the NRC staff reviewers in 2001 used their expert engineering judgement to determine that it was not necessary to assume that the PSVs or PORVs would stick open with water discharge, based on EPRI test informa1ion. licensee supplemental information, and their own technical experience.
 
been able to justify additional actions, but they determined that it was not necessary. Instead, the NRC staff reviewers in 2001 used their expert engineering judgement to determine that it was not necessary to assume that the PSVs or PORVs would stick open with water discharge, based on EPRI test informa1ion. licensee supplemental information, and their own technical experience.
In discussions with the Panel, NRR staff raised a concern that the Setpoint SE does not document a re-review of the qualification of the PSVs and noted that if the Uprate SE had not found water discharge through the PSVs to be acceptable, it is unlikely that the NRC staff would 1
In discussions with the Panel, NRR staff raised a concern that the Setpoint SE does not document a re-review of the qualification of the PSVs and noted that if the Uprate SE had not found water discharge through the PSVs to be acceptable, it is unlikely that the NRC staff would 1
have approved this 2004 amendment. In Appendix C to this report, the Panel summarizes the discussion in the Setpoint SE of the PSV water discharge capability. The Panel recognizes that a staff review may rely on a previous more extensive review to determine the acceptability of a similar request. The Panel does not consider the review approach used in 2004 to challenge the acceptability of the 2001 review.
have approved this 2004 amendment. In Appendix C to this report, the Panel summarizes the discussion in the Setpoint SE of the PSV water discharge capability. The Panel recognizes that a staff review may rely on a previous more extensive review to determine the acceptability of a similar request. The Panel does not consider the review approach used in 2004 to challenge the acceptability of the 2001 review.
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* In the NRR Appeal Decision, the NRC staff claims that "[t]he NRC erred in approving a sequence of events that allowed the [IOECCS], [CVCS] malfunction, and inadvertent opening of a pressurizer safety or relief valve analyses in the 2001 and 2004 [SEs]" and "the NRC staff understood the PSVs to be qualified for water relief when. in fact, they were not."
* In the NRR Appeal Decision, the NRC staff claims that "[t]he NRC erred in approving a sequence of events that allowed the [IOECCS], [CVCS] malfunction, and inadvertent opening of a pressurizer safety or relief valve analyses in the 2001 and 2004 [SEs]" and "the NRC staff understood the PSVs to be qualified for water relief when. in fact, they were not."
* Exelon claims in the NRR Backfit Appeal that "the compliance exception requires more than simply asserting that the prior staff approvals were wrong- the NRC must demonstrate that the prior approvals were erroneous because of an omission or mistake of fact at the time of the approval. The NRC has not made that case here."
* Exelon claims in the NRR Backfit Appeal that "the compliance exception requires more than simply asserting that the prior staff approvals were wrong- the NRC must demonstrate that the prior approvals were erroneous because of an omission or mistake of fact at the time of the approval. The NRC has not made that case here."
On the basis of its independent review, the Panel concluded that, in 2001 and 2004, the NRC staff did not misunderstand 1he qualification status of the PSVs and that it was not a mistake to undertake a review of or make a technically based safety finding on the likely successful performance of the valves. In the Panel's opinion, the actions of the Reactor Systems Branch in 2001 to reach out to the Division of Engineering's Mechanical Engineering Branch for expert technical review assistance was both appropriate and commendable. ~ fter      I 1cr2sisM2sisw27Jconsidering the materials presented by the licensee in support of the 2001 and 2004 requests and discussing the 2001 review with one of the involved managers, the Panel found no indication that the senior reviewer evaluating the topic was misled regarding the qualification status of the PSVs, but rather used his expert judgment in determining the appropriate level of qualification for a technically complex topic for which there was not a single accepted approach. For these reasons, the Panel concluded that the NRC staff reviews and
On the basis of its independent review, the Panel concluded that, in 2001 and 2004, the NRC staff did not misunderstand 1he qualification status of the PSVs and that it was not a mistake to undertake a review of or make a technically based safety finding on the likely successful performance of the valves. In the Panel's opinion, the actions of the Reactor Systems Branch in 2001 to reach out to the Division of Engineering's Mechanical Engineering Branch for expert technical review assistance was both appropriate and commendable. ~ fter      I 1cr2sisM2sisw27Jconsidering the materials presented by the licensee in support of the 2001 and 2004 requests and discussing the 2001 review with one of the involved managers, the Panel found no indication that the senior reviewer evaluating the topic was misled regarding the qualification status of the PSVs, but rather used his expert judgment in determining the appropriate level of qualification for a technically complex topic for which there was not a single accepted approach. For these reasons, the Panel concluded that the NRC staff reviews and approvals of the 2001 and 2004 license amendments were not based on omissions or mistakes of fact.
 
approvals of the 2001 and 2004 license amendments were not based on omissions or mistakes of fact.
4.2    What is the known and established standard for water qualification of PSVs?
4.2    What is the known and established standard for water qualification of PSVs?
The Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Commission has not established a more detailed or prescriptive standard.
The Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Commission has not established a more detailed or prescriptive standard.
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4.5    Given that Exelon suggests that the NRC pursue a cost-justified substantial safety enhancement backflt, what is the contribution to overall plant risk of the current configuration at Braidwood and Byron?
4.5    Given that Exelon suggests that the NRC pursue a cost-justified substantial safety enhancement backflt, what is the contribution to overall plant risk of the current configuration at Braidwood and Byron?
The Panel requested RES to provide information and insights on the risk significance of the sequence at issue, to assure that the Panel's judgments were being made with a full understanding of their signiffcance, and to assist in responding to the EDO question.
The Panel requested RES to provide information and insights on the risk significance of the sequence at issue, to assure that the Panel's judgments were being made with a full understanding of their signiffcance, and to assist in responding to the EDO question.
The RES studfBtcn11 suggests that the most significant IOECCS sequence, assuming that all pressurizer overfill events lead to a small LOCA, contributes approximately 1 percent of the total internal event core damage frequency (CDF). In its report, RES estimated that the maximum benefit (CDF reduction) of 1.5E-07 per year would be achieved if the. plants were modified (jbackfitj is perfectl y effective such that pressurizer overfi'lling wasi..&sect; always prevented.,i If the PSVs are not assumed to always fail following water discharge (consistent with the NRC staff expert judgment in 2001) or if the plants were modified in a different way that did not prevent pressurizer o*.ierfillin~backfit is less than perfectly effective. JMAS32rthe risk-reduction benefit of implementing the backfit would be even smaller.
The RES studfBtcn11 suggests that the most significant IOECCS sequence, assuming that all pressurizer overfill events lead to a small LOCA, contributes approximately 1 percent of the total internal event core damage frequency (CDF). In its report, RES estimated that the maximum benefit (CDF reduction) of 1.5E-07 per year would be achieved if the. plants were modified (jbackfitj is perfectl y effective such that pressurizer overfi'lling wasi..&sect; always prevented.,i If the PSVs are not assumed to always fail following water discharge (consistent with the NRC staff expert judgment in 2001) or if the plants were modified in a different way that did not prevent pressurizer o*.ierfillin~backfit is less than perfectly effective. JMAS32rthe risk-reduction benefit of implementing the backfit would be even smaller.
The Panel is aware of and sensitive to two important issues related to this question. First, NRR, not the Panel, is responsible for any decisions on alternative application of the backfit rule to this issue (through the other categories of adequate protection or cost-justified substantial safety enhancement). Second, the Panel does not wish to imply that ''the contribution to plant risk" should be seen as the only measure of enhanced safety. The issues of event classification and the non-escalation of events are essentially defense-in-depth concepts. Defense in depth has a recognized role and value in the regulatory process. The Panel is also aware that not every defense-in-depth feature has the same safety significance, and that the estimated risk significance (measured in core damage frequency) is very relevant.
The Panel is aware of and sensitive to two important issues related to this question. First, NRR, not the Panel, is responsible for any decisions on alternative application of the backfit rule to this issue (through the other categories of adequate protection or cost-justified substantial safety enhancement). Second, the Panel does not wish to imply that ''the contribution to plant risk" should be seen as the only measure of enhanced safety. The issues of event classification and the non-escalation of events are essentially defense-in-depth concepts. Defense in depth has a recognized role and value in the regulatory process. The Panel is also aware that not every defense-in-depth feature has the same safety significance, and that the estimated risk significance (measured in core damage frequency) is very relevant.
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In the first of these determinations, the NRC staff's compliance backfit is based on the assumption in the Backfit SE that the PSV fails to reclose given the absence of "ASME water 58 NRC 2016f 59 The RES stud y explains that "any practical backfit remedy is not expected to be completely effective.
In the first of these determinations, the NRC staff's compliance backfit is based on the assumption in the Backfit SE that the PSV fails to reclose given the absence of "ASME water 58 NRC 2016f 59 The RES stud y explains that "any practical backfit remedy is not expected to be completely effective.
Therefore. this delta CDF represents the maximum oossible benefit from any backfit plant change."
Therefore. this delta CDF represents the maximum oossible benefit from any backfit plant change."
qualification documentation." As indicated in the Backfit SE, the Uprate SE involved a technical evaluation of safety valve capability and likely performance under water-discharge conditions rather than a simple assumption of a failure. The NRR Appeal Decision indicates that "the 2001 and 2004 [license amendment] approvals occurred because the NRC staff understood the PSVs to be qualified for water relief when, in fact. they were not."
qualification documentation." As indicated in the Backfit SE, the Uprate SE involved a technical evaluation of safety valve capability and likely performance under water-discharge conditions rather than a simple assumption of a failure. The NRR Appeal Decision indicates that "the 2001 and 2004 [license amendment] approvals occurred because the NRC staff understood the PSVs to be qualified for water relief when, in fact. they were not."
The Panel carefully considered these views and has reviewed the relevant documents including the licensee's responses to the NRC staffs RAls,60 the NRR technical branch's SE input,61 and the Uprate SE. The Panel did not find any evidence that the licensee had claimed or the NRC staff had believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows thorough consideration of the testing conducted on valves of the type installed at the plant and a well-informed technical judgment that this testing provided appropriate qualification.
The Panel carefully considered these views and has reviewed the relevant documents including the licensee's responses to the NRC staffs RAls,60 the NRR technical branch's SE input,61 and the Uprate SE. The Panel did not find any evidence that the licensee had claimed or the NRC staff had believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows thorough consideration of the testing conducted on valves of the type installed at the plant and a well-informed technical judgment that this testing provided appropriate qualification.
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Operator training, control room procedures to terminate the event before pressurizer filling, and use of PORVs rather than reliance on PSVs, are clearly preferred and prudent measures, whether they form the facilities' UFSAR licensing basis and are assumed in the accident analyses or not.
Operator training, control room procedures to terminate the event before pressurizer filling, and use of PORVs rather than reliance on PSVs, are clearly preferred and prudent measures, whether they form the facilities' UFSAR licensing basis and are assumed in the accident analyses or not.
The PSVs in question were designed for steam service. Steam relief is their normal service condition and applies to their ASME BPV Code certification. The Panel supports the previous NRC staff determinations for Byron and Braidwood and certain other plants that PSVs experiencing water discharge during an abnormal or accident condition need not be assumed to fail since there was a reasonable and technically well-informed engineering judgement to the contrary. However, the Panel also considers the actions by various licensees to improve the reliability and performance of the PORVs to avoid water discharge through the PSVs to be prudent in light of the design specifications of the PSVs.
The PSVs in question were designed for steam service. Steam relief is their normal service condition and applies to their ASME BPV Code certification. The Panel supports the previous NRC staff determinations for Byron and Braidwood and certain other plants that PSVs experiencing water discharge during an abnormal or accident condition need not be assumed to fail since there was a reasonable and technically well-informed engineering judgement to the contrary. However, the Panel also considers the actions by various licensees to improve the reliability and performance of the PORVs to avoid water discharge through the PSVs to be prudent in light of the design specifications of the PSVs.
The Panel considered but could not determine the extent to which the licensee for Byron and Braidwood addressed crediting water discharge through the PSVs, PORVs, or PORV block valves in the Byron and Braidwood inservice testing programs. The Panel recognizes that the difference between the intended use of these valves for overpressure protection and their infrequent use in response to certain plant events might be considered in implementing appropriate inservice testing activities.
The Panel considered but could not determine the extent to which the licensee for Byron and Braidwood addressed crediting water discharge through the PSVs, PORVs, or PORV block valves in the Byron and Braidwood inservice testing programs. The Panel recognizes that the difference between the intended use of these valves for overpressure protection and their infrequent use in response to certain plant events might be considered in implementing appropriate inservice testing activities.
The Panel notes that water discharge through various pressurizer valves is not a new issue because water discharge has always been credited (by the licensee for Byron and Braidwood and other licensees) for the feedwater line break analysis in UFSAR Section 15.2.8.
The Panel notes that water discharge through various pressurizer valves is not a new issue because water discharge has always been credited (by the licensee for Byron and Braidwood and other licensees) for the feedwater line break analysis in UFSAR Section 15.2.8.
On the basis of Its review, the Panel also noted that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. -The Panel included the information it gathered and assessed to reach its conclusion regarding the generic nature of the issue in Appendices Band C of this report. Should the NRC staff undertake a generic look of the issues, it should, among other things, consider the information presented and questions raised in those appendices.- The review should also include a reassessment of the information and staff positions communicated in RIS 2005-29, as well as those included in its proposed Revision 1, which is currently under development, to determine whether or not these documents include new staff positions with the potential for inappropriate or unintended backfitting. As part of any generic assessment, the Panel also recommends that staff determine whether the information in RIS 2005-29 and its proposed Revision 1 should be incorporated into a regulatory guide or another guidance document.
On the basis of Its review, the Panel also noted that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. -The Panel included the information it gathered and assessed to reach its conclusion regarding the generic nature of the issue in Appendices Band C of this report. Should the NRC staff undertake a generic look of the issues, it should, among other things, consider the information presented and questions raised in those appendices.- The review should also include a reassessment of the information and staff positions communicated in RIS 2005-29, as well as those included in its proposed Revision 1, which is currently under development, to determine whether or not these documents include new staff positions with the potential for inappropriate or unintended backfitting. As part of any generic assessment, the Panel also recommends that staff determine whether the information in RIS 2005-29 and its proposed Revision 1 should be incorporated into a regulatory guide or another guidance document.
APPENDIX A: HISTORY OF THE BACKFIT RULE AND THE COMPLIANCE EXCEPTION The Backfit Rule Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, "Backfitting," was originally promulgated in 1970.62 Because of perceived deficiencies in the rule, the U.S. Nuclear Regulatory Commission (NRC) substantially revised it in 1985.63 The 1985 rule was challenged in court, and the U.S. Circuit Court for the District of Columbia (D.C. Circuit) vacated this rule in its entirety. The D.C. Circuit took this action because it concluded that the revised rule could be interpreted to allow the NRC to consider costs in defining or redefining what is required for adequate protection of the public health and safety.64 In response, the NRC revised the Backfit Rule in 1988 to remove any implication that costs could be considered in defining or redefining adequate protection.65 The 1988 revisions only differed firom the 1985 rule to the extent necessary to address the court's concerns. The 1988 rule was also challenged in cou rt, but this time the D.C. Circuit upheld the rule.66 In its current form, 10 CFR 50.109(a)(1) defines backfitting as
APPENDIX A: HISTORY OF THE BACKFIT RULE AND THE COMPLIANCE EXCEPTION The Backfit Rule Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, "Backfitting," was originally promulgated in 1970.62 Because of perceived deficiencies in the rule, the U.S. Nuclear Regulatory Commission (NRC) substantially revised it in 1985.63 The 1985 rule was challenged in court, and the U.S. Circuit Court for the District of Columbia (D.C. Circuit) vacated this rule in its entirety. The D.C. Circuit took this action because it concluded that the revised rule could be interpreted to allow the NRC to consider costs in defining or redefining what is required for adequate protection of the public health and safety.64 In response, the NRC revised the Backfit Rule in 1988 to remove any implication that costs could be considered in defining or redefining adequate protection.65 The 1988 revisions only differed firom the 1985 rule to the extent necessary to address the court's concerns. The 1988 rule was also challenged in cou rt, but this time the D.C. Circuit upheld the rule.66 In its current form, 10 CFR 50.109(a)(1) defines backfitting as
           ... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the im position of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position ....
           ... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the im position of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position ....
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63 NRC 1985 e4 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 824 F.2d 108, 119-20 (1987).
63 NRC 1985 e4 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 824 F.2d 108, 119-20 (1987).
65 NRC 1988b 66 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 880 F.2d 552 (1989).
65 NRC 1988b 66 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 880 F.2d 552 (1989).
Commission Polley The Commission addressed its intended application of the compliance exception in the 1985 rulemaking :67 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.
Commission Polley The Commission addressed its intended application of the compliance exception in the 1985 rulemaking :67 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.
In the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."68 The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic] regulations."69 Backfitting Guidance Extensive information regarding the appropriate implementation of backfitting is provided in NUREG-1409.70 Relevant excerpts from this guidance are provided below.
In the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."68 The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic] regulations."69 Backfitting Guidance Extensive information regarding the appropriate implementation of backfitting is provided in NUREG-1409.70 Relevant excerpts from this guidance are provided below.
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* NRC staff positions that are documented explicit interpretations of more general regulations and are contained in documents such as the Standard Review Plan, branch technical positions, regulatory guides, generic letters, and bulletins e1 NRC 1985, at 38103 68  /d. at 38102 69
* NRC staff positions that are documented explicit interpretations of more general regulations and are contained in documents such as the Standard Review Plan, branch technical positions, regulatory guides, generic letters, and bulletins e1 NRC 1985, at 38103 68  /d. at 38102 69
     /d. at 38103. The 1988 rulemaking neither revised the compliance exception as stated in the 1985 rule nor provided additional guidance on its interpretation.
     /d. at 38103. The 1988 rulemaking neither revised the compliance exception as stated in the 1985 rule nor provided additional guidance on its interpretation.
70 NRC 1990c
70 NRC 1990c A similar list of examples is provided in Manual Chapter 0514,71 which is also included as Appendix D to NUREG-1409. Manual Chapter 0514 was referenced in the 1988 rulemaking, and a working draft was provided to the Commission for information in SECY-88-102.72 Manual Chapter 0514 provides a definition of "applicable regulatory staff positions" that is slightly more detailed than the definition in NUREG-1409. This definition from Manual Chapter 0514 is quoted below, with additional detail beyond NUREG-1409 emphasized in underlined text.
 
A similar list of examples is provided in Manual Chapter 0514,71 which is also included as Appendix D to NUREG-1409. Manual Chapter 0514 was referenced in the 1988 rulemaking, and a working draft was provided to the Commission for information in SECY-88-102.72 Manual Chapter 0514 provides a definition of "applicable regulatory staff positions" that is slightly more detailed than the definition in NUREG-1409. This definition from Manual Chapter 0514 is quoted below, with additional detail beyond NUREG-1409 emphasized in underlined text.
Applicable regulatory staff positions are those already specifically imposed upon or committed to by a licensee at the time of the identification of a plant-specific backfit, and are of several different types and sources:
Applicable regulatory staff positions are those already specifically imposed upon or committed to by a licensee at the time of the identification of a plant-specific backfit, and are of several different types and sources:
: a. Legal requirements such as in explicit regulations, orders, plant licenses (amendments, conditions, technical specifications). Note that some regulations have update features built in , as for example, 10 CFR 50.55a, Codes and Standards. Such update requirements are applicable as described in the regulation.
: a. Legal requirements such as in explicit regulations, orders, plant licenses (amendments, conditions, technical specifications). Note that some regulations have update features built in , as for example, 10 CFR 50.55a, Codes and Standards. Such update requirements are applicable as described in the regulation.
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Informal or formal communications to one licensee are not official positions to all licensees. Section 053 of Manual Chapter 0514 identifies what can be applied as official staff positions in a plant-specific context. They are legal requirements such as contained in explicit regulations, orders, and plant licenses; written commitments such as contained in final safety analysis reports, licenses event reports, and docketed correspondence; and documented, approved explicit interpretations such as contained in the [Standard Review Plan], branch technical 71 NRC 1988c 72 NRC 1988a 73 Requirements may be imposed by rule or order. Staff interpretations such as examples of acceptable ways to meet requirements are not requirements in and of themselves.
Informal or formal communications to one licensee are not official positions to all licensees. Section 053 of Manual Chapter 0514 identifies what can be applied as official staff positions in a plant-specific context. They are legal requirements such as contained in explicit regulations, orders, and plant licenses; written commitments such as contained in final safety analysis reports, licenses event reports, and docketed correspondence; and documented, approved explicit interpretations such as contained in the [Standard Review Plan], branch technical 71 NRC 1988c 72 NRC 1988a 73 Requirements may be imposed by rule or order. Staff interpretations such as examples of acceptable ways to meet requirements are not requirements in and of themselves.
74 Imposition of a staff position from which a licensee has previously been excepted is a backfit.
74 Imposition of a staff position from which a licensee has previously been excepted is a backfit.
positions, regulatory guides, generic letters, and bulletins. Orders, licenses, and written commitments are applicable only to a particular licensee.
positions, regulatory guides, generic letters, and bulletins. Orders, licenses, and written commitments are applicable only to a particular licensee.
If the NRC staff previously exempted a licensee from a legal requirement or approved position, it is not applicable to that licensee for the purpose of backfit consideration. Explicit exemption would be done formally in writing. The Appendix to NRC Manual Chapter 0514 discusses tacit approval under reanalysis of issues. Two situations are covered. In the first case, staff review of a previously accepted licensee action or program may result in a requested change. This would be classified as a backfit because it represents a change in a previous staff position and would require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed in the backfit rule). In the second case, a licensee submittal committing to a specific course of action that has not received timely NRC staff review is implemented by the licensee. In this case, It is considered that the NRC staff tacitly accepted the licensee's action since timely notice to the contrary was not given. If the NRC staff subsequently adopts a different position and requests a change in the licensee action, this change may be classified as a backfit and thus require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed In the backfit rule}.
If the NRC staff previously exempted a licensee from a legal requirement or approved position, it is not applicable to that licensee for the purpose of backfit consideration. Explicit exemption would be done formally in writing. The Appendix to NRC Manual Chapter 0514 discusses tacit approval under reanalysis of issues. Two situations are covered. In the first case, staff review of a previously accepted licensee action or program may result in a requested change. This would be classified as a backfit because it represents a change in a previous staff position and would require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed in the backfit rule). In the second case, a licensee submittal committing to a specific course of action that has not received timely NRC staff review is implemented by the licensee. In this case, It is considered that the NRC staff tacitly accepted the licensee's action since timely notice to the contrary was not given. If the NRC staff subsequently adopts a different position and requests a change in the licensee action, this change may be classified as a backfit and thus require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed In the backfit rule}.
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Explicit approval cou ld be provided in an Inspection report that states that a particular approach is acceptable. However, conclusions of that nature are usually made in [safety evaluations] rather than inspection reports.
Explicit approval cou ld be provided in an Inspection report that states that a particular approach is acceptable. However, conclusions of that nature are usually made in [safety evaluations] rather than inspection reports.
Comp/lance Backfit Guidance NUREG-1409 gives the following response to the question, "[h]ow does the backfit rule apply to new staff positions that reflect an evolving understanding of technical issues?"
Comp/lance Backfit Guidance NUREG-1409 gives the following response to the question, "[h]ow does the backfit rule apply to new staff positions that reflect an evolving understanding of technical issues?"
An evolving understanding of issues does not, by itself, define which category fits a particular backfit. Judgment must be applied to the facts of each particular case to determine whether the backfit is for compliance, to provide adequate protection, to redefine adequate protection, or to achieve a cost-justified substantial safety enhancement. For example, with regard to compliance, the 1985 statement of considerations for 1O CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact....new or modified interpretations of what constitutes compliance would not fall within the exception...."
An evolving understanding of issues does not, by itself, define which category fits a particular backfit. Judgment must be applied to the facts of each particular case to determine whether the backfit is for compliance, to provide adequate protection, to redefine adequate protection, or to achieve a cost-justified substantial safety enhancement. For example, with regard to compliance, the 1985 statement of considerations for 10 CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact....new or modified interpretations of what constitutes compliance would not fall within the exception...."
 
NUREG-1409 also provides. an example where an evolving understanding of technical issues resulted in a compliance backfit that was apparently justified for at least some licensees. In response to industry claims that Bulletin 88-11 75 lacked any backfitting justification, the NRC staff responded:
NUREG-1409 also provides. an example where an evolving understanding of technical issues resulted in a compliance backfit that was apparently justified for at least some licensees. In response to industry claims that Bulletin 88-11 75 lacked any backfitting justification, the NRC staff responded:
Although the justification was not printed in the bulletin, NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," was justified as a backfit. It is an example of a backfit that was determined by the responsible NRC official to be required as a matter of compliance with existing requirements and commitments.
Although the justification was not printed in the bulletin, NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," was justified as a backfit. It is an example of a backfit that was determined by the responsible NRC official to be required as a matter of compliance with existing requirements and commitments.
The CRGR reviewed the bulletin and concurred. The regulations currently require licensees to meet the applicable codes of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code. Because of the NRC staffs concern with the integrity of the surge line, licensees were requested to perform their fatigue analysis in accordance with the latest ASME Section Ill requirements that incorporate high cycle fatigue analysis. The justification provided by the NRC staff was that previously unconsidered thermal stratification phenomenon may invalidate the existing analysis performed to confirm the integrity of the surge line.
The CRGR reviewed the bulletin and concurred. The regulations currently require licensees to meet the applicable codes of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code. Because of the NRC staffs concern with the integrity of the surge line, licensees were requested to perform their fatigue analysis in accordance with the latest ASME Section Ill requirements that incorporate high cycle fatigue analysis. The justification provided by the NRC staff was that previously unconsidered thermal stratification phenomenon may invalidate the existing analysis performed to confirm the integrity of the surge line.
Subsequently, it was understood that some licensees believed that the NRC staff's rationale was in error because they were not committed to the latest ASME Section Ill requirements by virtue of their license commitment. However, the issue became moot because these licensees undertook the analysis voluntarily in view of the safety importance of the issue and the fact that previous versions of the ASME Code did not completely address the concern.
Subsequently, it was understood that some licensees believed that the NRC staff's rationale was in error because they were not committed to the latest ASME Section Ill requirements by virtue of their license commitment. However, the issue became moot because these licensees undertook the analysis voluntarily in view of the safety importance of the issue and the fact that previous versions of the ASME Code did not completely address the concern.
75 NRC 1988e
75 NRC 1988e APPENDIX B: QUALIFICATION OF PRESSURE RELIEF VALVES IN NUCLEAR POWER PLANTS IN RESPONSE TO THE TMl-2 ACCIDENT Byron and Braidwood Design and Code Requirements Nuclear power plants in the United States use various types of pressure relief valves to protect personnel and equipment from overpressure events within reactor fluid systems. Pressure relief valves include safety valves, safety relief valves, and relief valves, with different designs, operating conditions, and requirements. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), Section Ill, Division 1, specifies requirements for the design, operation, installation, and testing of pressure relief valves used for various functions in nuclear power plants.76 For example, the ASME BPV Code (2007 Edition) in Article NB-7000, Overpressure Protection, specifies requirements for several service conditions:
 
APPENDIX B: QUALIFICATION OF PRESSURE RELIEF VALVES IN NUCLEAR POWER PLANTS IN RESPONSE TO THE TMl-2 ACCIDENT Byron and Braidwood Design and Code Requirements Nuclear power plants in the United States use various types of pressure relief valves to protect personnel and equipment from overpressure events within reactor fluid systems. Pressure relief valves include safety valves, safety relief valves, and relief valves, with different designs, operating conditions, and requirements. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), Section Ill, Division 1, specifies requirements for the design, operation, installation, and testing of pressure relief valves used for various functions in nuclear power plants.76 For example, the ASME BPV Code (2007 Edition) in Article NB-7000, Overpressure Protection, specifies requirements for several service conditions:
* steam and air or gas service for safety valves;
* steam and air or gas service for safety valves;
* steam, air or gas, and liquid service for safety relief valves;
* steam, air or gas, and liquid service for safety relief valves;
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A. Performance Testing of Relief and Safety Valves-The following information must be provided in report form by October 1, 1981 :
A. Performance Testing of Relief and Safety Valves-The following information must be provided in report form by October 1, 1981 :
77 NRC 1981 band NRC 1981 c 78 NRC 2007b and NRC 2007c 79 NRC 1979a 80 NRC 1980b and NRC 1980c
77 NRC 1981 band NRC 1981 c 78 NRC 2007b and NRC 2007c 79 NRC 1979a 80 NRC 1980b and NRC 1980c
( 1) Evidence supported by test of safety and relief valve functionability for expected operating and accident (non-[anticipated transient without scram])
( 1) Evidence supported by test of safety and relief valve functionability for expected operating and accident (non-[anticipated transient without scram])
conditions must be provided to NRC. The testing should demonstrate that the valves will open and reclose under the expected flow conditions.
conditions must be provided to NRC. The testing should demonstrate that the valves will open and reclose under the expected flow conditions.
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Pre-implementation review will be based on EPRI, BWR, and applicant submittals with regard to the various test programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following valve qualification dates can be met:
Pre-implementation review will be based on EPRI, BWR, and applicant submittals with regard to the various test programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following valve qualification dates can be met:
Final PWR (EPRI) Test Program--July 1, 1980 Final BWR Test Program--October 1, 1980 Block Valve Qualification Program--Janua ry 1, 1981 Post-implementation review will be based on the applicants' plant-specific submittals for qualification of safety relief valves and block valves. To properly evaluate these plant-specific applications, the test data and results of the various programs will also be required by the following dates:
Final PWR (EPRI) Test Program--July 1, 1980 Final BWR Test Program--October 1, 1980 Block Valve Qualification Program--Janua ry 1, 1981 Post-implementation review will be based on the applicants' plant-specific submittals for qualification of safety relief valves and block valves. To properly evaluate these plant-specific applications, the test data and results of the various programs will also be required by the following dates:
PWR (EPRl)/BWR Generic Test Program Results--July 1, 1981 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee/applicant preliminary review of generic test program results--July 1, 1981
PWR (EPRl)/BWR Generic Test Program Results--July 1, 1981 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee/applicant preliminary review of generic test program results--July 1, 1981 Plant-specific reports for safety and relief valve qualification--October 1, 1981 Plant-specific submittals for piping and support evaluations--January 1, 1982 Plant-specific submittals for block valve qualification--July 1, 1982 EPRI Testing In October 1982, EPRI issued NP-2670-LD to address testing of PORVs.81 This report has been referenced by certain licensees (e.g., Section 15.2.14 of the North Anna, Units 1 and 2 Updated Final Safety Analysis Report (UFSAR)B2 ).
 
Plant-specific reports for safety and relief valve qualification--October 1, 1981 Plant-specific submittals for piping and support evaluations--January 1, 1982 Plant-specific submittals for block valve qualification--July 1, 1982 EPRI Testing In October 1982, EPRI issued NP-2670-LD to address testing of PORVs.81 This report has been referenced by certain licensees (e.g., Section 15.2.14 of the North Anna, Units 1 and 2 Updated Final Safety Analysis Report (UFSAR)B2 ).
In December 1982, EPRI issued NP-2628-SR, which described safety and relief valve tests for types of valves in service at nuclear power plants.83 In particular, Section 3.5 documented the testing of Crosby safety valves similar to the PSVs at Byron and Braidwood, including two water tests. The report indicated chattering of the safety valves, with subsequent inspection finding galled surfaces and damage to internal parts. Section 4.6 documented testing of Copes-Vulcan relief valves similar to the pressurizer PORVs at Byron and Braidwood, although the extent of water testing was not fully described. The report indicated no damage found during the inspection of the Copes-Vulcan relief valves. The report did not indicate any failures of the Crosby or Copes-Vulcan valves to reseat after discharging water during the testing.
In December 1982, EPRI issued NP-2628-SR, which described safety and relief valve tests for types of valves in service at nuclear power plants.83 In particular, Section 3.5 documented the testing of Crosby safety valves similar to the PSVs at Byron and Braidwood, including two water tests. The report indicated chattering of the safety valves, with subsequent inspection finding galled surfaces and damage to internal parts. Section 4.6 documented testing of Copes-Vulcan relief valves similar to the pressurizer PORVs at Byron and Braidwood, although the extent of water testing was not fully described. The report indicated no damage found during the inspection of the Copes-Vulcan relief valves. The report did not indicate any failures of the Crosby or Copes-Vulcan valves to reseat after discharging water during the testing.
EPRI also published NP-2770-LD in the early 1980s to describe the testing of PWR primary system safety valves. Volume 1, issued in December 1982, provides a summary of the test program and its results.84 Section 4.5 of Volume 1 Indicates that the following tests were performed on the Crosby 6M6 PSV: 11 steam tests with lfilled loop seals, 3 steam-to-water transition tests, and 2 water tests. The report states that the valve experienced chatter during the tests, and one water test had to be terminated. The individual volumes of EPRI NP-2770-LD discuss the test results for each specific PSV type. Volume 6, issued in March 1983, provides the test details for the Crosby 6M6 PSV.
EPRI also published NP-2770-LD in the early 1980s to describe the testing of PWR primary system safety valves. Volume 1, issued in December 1982, provides a summary of the test program and its results.84 Section 4.5 of Volume 1 Indicates that the following tests were performed on the Crosby 6M6 PSV: 11 steam tests with lfilled loop seals, 3 steam-to-water transition tests, and 2 water tests. The report states that the valve experienced chatter during the tests, and one water test had to be terminated. The individual volumes of EPRI NP-2770-LD discuss the test results for each specific PSV type. Volume 6, issued in March 1983, provides the test details for the Crosby 6M6 PSV.
Westinghouse Evaluation of EPRI Testing In July 1982, the Westinghouse Owners Group (WOG) submitted WCAP-10105.85 In WCAP-10105, the WOG indicated that the design specification for PSVs in Westinghouse-designed nuclear power plants is for steam service only. Based on a review of the EPRI test data, the WOG concluded that the valves performed with chatter, but did not identify any valve damage.
Westinghouse Evaluation of EPRI Testing In July 1982, the Westinghouse Owners Group (WOG) submitted WCAP-10105.85 In WCAP-10105, the WOG indicated that the design specification for PSVs in Westinghouse-designed nuclear power plants is for steam service only. Based on a review of the EPRI test data, the WOG concluded that the valves performed with chatter, but did not identify any valve damage.
In January 1988, Westinghouse issued WCAP-11677, which compared the EPRI test data with feedwater line break safety analyses.86 Westinghouse determined that all nuclear power plants addressed in the EPRI testing had PSVs that would operate reliably during water discharge.
In January 1988, Westinghouse issued WCAP-11677, which compared the EPRI test data with feedwater line break safety analyses.86 Westinghouse determined that all nuclear power plants addressed in the EPRI testing had PSVs that would operate reliably during water discharge.
Westinghouse evaluated the performance of the Crosby 6M6 PSVs during the EPRI tests, and 81 EPRI 1982a 62 VEPCO 2015 63 EPRI 1982b 64 EPRI 1982c as woG 1982 86 Westinghouse 1988
Westinghouse evaluated the performance of the Crosby 6M6 PSVs during the EPRI tests, and 81 EPRI 1982a 62 VEPCO 2015 63 EPRI 1982b 64 EPRI 1982c as woG 1982 86 Westinghouse 1988 considered that the performance involved less significant flutter (half lift motion) than the chatter (full lift motion) determined in the EPRI report. Westinghouse concluded that the Crosby 6M6 PSV can pass slightly subcooled water at a minimum up to three times without damage.
 
considered that the performance involved less significant flutter (half lift motion) than the chatter (full lift motion) determined in the EPRI report. Westinghouse concluded that the Crosby 6M6 PSV can pass slightly subcooled water at a minimum up to three times without damage.
Byron and Braidwood Licensing and Response to TMI Requirements The NRC safety evaluation reports (SERs) associated with the issuance of the operating licenses for Byron and Braidwood included evaluation of the TMI Action Plan items. 87 In the introduction to the Braidwood SER, the NRC staff stated that the review and evaluation of compliance by the applicant with the licensing requirements established in NUREG-066088 and TMI Action Plan Item 11.D.1 were incorporated into the reviews summarized throughout the SER.
Byron and Braidwood Licensing and Response to TMI Requirements The NRC safety evaluation reports (SERs) associated with the issuance of the operating licenses for Byron and Braidwood included evaluation of the TMI Action Plan items. 87 In the introduction to the Braidwood SER, the NRC staff stated that the review and evaluation of compliance by the applicant with the licensing requirements established in NUREG-066088 and TMI Action Plan Item 11.D.1 were incorporated into the reviews summarized throughout the SER.
Appendix E, "Requirements Resulting from TMl-2 Accident," to the Byron and Braidwood UFSAR in Section E.23, "Relief and Safety Valve Test Requirements (11.D.1 )," references the 1982 transmittal from Consumers Power of a test report for the EPRI safety and relief valve test program.89 The UFSAR states that the final evaluation of the data indicated that the relief and safety valves will perform their intended functions for all expected fluid inlet conditions. The UFSAR also references the October 1982 licensee evaluation of the adequacy of the relief and safety valves that had been submitted to the NRC.90 In Supplement 1 to the Braidwood SER,91 in Section 3.9.3.3, "Design and Installation of Pressure Relief Devices," the NRC staff stated that EPRI had completed a full-scale valve testing program and referenced the July 1982 submittal of WCAP-10105. The NRC staff stated that the applicant responded to a requirement to demonstrate operability of these valves through submittals dated July 1, 1982, October 26, 1982, and December 30, 1983. On the basis of a preliminary review, the NRC staff concluded that the applicant's general approach to responding to this item was acceptable, and provided adequate assurance that the RCS overpressure protection systems at Braidwood could adequately perform their intended functions. The NRC staff stated that if the detailed review revealed that modifications or adjustments to safety valves, PORVs, PORV block valves, or associated piping, would be needed to ensure that all intended design margins were present, the NRG staff would require that the applicant make appropriate modifications. The NRC staff categorized this issue as a Confirmatory Item. The NRC issued operating licenses for all four Byron and Braidwood Units between February 1985 and May 1988.
Appendix E, "Requirements Resulting from TMl-2 Accident," to the Byron and Braidwood UFSAR in Section E.23, "Relief and Safety Valve Test Requirements (11.D.1 )," references the 1982 transmittal from Consumers Power of a test report for the EPRI safety and relief valve test program.89 The UFSAR states that the final evaluation of the data indicated that the relief and safety valves will perform their intended functions for all expected fluid inlet conditions. The UFSAR also references the October 1982 licensee evaluation of the adequacy of the relief and safety valves that had been submitted to the NRC.90 In Supplement 1 to the Braidwood SER,91 in Section 3.9.3.3, "Design and Installation of Pressure Relief Devices," the NRC staff stated that EPRI had completed a full-scale valve testing program and referenced the July 1982 submittal of WCAP-10105. The NRC staff stated that the applicant responded to a requirement to demonstrate operability of these valves through submittals dated July 1, 1982, October 26, 1982, and December 30, 1983. On the basis of a preliminary review, the NRC staff concluded that the applicant's general approach to responding to this item was acceptable, and provided adequate assurance that the RCS overpressure protection systems at Braidwood could adequately perform their intended functions. The NRC staff stated that if the detailed review revealed that modifications or adjustments to safety valves, PORVs, PORV block valves, or associated piping, would be needed to ensure that all intended design margins were present, the NRG staff would require that the applicant make appropriate modifications. The NRC staff categorized this issue as a Confirmatory Item. The NRC issued operating licenses for all four Byron and Braidwood Units between February 1985 and May 1988.
Closure of TMI Action Plan Item 11.D.1 for Byron and Braidwood Following the issuance of the operating licenses, the NRC staff documented its review of the response to TMI Action Plan Item 11.D.1 for Byron and Braidwood via two letters that transmitted similar Technical Evaluation Reports (TERs) developed by Idaho National Engineering Laboratory (INEL).92 In its letters, the NRC staff indicated that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The TERs described the INEL review of the EPRI testing of PSVs and PORVs 87 NRC 1983 and NRC 1986b (Braidwood), NRC 1984 and NRC 1987a (Byron) 88  NRC 1980a S9 Consumers 1982 9
Closure of TMI Action Plan Item 11.D.1 for Byron and Braidwood Following the issuance of the operating licenses, the NRC staff documented its review of the response to TMI Action Plan Item 11.D.1 for Byron and Braidwood via two letters that transmitted similar Technical Evaluation Reports (TERs) developed by Idaho National Engineering Laboratory (INEL).92 In its letters, the NRC staff indicated that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The TERs described the INEL review of the EPRI testing of PSVs and PORVs 87 NRC 1983 and NRC 1986b (Braidwood), NRC 1984 and NRC 1987a (Byron) 88  NRC 1980a S9 Consumers 1982 9
   &deg; ComEd 1982 91 NRC 1986b. Similar discussion appears in NRC 1984 for Byron, and NRG 1987a (also for Byron) states that TMI Action Plan Item 11.D.1 had been closed in NRC 1984.
   &deg; ComEd 1982 91 NRC 1986b. Similar discussion appears in NRC 1984 for Byron, and NRG 1987a (also for Byron) states that TMI Action Plan Item 11.D.1 had been closed in NRC 1984.
92 NRG 1988c (Byron) and NRG 1990a (Braidwood)
92 NRG 1988c (Byron) and NRG 1990a (Braidwood) similar to the Byron and Braidwood pressurizer valves. The TERs concluded that Byron and Braidwood had provided an acceptable response to TMI Action Plan Item 11.D.1.
 
similar to the Byron and Braidwood pressurizer valves. The TERs concluded that Byron and Braidwood had provided an acceptable response to TMI Action Plan Item 11.D.1.
Section 4.2.3, "Extended High Pressure Injection [HPI] Event," of the TERs stated that the potential for water discharge in extended HPI events can be disregarded for an extended high pressure injection event because at least 20 minutes would be available for operator action.
Section 4.2.3, "Extended High Pressure Injection [HPI] Event," of the TERs stated that the potential for water discharge in extended HPI events can be disregarded for an extended high pressure injection event because at least 20 minutes would be available for operator action.
Water discharge was evaluated, however, in Section 4.2.2, "FSAR Liquid Transients," of the TERs. This section discussed the evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs.
Water discharge was evaluated, however, in Section 4.2.2, "FSAR Liquid Transients," of the TERs. This section discussed the evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs.
In addition, Section 4.3.1 , "Safety Valves," and Section 4.3.2, "Power Operated Relief Valves,"
In addition, Section 4.3.1 , "Safety Valves," and Section 4.3.2, "Power Operated Relief Valves,"
of the TERs determined that the performance of the PSVs and PORVs was acceptable based on the EPRI tests, including water discharge tests. The TERs indicated that the PSV had two applicable tests: a loop seal steam-water transition test where the valve opened, chattered and stabilized to close; and a saturated water test where the valve opened with water, chattered, and stabilized. The TERs indicated that the PORV opened and closed on demand in the loop seal steam-water transition test, with a bending moment that was evaluated by analysis.
of the TERs determined that the performance of the PSVs and PORVs was acceptable based on the EPRI tests, including water discharge tests. The TERs indicated that the PSV had two applicable tests: a loop seal steam-water transition test where the valve opened, chattered and stabilized to close; and a saturated water test where the valve opened with water, chattered, and stabilized. The TERs indicated that the PORV opened and closed on demand in the loop seal steam-water transition test, with a bending moment that was evaluated by analysis.
APPENDIX C: CONCERNS REGARDING PERFORMANCE OF PRESSURIZER VALVES UNDER WATER FLOW CONDITIONS Westinghouse Nuclear Safety Advisory Letter In 1993 and 1994, Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 93-013 and its Supplement 1 to operating nuclear power plants (including Byron and Braidwood). 93 These advisories resulted from Westinghouse's discovery that potentially nonconservative assumptions were used in the licensing analysis of the Inadvertent Operation of the Emergency Core Cooling System at Power (IOECCS) event.
APPENDIX C: CONCERNS REGARDING PERFORMANCE OF PRESSURIZER VALVES UNDER WATER FLOW CONDITIONS Westinghouse Nuclear Safety Advisory Letter In 1993 and 1994, Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 93-013 and its Supplement 1 to operating nuclear power plants (including Byron and Braidwood). 93 These advisories resulted from Westinghouse's discovery that potentially nonconservative assumptions were used in the licensing analysis of the Inadvertent Operation of the Emergency Core Cooling System at Power (IOECCS) event.
In NSAL-93-013, Westinghouse recommended that licensees determine whether their pressurizer safety relief valves (PSRVs)~ are capable of closing following discharge of subcooled water. Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse indicated that water discharge through the power-operated relief valves (PORVs) is not a concern, becaus,e the PORV block valves can be used to isolate the PORVs if they fail to close. If the PSRVs are not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees re-evaluate the IOECCS event with three possible options of (1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the event.
In NSAL-93-013, Westinghouse recommended that licensees determine whether their pressurizer safety relief valves (PSRVs)~ are capable of closing following discharge of subcooled water. Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse indicated that water discharge through the power-operated relief valves (PORVs) is not a concern, becaus,e the PORV block valves can be used to isolate the PORVs if they fail to close. If the PSRVs are not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees re-evaluate the IOECCS event with three possible options of (1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the event.
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In 2005, the NRC issued Regulatory Issue Summary (RIS) 2005-29 to notify nuclear power plant licensees of a concern identified during reviews of power uprate requests. 96 In RIS 2005-29, the NRC staff stated that typically Condition II scenarJos97 involve discharging water through relief or safety valves that are not qualified for water discharge. The NRC staff stated that these valves are then assumed to fail in the open position and create a small-break loss-of-coolant 93  Westinghouse 1993 and Westinghouse 1994 94  Although W estinghouse used the term PSRVs. the specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code. and by the licensee.
In 2005, the NRC issued Regulatory Issue Summary (RIS) 2005-29 to notify nuclear power plant licensees of a concern identified during reviews of power uprate requests. 96 In RIS 2005-29, the NRC staff stated that typically Condition II scenarJos97 involve discharging water through relief or safety valves that are not qualified for water discharge. The NRC staff stated that these valves are then assumed to fail in the open position and create a small-break loss-of-coolant 93  Westinghouse 1993 and Westinghouse 1994 94  Although W estinghouse used the term PSRVs. the specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code. and by the licensee.
95 NRC 2003 96 NRC 2005b 97 As defined in American Nuclear Society (ANS) Standard 51 .1/N18.2-1973 (ANS 1973).
95 NRC 2003 96 NRC 2005b 97 As defined in American Nuclear Society (ANS) Standard 51 .1/N18.2-1973 (ANS 1973).
accident (LOCA). The NRC staff stated that it was concerned that some licensees may be crediting PORVs without qualification for water discharge and without establishing additional restrictions to ensure the availability of PORVs and block valves. The NRC staff stated that the advice in Westinghouse NSAL-93-013 to use the PORV block valves to isolate the PORVs is inconsistent with the non-escalation position.
accident (LOCA). The NRC staff stated that it was concerned that some licensees may be crediting PORVs without qualification for water discharge and without establishing additional restrictions to ensure the availability of PORVs and block valves. The NRC staff stated that the advice in Westinghouse NSAL-93-013 to use the PORV block valves to isolate the PORVs is inconsistent with the non-escalation position.
In draft Revision 1 to RIS 2005-29, the NRC staff addresses the specific ANS Condition II scenarios of chemical volume and control system (CVCS) malfunction, inadvertent opening of a PORV or PSV (IOPSRV), and the IOECCS event.98 Regarding the eves malfunction, the NRC staff states that performing only a reactivity anomaly analysis or assuming that this malfunction is not as severe as the IOECCS event is not acceptable. Regarding the IOPSRV event, the NRC staff stated that inadvertent opening of PSV or PORV could continue as an ANS Condition Ill small break LOCA and fails to meet the non-escalation position. Regarding the IOECCS event, the NRC staff states that five of the alternative approaches in NSAL-93-013 fail to meet the non-escalation position. The NRC staff indicated that these unacceptable alternative approaches are:
In draft Revision 1 to RIS 2005-29, the NRC staff addresses the specific ANS Condition II scenarios of chemical volume and control system (CVCS) malfunction, inadvertent opening of a PORV or PSV (IOPSRV), and the IOECCS event.98 Regarding the eves malfunction, the NRC staff states that performing only a reactivity anomaly analysis or assuming that this malfunction is not as severe as the IOECCS event is not acceptable. Regarding the IOPSRV event, the NRC staff stated that inadvertent opening of PSV or PORV could continue as an ANS Condition Ill small break LOCA and fails to meet the non-escalation position. Regarding the IOECCS event, the NRC staff states that five of the alternative approaches in NSAL-93-013 fail to meet the non-escalation position. The NRC staff indicated that these unacceptable alternative approaches are:
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Additional General PSV/PORV Information In 2004, EPRI issued Technical Report 1011047, which evaluated the potential increase in failure rates following steam and liquid relief through safety valves based on expert judgement.99 The report found that the increase in failure rates is difficult to estimate because of limited data.
Additional General PSV/PORV Information In 2004, EPRI issued Technical Report 1011047, which evaluated the potential increase in failure rates following steam and liquid relief through safety valves based on expert judgement.99 The report found that the increase in failure rates is difficult to estimate because of limited data.
However, the experts considered that repeated water discharge through safety valves might cause increased chatter, and therefore, an increased failure rate.
However, the experts considered that repeated water discharge through safety valves might cause increased chatter, and therefore, an increased failure rate.
In 2011 , the NRC summarized relief valve performance data in NUREG/CR-7037, based on a study by the Idaho National Laboratory. 100 With respect to pressurizer PORVs, the report found four separate water discharge events at four PWR plants. The report estimated 698 total demands on these PORVs during their water discharge events with no failures to close. The report also summarized test data for three valve types from the Equipment Performance and Information Exchange (EPIX) database maintained by the Institute of Nuclear Power Operations. The report indicates two failures of PORVs to reclose during 2070 demands, but does not specify water or steam service for the EPIX test information. With respect to PSVs, the report indicates two failures out of four total demands following plant scrams, but does not indicate water or steam service. Following a request by the Panel, NRC staff from the Office of Nuclear Regulatory Research provided Licensee Event Report information indicating that the two PSV failures involved incomplete reseating of the valves with leakage of 25 and 200 gallons 98 NRC 2015a 99 EPRI 2004 100 NRC 2011
In 2011 , the NRC summarized relief valve performance data in NUREG/CR-7037, based on a study by the Idaho National Laboratory. 100 With respect to pressurizer PORVs, the report found four separate water discharge events at four PWR plants. The report estimated 698 total demands on these PORVs during their water discharge events with no failures to close. The report also summarized test data for three valve types from the Equipment Performance and Information Exchange (EPIX) database maintained by the Institute of Nuclear Power Operations. The report indicates two failures of PORVs to reclose during 2070 demands, but does not specify water or steam service for the EPIX test information. With respect to PSVs, the report indicates two failures out of four total demands following plant scrams, but does not indicate water or steam service. Following a request by the Panel, NRC staff from the Office of Nuclear Regulatory Research provided Licensee Event Report information indicating that the two PSV failures involved incomplete reseating of the valves with leakage of 25 and 200 gallons 98 NRC 2015a 99 EPRI 2004 100 NRC 2011 per minute, respectively. The report summarized EPIX test data for PSVs as no failures to reclose during 1805 demands.
 
per minute, respectively. The report summarized EPIX test data for PSVs as no failures to reclose during 1805 demands.
Plant-Specific Actions Diablo Canyon In 1996, the licensee for Diablo Canyon Power Plant (Diablo Canyon) submitted a report of its evaluation under Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, "Changes, tests and experiments," of the potential for an IOECCS event. 101 The submittal included NSAL-93-013 and its Supplement 1 as enclosures. The licensee indicated that the PSVs had not been initially qualified for water discharge, but were subsequently qualified to discharge water for a brief period. The licensee indicated that WCAP-11677 (which evaluated the EPRI testing) was applicable and demonstrated that the PSVs were operable.
Plant-Specific Actions Diablo Canyon In 1996, the licensee for Diablo Canyon Power Plant (Diablo Canyon) submitted a report of its evaluation under Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, "Changes, tests and experiments," of the potential for an IOECCS event. 101 The submittal included NSAL-93-013 and its Supplement 1 as enclosures. The licensee indicated that the PSVs had not been initially qualified for water discharge, but were subsequently qualified to discharge water for a brief period. The licensee indicated that WCAP-11677 (which evaluated the EPRI testing) was applicable and demonstrated that the PSVs were operable.
In 2004, the NRC issued a license amendment for Diablo Canyon that allowed credit for actuation of the PORVs in response to inadvertent safety injection (SI) actuation, to avoid challenges to the PSVs. 102 To support the NRC staff's review, the licensee submitted additional information related to the capability of the PORVs to function adequately under conditions predicted for design-basis transients and accidents. 103 In response to a question regarding the design adequacy of the PORVs if the pressurizer becomes water solid, the licensee s.tated that the PORV had no requirements for ASME BPV Code certification, but referenced a January 1986 NRC letter that had accepted the adequacy of the PORV and block valve design and confirmatory testing for a range of fluid conditions (full pressure steam, steam to water transition, and subcooled water fluid).104 Salem In 1997, the NRC issued a license amendment revising the technical specification (TS) for Salem Nuclear Generating Station, Units 1 and 2 (Salem) to ensure that the automatic capability of the PORVs to relieve pressure would be malntained. 1015 In response to NSAL-93-013, the licensee determined that an inadvertent SI actuation at power could cause the pressurizer to become water solid. The PSVs would lift and discharge water if the automatic operation of the PORVs were not made available for reactor coolant system (RCS) depressurization early in the transient. In that the Salem PSVs were not designed to relieve water, it was noted that water discharge could cause the PSVs to fail in the open position.
In 2004, the NRC issued a license amendment for Diablo Canyon that allowed credit for actuation of the PORVs in response to inadvertent safety injection (SI) actuation, to avoid challenges to the PSVs. 102 To support the NRC staff's review, the licensee submitted additional information related to the capability of the PORVs to function adequately under conditions predicted for design-basis transients and accidents. 103 In response to a question regarding the design adequacy of the PORVs if the pressurizer becomes water solid, the licensee s.tated that the PORV had no requirements for ASME BPV Code certification, but referenced a January 1986 NRC letter that had accepted the adequacy of the PORV and block valve design and confirmatory testing for a range of fluid conditions (full pressure steam, steam to water transition, and subcooled water fluid).104 Salem In 1997, the NRC issued a license amendment revising the technical specification (TS) for Salem Nuclear Generating Station, Units 1 and 2 (Salem) to ensure that the automatic capability of the PORVs to relieve pressure would be malntained. 1015 In response to NSAL-93-013, the licensee determined that an inadvertent SI actuation at power could cause the pressurizer to become water solid. The PSVs would lift and discharge water if the automatic operation of the PORVs were not made available for reactor coolant system (RCS) depressurization early in the transient. In that the Salem PSVs were not designed to relieve water, it was noted that water discharge could cause the PSVs to fail in the open position.
During the review, the NRC staff noted that the PORVs were not designed to "safety related" standards and, thus, could not be credited for automatic mitigation of an inadvertent SI actuation at power. In response, the licensee proposed an upgrade of the PORVs to eliminate the possibility that a single active failure of a PORV component could prevent the mitigation of an inadvertent SI actuation at power. As discussed in the NRC staff's safety evaluation (SE), the licensee implemented modifications to the PORV circuitry to qualify the upgraded circuitry as safety-related.
During the review, the NRC staff noted that the PORVs were not designed to "safety related" standards and, thus, could not be credited for automatic mitigation of an inadvertent SI actuation at power. In response, the licensee proposed an upgrade of the PORVs to eliminate the possibility that a single active failure of a PORV component could prevent the mitigation of an inadvertent SI actuation at power. As discussed in the NRC staff's safety evaluation (SE), the licensee implemented modifications to the PORV circuitry to qualify the upgraded circuitry as safety-related.
10 1 PG&E 1996 10 2 NRC 2004a 103  PG&E 2003 104 NRC 1986a 10s NRC 1997
10 1 PG&E 1996 10 2 NRC 2004a 103  PG&E 2003 104 NRC 1986a 10s NRC 1997 Regarding PORV performance, the licensee evaluated the PORV air accumulators and determined that they had sufficient capacity for the inadvertent SI event. The licensee also reported that endurance tests had been performed with five different trims (with different trim materials) on one PORV at Wyle Laboratories to demonstrate that (1) after 2000 consecutive operations, there were no packing leaks or packing gland adjustments required; (2) there was no diaphragm failure; and (3) the solenoid valve withstood 10,000 operations without any loss of function. Based on this information. the NRC staff concluded that the PORV performance was acceptable to mitigate an inadvertent SI event.
 
Regarding PORV performance, the licensee evaluated the PORV air accumulators and determined that they had sufficient capacity for the inadvertent SI event. The licensee also reported that endurance tests had been performed with five different trims (with different trim materials) on one PORV at Wyle Laboratories to demonstrate that (1) after 2000 consecutive operations, there were no packing leaks or packing gland adjustments required; (2) there was no diaphragm failure; and (3) the solenoid valve withstood 10,000 operations without any loss of function. Based on this information. the NRC staff concluded that the PORV performance was acceptable to mitigate an inadvertent SI event.
Millstone 3 In 1998, the NRC issued a license amendment for Millstone Nuclear Power Station, Unit 3 (Millstone 3) that revised the TS to ensure that the capability of the PORVs to relieve pressure would be maintained.106 The revised TS Bases stated that the PORVs and their associated piping had been demonstrated to be "qualified" for water discharge. The PORVs would prevent water discharge from the PSVs, for which qualification for water discharge had not been demonstrated. The TS Bases also stated that the prime i'mportance for the capability to close the block valve is to isolate a stuck-open PORV. In the SE, the NRC staff referenced a December 1997 Licensee Event Report that notified the NRC of the issue of potential failure of PSVs following water discharge.10 7 As part of this license amendment, the licensee upgraded the PORV circuitry, added additional PORV surveillance requirements, qualified the PORVs and associated piping for water discharge, and revised emergency procedures to allow plant operators additional time to terminate the event. With respect to the PORV circuitry, the NRC staff concluded that the PORV circuitry modifications qualifiied the PORV control circuitry as safety-related. With respect to PORV performance, the licensee reanalyzed the inadvertent SI event with the LOFTRAN computer code to determine the time available for operator action to make a PORV available and provide the mass and energy releases needed to qualify the PORVs and associated piping for water discharge. The licensee referenced EPRI testing that was said to generically resolve TMI Action Plan Items associated with PORVs and safety valve qualification for water and steam discharge, specifically the results from four tests of a Garrett PORV (such as used at Millstone 3) for water discharge.108 The licensee determined that the PORVs and associated piping are qualified for 1 hour of water discharge for an IOECCS event. The licensee also stated that the PORV manufacturer performed numerous cycle tests to verify the performance of the valve design, and also verified that valve seat leakage was acceptable. The licensee stated that the PORV block valves had been evaluated for water discharge in accordance with the program established in response to Generic Letter (GL) 89-10.109 The NRC staff found the licensee information regarding the qualification of the PORVs for water discharge during the inadvertent SI event to be acceptable.
Millstone 3 In 1998, the NRC issued a license amendment for Millstone Nuclear Power Station, Unit 3 (Millstone 3) that revised the TS to ensure that the capability of the PORVs to relieve pressure would be maintained.106 The revised TS Bases stated that the PORVs and their associated piping had been demonstrated to be "qualified" for water discharge. The PORVs would prevent water discharge from the PSVs, for which qualification for water discharge had not been demonstrated. The TS Bases also stated that the prime i'mportance for the capability to close the block valve is to isolate a stuck-open PORV. In the SE, the NRC staff referenced a December 1997 Licensee Event Report that notified the NRC of the issue of potential failure of PSVs following water discharge.10 7 As part of this license amendment, the licensee upgraded the PORV circuitry, added additional PORV surveillance requirements, qualified the PORVs and associated piping for water discharge, and revised emergency procedures to allow plant operators additional time to terminate the event. With respect to the PORV circuitry, the NRC staff concluded that the PORV circuitry modifications qualifiied the PORV control circuitry as safety-related. With respect to PORV performance, the licensee reanalyzed the inadvertent SI event with the LOFTRAN computer code to determine the time available for operator action to make a PORV available and provide the mass and energy releases needed to qualify the PORVs and associated piping for water discharge. The licensee referenced EPRI testing that was said to generically resolve TMI Action Plan Items associated with PORVs and safety valve qualification for water and steam discharge, specifically the results from four tests of a Garrett PORV (such as used at Millstone 3) for water discharge.108 The licensee determined that the PORVs and associated piping are qualified for 1 hour of water discharge for an IOECCS event. The licensee also stated that the PORV manufacturer performed numerous cycle tests to verify the performance of the valve design, and also verified that valve seat leakage was acceptable. The licensee stated that the PORV block valves had been evaluated for water discharge in accordance with the program established in response to Generic Letter (GL) 89-10.109 The NRC staff found the licensee information regarding the qualification of the PORVs for water discharge during the inadvertent SI event to be acceptable.
106  NRC 1998 10 7 Northeast 1997 10a EPRI 1982a (Volume 11) 109 NRC 1989
106  NRC 1998 10 7 Northeast 1997 10a EPRI 1982a (Volume 11) 109 NRC 1989 Callaway In 2000, the NRG issued a license amendment for Callaway Plant, Unit 1 (Callaway) that revised the TS to change the PSV lift setting range. 110 The changes also credited automatic actuation of at least one PORV during an IOECCS event to prevent water discharge through the PSVs; to enable this credit, the licensee modified and upgraded the PORV circuitry to full Class 1E. In its license amendment request, 111 the licensee had stated that the design function of the valves was not being changed and the conclusions documented in the NRG staff's previous evaluation of Callaway's response to TMI Action Plan Item II.D.1112 were also unchanged. As a result, the licensee stated that the PORVs and associated discharge piping can accommodate water discharge.
 
Callaway In 2000, the NRG issued a license amendment for Callaway Plant, Unit 1 (Callaway) that revised the TS to change the PSV lift setting range. 110 The changes also credited automatic actuation of at least one PORV during an IOECCS event to prevent water discharge through the PSVs; to enable this credit, the licensee modified and upgraded the PORV circuitry to full Class 1E. In its license amendment request, 111 the licensee had stated that the design function of the valves was not being changed and the conclusions documented in the NRG staff's previous evaluation of Callaway's response to TMI Action Plan Item II.D.1112 were also unchanged. As a result, the licensee stated that the PORVs and associated discharge piping can accommodate water discharge.
Byron and Braidwood In 1998, the licensee for Byron and Braidwood requested an amendment to its TS to take credit for automatic operation of the PORVs to mitigate an IOECCS event. 113 In the amendment request, the licensee stated that the PSVs had not been qualified to reseat after passing subcooled liquid. The licensee stated that the PORVs at Byron and Braidwood are safety-related components with safety-related actuators and accumulator tanks, with PORV control circuits classified as safety-related. The licensee noted that some portions of the PORV circuitry are nonsafety-related, with improvements implemented in response to GL 90-06. 114 The licensee stated that the PORV block valves are within the scope of the GL 89-10 program.
Byron and Braidwood In 1998, the licensee for Byron and Braidwood requested an amendment to its TS to take credit for automatic operation of the PORVs to mitigate an IOECCS event. 113 In the amendment request, the licensee stated that the PSVs had not been qualified to reseat after passing subcooled liquid. The licensee stated that the PORVs at Byron and Braidwood are safety-related components with safety-related actuators and accumulator tanks, with PORV control circuits classified as safety-related. The licensee noted that some portions of the PORV circuitry are nonsafety-related, with improvements implemented in response to GL 90-06. 114 The licensee stated that the PORV block valves are within the scope of the GL 89-10 program.
In 1999, the NRG staff requested additional information related to concerns that the PORV circuitry did not meet the single failure criterion. 115 The licensee reevaluated its approach and withdrew its TS amendment request. 116 No further action regarding this amendment request was identified by the Panel. However, in a public meeting during the review of the NRR Appeal,117 the licensee stated that the PORVs and their block valves at Byron and Braidwood are safety-related with the exception of one circuitry aspect of the PORV. 118 In 2001, the NRC issued a license amendment for Byron and Braidwood to increase the maximum thermal power for each unit from 3411 megawatts thermal (MWt) to 3586.6 MWt (commonly referred !Q_as a stretch power uprate). 119 During its review, the NRG staff requested that the licensee address water solid conditions in the pressurizer, because Uthe NRG staff had generally not accepted a solid pressurizer for an IOECCS event given the potential for all three PSVs to be stuck open due to liquid relief through these safety valves. In response, the licensee stated that Section 15.5.1 , "Inadvertent Operation of Emergency Core Cooling System During Power Operation," of the UFSAR had been revised to credit the PSVs to pass water. 120 The licensee discussed the EPRI testing program in response to TMI Action Plan Item 11.D.1, with 110  NRC 2000 111  Union Electric 2000 112 NRC 1987b 11 3 ComEd 1998 114 NRC 1990b 115 NRC 1999 11e ComEd 1999 m Exelon 2015 118 NRC 2016a 119 NRC 2001 b 120 ComEd 2000b
In 1999, the NRG staff requested additional information related to concerns that the PORV circuitry did not meet the single failure criterion. 115 The licensee reevaluated its approach and withdrew its TS amendment request. 116 No further action regarding this amendment request was identified by the Panel. However, in a public meeting during the review of the NRR Appeal,117 the licensee stated that the PORVs and their block valves at Byron and Braidwood are safety-related with the exception of one circuitry aspect of the PORV. 118 In 2001, the NRC issued a license amendment for Byron and Braidwood to increase the maximum thermal power for each unit from 3411 megawatts thermal (MWt) to 3586.6 MWt (commonly referred !Q_as a stretch power uprate). 119 During its review, the NRG staff requested that the licensee address water solid conditions in the pressurizer, because Uthe NRG staff had generally not accepted a solid pressurizer for an IOECCS event given the potential for all three PSVs to be stuck open due to liquid relief through these safety valves. In response, the licensee stated that Section 15.5.1 , "Inadvertent Operation of Emergency Core Cooling System During Power Operation," of the UFSAR had been revised to credit the PSVs to pass water. 120 The licensee discussed the EPRI testing program in response to TMI Action Plan Item 11.D.1, with 110  NRC 2000 111  Union Electric 2000 112 NRC 1987b 11 3 ComEd 1998 114 NRC 1990b 115 NRC 1999 11e ComEd 1999 m Exelon 2015 118 NRC 2016a 119 NRC 2001 b 120 ComEd 2000b the results summarized in EPRI NP-2628-SR. 121 The licensee referenced previous NRC approvals related to TMI Action Plan Item 11.0.1 .122 The NRC staff made a further request regarding the temperature of water that would be discharged by the PSVs and the length of time that the PSVs would be expected to discharge water. The NRC staff also asked the licensee to discuss which EPRI tests are applicable to the Byron and Braidwood condition. In response, the licensee stated that the PSVs would close after discharging water, although they may not be leaktight. 123 The licensee stated that the leakage from up to three leaking PSVs is bounded by one fully open PSV. The licensee indicated that the EPRI testing of the Crosby safety valves in EPRI NP-2770-LD, Volumes 1 and 6,124 are applicable. The licensee indicated that valve chatter occurred during the tests with damage to the internals, but that the safety valve closed in response to system depressurization. The licensee stated that the Byron and Braidwood pressurizer water temperature of 590 &deg;Fis higher than the EPRI tests (530 &deg;F). The licensee stated that the assumed length of the event is 20 minutes from initial SI signal to when the system pressure is restored below PSV lift setpoint.
 
the results summarized in EPRI NP-2628-SR. 121 The licensee referenced previous NRC approvals related to TMI Action Plan Item 11.0.1 .122 The NRC staff made a further request regarding the temperature of water that would be discharged by the PSVs and the length of time that the PSVs would be expected to discharge water. The NRC staff also asked the licensee to discuss which EPRI tests are applicable to the Byron and Braidwood condition. In response, the licensee stated that the PSVs would close after discharging water, although they may not be leaktight. 123 The licensee stated that the leakage from up to three leaking PSVs is bounded by one fully open PSV. The licensee indicated that the EPRI testing of the Crosby safety valves in EPRI NP-2770-LD, Volumes 1 and 6,124 are applicable. The licensee indicated that valve chatter occurred during the tests with damage to the internals, but that the safety valve closed in response to system depressurization. The licensee stated that the Byron and Braidwood pressurizer water temperature of 590 &deg;Fis higher than the EPRI tests (530 &deg;F). The licensee stated that the assumed length of the event is 20 minutes from initial SI signal to when the system pressure is restored below PSV lift setpoint.
In Section 3.2 of the SE accompanying the license amendment, the NRC staff discussed its review of the performance of the PORVs and PSVs to discharge liquid water for approximately 20 minutes. The NRC staff discussed the EPRI testing program, with the conclusion that the PSV would close in response to system depressurization. The NRC staff reviewed the licensee's evaluation of the performance of the PSVs for liquid water conditions. The NRC staff found that the EPRI tests adequately demonstrated the performance of the valves for the expected water temperature conditions, and that there was reasonable assurance that the valves would adequately reseat following the spurious SI event. The NRC staff determined that EPRI test data indicated that the PSVs might chatter for the expected fluid inlet temperature, but that the resulting PSV seat leakage following the water discharge would be less than the discharge from one stuck-open PSV. Therefore, the NRC staff found the licensee's crediting of the PSVs to discharge liquid water during the spurious SI event to be acceptable. This portion of the SE was based on input provided by the Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch, with techn ical input from the responsible staff member for safety valves in the NRR Division of Engineering.12s As noted by the licensee, Section 15.5.1 of the Byron and Braidwood UFSAR at the time of the stretch power uprate includes PSV water discharge and references the TMI Action Plan Item 11.D.1 approvals. 126 The current UFSAR Revision 15 concludes that the IOECCS event does not progress into a stuck-open PSV LOCA event. 127 The UFSAR states that all three PSVs may lift but will reclose, and that the leakage is bounded by one fully open valve with the consequences bounded by the IOPSRV event. The UFSAR also specifies that if SI results in discharge of coolant through the pressurizer valves, the operators will bring the plant to cold shutdown to inspect the valves.
In Section 3.2 of the SE accompanying the license amendment, the NRC staff discussed its review of the performance of the PORVs and PSVs to discharge liquid water for approximately 20 minutes. The NRC staff discussed the EPRI testing program, with the conclusion that the PSV would close in response to system depressurization. The NRC staff reviewed the licensee's evaluation of the performance of the PSVs for liquid water conditions. The NRC staff found that the EPRI tests adequately demonstrated the performance of the valves for the expected water temperature conditions, and that there was reasonable assurance that the valves would adequately reseat following the spurious SI event. The NRC staff determined that EPRI test data indicated that the PSVs might chatter for the expected fluid inlet temperature, but that the resulting PSV seat leakage following the water discharge would be less than the discharge from one stuck-open PSV. Therefore, the NRC staff found the licensee's crediting of the PSVs to discharge liquid water during the spurious SI event to be acceptable. This portion of the SE was based on input provided by the Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch, with techn ical input from the responsible staff member for safety valves in the NRR Division of Engineering.12s As noted by the licensee, Section 15.5.1 of the Byron and Braidwood UFSAR at the time of the stretch power uprate includes PSV water discharge and references the TMI Action Plan Item 11.D.1 approvals. 126 The current UFSAR Revision 15 concludes that the IOECCS event does not progress into a stuck-open PSV LOCA event. 127 The UFSAR states that all three PSVs may lift but will reclose, and that the leakage is bounded by one fully open valve with the consequences bounded by the IOPSRV event. The UFSAR also specifies that if SI results in discharge of coolant through the pressurizer valves, the operators will bring the plant to cold shutdown to inspect the valves.
12 1 EPRI 1982b 122  NRC 1998c and NRC 1990a 12 3 Exelon 2001 124 EPRI 1982c and EPRI 1983 125 NRC 2001 a 126 Exelon 2002 127 Exelon 2014
12 1 EPRI 1982b 122  NRC 1998c and NRC 1990a 12 3 Exelon 2001 124 EPRI 1982c and EPRI 1983 125 NRC 2001 a 126 Exelon 2002 127 Exelon 2014 In 2004, the NRC issued a license amendment for Byron and Braidwood granting an adjustment to the PSV setpoints. 128 As documented in the SE, the NRC staff requested during its review that the licensee perform a quantitative analysis regarding PSV water cycles and discharge water temperature. For the loss of ac power (LOAC) with reactor coolant pump (RCP) seal injection event, the licensee's analysis indicated that continued injection of water into the RCS through the RCP seals would result in a water-solid pressurizer and water discharge through the PSVs. The proposed PSV setpoint tolerance assuming negative tolerance would result in a lower PSV lift setpoint. With the lower setpoint. the PSV would open earlier, and a larger number of PSV water cycles with a lower water discharge temperature could result during the transient. The licensee performed an analysis of the LOAC with RCP seal injection event. and determined the revised PSV setpoint would result in an increase of about one PSV water cycle and a reduction in the water discharge temperature of about 0.5 &deg;F. A comparison of the reanalysis showed that the spurious SI event remained the limiting event since it resulted in a greater increase in the number of PSV water cycles (two cycles vs. one cycle) and a greater decrease in the PSV discharge water temperature (3.0 &deg;F vs. 0.5 &deg;F) than that calculated for the LOAC with RCP seal injection event. The water discharge temperature in the analysis of record for the spurious SI event was 590 &deg;F. The lowest discharge water temperature for the spurious SI event with the revised PSV setpoint was 587 &deg;F. The NRC staff found that the calculated water discharge temperature (587 &deg;F} was significantly higher than the discharge water temperature of 530 &deg;F that was used to support operability of the PSVs as discussed in the analysis of record. As a result, the NRC staff concluded that the analysis was acceptable to assure that the PSVs will remain operable following a spurious SI event.
 
In 2004, the NRC issued a license amendment for Byron and Braidwood granting an adjustment to the PSV setpoints. 128 As documented in the SE, the NRC staff requested during its review that the licensee perform a quantitative analysis regarding PSV water cycles and discharge water temperature. For the loss of ac power (LOAC) with reactor coolant pump (RCP) seal injection event, the licensee's analysis indicated that continued injection of water into the RCS through the RCP seals would result in a water-solid pressurizer and water discharge through the PSVs. The proposed PSV setpoint tolerance assuming negative tolerance would result in a lower PSV lift setpoint. With the lower setpoint. the PSV would open earlier, and a larger number of PSV water cycles with a lower water discharge temperature could result during the transient. The licensee performed an analysis of the LOAC with RCP seal injection event. and determined the revised PSV setpoint would result in an increase of about one PSV water cycle and a reduction in the water discharge temperature of about 0.5 &deg;F. A comparison of the reanalysis showed that the spurious SI event remained the limiting event since it resulted in a greater increase in the number of PSV water cycles (two cycles vs. one cycle) and a greater decrease in the PSV discharge water temperature (3.0 &deg;F vs. 0.5 &deg;F) than that calculated for the LOAC with RCP seal injection event. The water discharge temperature in the analysis of record for the spurious SI event was 590 &deg;F. The lowest discharge water temperature for the spurious SI event with the revised PSV setpoint was 587 &deg;F. The NRC staff found that the calculated water discharge temperature (587 &deg;F} was significantly higher than the discharge water temperature of 530 &deg;F that was used to support operability of the PSVs as discussed in the analysis of record. As a result, the NRC staff concluded that the analysis was acceptable to assure that the PSVs will remain operable following a spurious SI event.
In 2014, the NRC issued a license amendment for Byron and Braidwood granting a measurement uncertainty recapture (MUR} power uprate.129 The NRC staff determined that the IOECCS event was outside of the scope of the MUR power uprate, because the licensee did not propose to modify the Chapter 15 analyses related to PSV and PORV water discharge.
In 2014, the NRC issued a license amendment for Byron and Braidwood granting a measurement uncertainty recapture (MUR} power uprate.129 The NRC staff determined that the IOECCS event was outside of the scope of the MUR power uprate, because the licensee did not propose to modify the Chapter 15 analyses related to PSV and PORV water discharge.
With respect to inservice testing (1ST) activities, the Byron 1ST program 130 references the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code}, 2004 Edition through 2006 Addenda; and the Braidwood 1ST program 131 references the ASME OM Code, 2001 Edition through 2003 Addenda. The Byron 1ST Program specifies the following testing and intervals for the PORVs, PORV block valves, and PSVs:
With respect to inservice testing (1ST) activities, the Byron 1ST program 130 references the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code}, 2004 Edition through 2006 Addenda; and the Braidwood 1ST program 131 references the ASME OM Code, 2001 Edition through 2003 Addenda. The Byron 1ST Program specifies the following testing and intervals for the PORVs, PORV block valves, and PSVs:
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The NRC staff specifically evaluated whether the PSVs could reasonably be expected to reseat to prevent the spurious SI actuation (an ANS Condition II event) from causing a stuck-open PSV (an ANS Condition Ill event). This issue was said to be further discussed in RIS 2005-29. While the PSVs for Beaver Valley were qualified to discharge steam, if the valves discharged water with sufficient subcooling, the NRC staff was concerned that they might not reseat properly.
The NRC staff specifically evaluated whether the PSVs could reasonably be expected to reseat to prevent the spurious SI actuation (an ANS Condition II event) from causing a stuck-open PSV (an ANS Condition Ill event). This issue was said to be further discussed in RIS 2005-29. While the PSVs for Beaver Valley were qualified to discharge steam, if the valves discharged water with sufficient subcooling, the NRC staff was concerned that they might not reseat properly.
Based the licensee's analysis, during a spurious SI event, the PSVs would be required to discharge steam followed by high temperature water after the pressurizer filled. The licensee provided plots of the pressuri.zer water temperatures for this event that indicated that the 132 NRC 2001 d 133 This term is used in the Shearon Harris SE. The licensee's RAI response (CP&L 2000) makes clear that the referenced SRVs and PORVs are pressurizer valves.
Based the licensee's analysis, during a spurious SI event, the PSVs would be required to discharge steam followed by high temperature water after the pressurizer filled. The licensee provided plots of the pressuri.zer water temperatures for this event that indicated that the 132 NRC 2001 d 133 This term is used in the Shearon Harris SE. The licensee's RAI response (CP&L 2000) makes clear that the referenced SRVs and PORVs are pressurizer valves.
134 NRC 2006
134 NRC 2006 minimum temperature of the discharged liquid for Beaver Valley was approximately 620 &deg;F. To evaluate the capability of the valves to discharge and reseat, the NRC staff reviewed the available data from the full-flow tests performed during the EPRI test program in 1981 for the specific PSV models representative of those installed at Beaver Valley. The licensee also used the methodology contained in WCAP-11677 and determined that the minimum acceptable liquid temperature for which the PSVs were expected to successfully discharge and reseat was less than the minimum expected temperature for the spurious SI event for Beaver Valley.
 
minimum temperature of the discharged liquid for Beaver Valley was approximately 620 &deg;F. To evaluate the capability of the valves to discharge and reseat, the NRC staff reviewed the available data from the full-flow tests performed during the EPRI test program in 1981 for the specific PSV models representative of those installed at Beaver Valley. The licensee also used the methodology contained in WCAP-11677 and determined that the minimum acceptable liquid temperature for which the PSVs were expected to successfully discharge and reseat was less than the minimum expected temperature for the spurious SI event for Beaver Valley.
The NRC staff agreed that both the minimum expected water discharge temperature and the minimum acceptable water temperature had been conservatively calculated. Therefore, the NRC staff determined that, for purposes of preventing the occurrence of a more serious ANS Condition Ill event, there was reasonable assurance that the PSVs would discharge water and reseat adequately following a spurious SI actuation. A consideration of the NRC staff in making this finding was that, in the unlikely event of a stuck-open PSV, the ECCS was fully capable of mitigating the resulting LOCA.
The NRC staff agreed that both the minimum expected water discharge temperature and the minimum acceptable water temperature had been conservatively calculated. Therefore, the NRC staff determined that, for purposes of preventing the occurrence of a more serious ANS Condition Ill event, there was reasonable assurance that the PSVs would discharge water and reseat adequately following a spurious SI actuation. A consideration of the NRC staff in making this finding was that, in the unlikely event of a stuck-open PSV, the ECCS was fully capable of mitigating the resulting LOCA.
Turkey Point In 2012, the NRC issued a license amendment authorizing an EPU for Turkey Point Nuclear Generating, Units 3 and 4 (Turkey Point), increasing the thermal power level of each unit approximately 15 percent to 2644 MWt. 135 In the SE accompanying the amendment, the NRC staff indicated that ECCS actuation was not a possible Initiator of inadvertent increase in reactor coolant inventory because the high head SI pumps have a shut-off head below the normal RCS operating pressure. The NRC staff stated that a eves malfunction that increases RCS inventory was evaluated for the effects of adding water inventory to the RCS. If the pressurizer filled and caused water to be relieved through the PORVs or PSVs, then these valves could stick open and create a small break LOCA. The NRC staff stated that this would violate the acceptance criterion that prohibits the escalation of an anticipated operational occurrence (AOO) into a more serious event. Satisfaction of this acceptance criterion was demonstrated by showing that sufficient time would exist for the operator to recognize the situation and end the charging flow before the pressurizer could fill.
Turkey Point In 2012, the NRC issued a license amendment authorizing an EPU for Turkey Point Nuclear Generating, Units 3 and 4 (Turkey Point), increasing the thermal power level of each unit approximately 15 percent to 2644 MWt. 135 In the SE accompanying the amendment, the NRC staff indicated that ECCS actuation was not a possible Initiator of inadvertent increase in reactor coolant inventory because the high head SI pumps have a shut-off head below the normal RCS operating pressure. The NRC staff stated that a eves malfunction that increases RCS inventory was evaluated for the effects of adding water inventory to the RCS. If the pressurizer filled and caused water to be relieved through the PORVs or PSVs, then these valves could stick open and create a small break LOCA. The NRC staff stated that this would violate the acceptance criterion that prohibits the escalation of an anticipated operational occurrence (AOO) into a more serious event. Satisfaction of this acceptance criterion was demonstrated by showing that sufficient time would exist for the operator to recognize the situation and end the charging flow before the pressurizer could fill.
The NRC staff concluded that the licensee's analyses of IOECCS and eves events adequately accounted for operation of the plant at the proposed power level.
The NRC staff concluded that the licensee's analyses of IOECCS and eves events adequately accounted for operation of the plant at the proposed power level.
Regarding an inadvertent opening of a PORV, the licensee initially proposed that the consequences of this event were bounded by the small break LOCA. The NRG staff did not accept this proposed disposition. If action were not taken to secure the open valve by either closing the PORV or its block valve, the NRC staff stated that this event could escalate to a small-break LOCA, which would be contrary to the non-escalation position. When the pressurizer filled, water would begin to flow through the open PORV. If the PORV were not qualified for water discharge, the NRC staff stated that it was likely the PORV would not close upon demand. In this way, the NRC staff stated that the inadvertent opening of a PORV, an AOO, would become a small break-LOCA at the top of the pressurizer, an ANS Condition Ill event. The NRC staff requested that the licensee address the inadvertent opening of the PORV with respect to the third criterion for.an ANS Condition II event.
Regarding an inadvertent opening of a PORV, the licensee initially proposed that the consequences of this event were bounded by the small break LOCA. The NRG staff did not accept this proposed disposition. If action were not taken to secure the open valve by either closing the PORV or its block valve, the NRC staff stated that this event could escalate to a small-break LOCA, which would be contrary to the non-escalation position. When the pressurizer filled, water would begin to flow through the open PORV. If the PORV were not qualified for water discharge, the NRC staff stated that it was likely the PORV would not close upon demand. In this way, the NRC staff stated that the inadvertent opening of a PORV, an AOO, would become a small break-LOCA at the top of the pressurizer, an ANS Condition Ill event. The NRC staff requested that the licensee address the inadvertent opening of the PORV with respect to the third criterion for.an ANS Condition II event.
The licensee provided an analysis performed largely in accordance with NRG-approved, Westinghouse analytic methodology using the RETRAN computer code; however, this analysis 135 NRG 2012a
The licensee provided an analysis performed largely in accordance with NRG-approved, Westinghouse analytic methodology using the RETRAN computer code; however, this analysis 135 NRG 2012a was performed assuming that the PORV opened instead of the PSV. The NRC staff stated that assuming the opening of the PORV is acceptable, because the PSV is differently qualified, and reseats mechanically. An additional independent fault would be required to cause the PSV to fail to close. The analysis indicated that the pressurizer would fill within about 240 seconds. The licensee stated that there were multiple alarms to indicate the opening of a PORV. The licensee stated that a prompt operator action would be needed to close the PORV and, if the PORV does not close, the operator would be directed to close the block valve. Because the necessary actions would be prompt and simple, the NRC staff agreed that there would be sufficient time to secure the inadvertently open PORV without filling the pressurizer.
 
was performed assuming that the PORV opened instead of the PSV. The NRC staff stated that assuming the opening of the PORV is acceptable, because the PSV is differently qualified, and reseats mechanically. An additional independent fault would be required to cause the PSV to fail to close. The analysis indicated that the pressurizer would fill within about 240 seconds. The licensee stated that there were multiple alarms to indicate the opening of a PORV. The licensee stated that a prompt operator action would be needed to close the PORV and, if the PORV does not close, the operator would be directed to close the block valve. Because the necessary actions would be prompt and simple, the NRC staff agreed that there would be sufficient time to secure the inadvertently open PORV without filling the pressurizer.
St. Lucie In 2012, the NRC issued a license amendment authorizing an EPU for St. Lucie Plant, Unit 2 (St. Lucie, Unit 2) that increased the authorized thermal power level about 12 percent to 3020 MWt.
St. Lucie In 2012, the NRC issued a license amendment authorizing an EPU for St. Lucie Plant, Unit 2 (St. Lucie, Unit 2) that increased the authorized thermal power level about 12 percent to 3020 MWt.
Regarding an IOECCS event, the high pressure SI pumps would be incapable during power operations of delivering flow to the RCS because the pumps' shut-off head would be less than the normal RCS operating pressure of 2250 pounds per square inch absolute. Therefore, the licensee determined that the inadvertent operation of the ECCS at power event was not a credible event and did not analyze it for the proposed EPU. The NRC staff found that the licensee's position for not analyzing the IOECCS event to be acceptable.
Regarding an IOECCS event, the high pressure SI pumps would be incapable during power operations of delivering flow to the RCS because the pumps' shut-off head would be less than the normal RCS operating pressure of 2250 pounds per square inch absolute. Therefore, the licensee determined that the inadvertent operation of the ECCS at power event was not a credible event and did not analyze it for the proposed EPU. The NRC staff found that the licensee's position for not analyzing the IOECCS event to be acceptable.
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If the PORV or its block valve was not closed, the NRC staff stated that the IOPORV event would enter the second pha:se with actuation of the ECCS. Based on its review, the NRC staff determined that the pressurizer overfill analysis, available alarming system, and procedures, in combination with simulator exercise results, provided reasonable assurance that the pressurizer would not be expected to fill to a water solid condition that could prevent the PORV or PSV from closing after they were open. The NRC staff therefore concluded that the event would not generate a more serious plant condition, meeting the non-escalation criterion. The NRC staff stated that it reviewed the licensee's analyses of the inadvertent opening of a pressurizer PORV event, and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models.
If the PORV or its block valve was not closed, the NRC staff stated that the IOPORV event would enter the second pha:se with actuation of the ECCS. Based on its review, the NRC staff determined that the pressurizer overfill analysis, available alarming system, and procedures, in combination with simulator exercise results, provided reasonable assurance that the pressurizer would not be expected to fill to a water solid condition that could prevent the PORV or PSV from closing after they were open. The NRC staff therefore concluded that the event would not generate a more serious plant condition, meeting the non-escalation criterion. The NRC staff stated that it reviewed the licensee's analyses of the inadvertent opening of a pressurizer PORV event, and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models.
The NRC staff concluded that the licensee demonstrated that all AOO acceptance criteria were satisfactorily met.
The NRC staff concluded that the licensee demonstrated that all AOO acceptance criteria were satisfactorily met.
North Anna North Anna Power Station, Units 1 and 2 (North Anna) UFSAR Section 15.2.14, "Spurious Operation of the Safety Injection System at Power," describes plant response to an inadvertent SI event. In particular, UFSAR Section 15.2.14.2.3, "Event Propagation," states the following:
North Anna North Anna Power Station, Units 1 and 2 (North Anna) UFSAR Section 15.2.14, "Spurious Operation of the Safety Injection System at Power," describes plant response to an inadvertent SI event. In particular, UFSAR Section 15.2.14.2.3, "Event Propagation," states the following:
Safety valve (Reference 18) and PORV (Reference 19) testing has revealed no instances of failure of the valves to reseat following water relief. Resulting leakage is within the capacity of the normal makeup system and is therefore not considered to be a small break loss of reactor coolant event. Therefore, the complete filling of the pressurizer and/or water relief via a safety valve as a result of a spurious safety injection does not constitute a failure to meet the event propagation acceptance criterion. Although primary credit for preventing the propagation of the event to a small break loss of reactor coolant event is the reseating of the PORVs and safety valves, It is noted that the PORVs (which open prior to the safety valves and, if open, preclude safety valve actuation for this event) are provided with block valves which the operator will close in the event of excessive PORV leakage.
Safety valve (Reference 18) and PORV (Reference 19) testing has revealed no instances of failure of the valves to reseat following water relief. Resulting leakage is within the capacity of the normal makeup system and is therefore not considered to be a small break loss of reactor coolant event. Therefore, the complete filling of the pressurizer and/or water relief via a safety valve as a result of a spurious safety injection does not constitute a failure to meet the event propagation acceptance criterion. Although primary credit for preventing the propagation of the event to a small break loss of reactor coolant event is the reseating of the PORVs and safety valves, It is noted that the PORVs (which open prior to the safety valves and, if open, preclude safety valve actuation for this event) are provided with block valves which the operator will close in the event of excessive PORV leakage.
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Conclusion In conclusion, the reliance by the licensee for Byron and Braidwood on the acceptable performance of the PSVs and PORVs following water discharge in response to abnormal events is not inconsistent with similar approaches by some other nuclear power plant licensees. In general, the review of activities by various nuclear power plant licensees related to PSV and PORV performance revealed reliance on EPRI, Wyle, and valve vendor testing to provide support for the performance of these valves under various service conditions. Specific certification for flow capacity of these valves for water discharge in accordance with the ASME BPV Code and National Board was not identified in the review of various justifications prepared by nuclear power plant licensees.
Conclusion In conclusion, the reliance by the licensee for Byron and Braidwood on the acceptable performance of the PSVs and PORVs following water discharge in response to abnormal events is not inconsistent with similar approaches by some other nuclear power plant licensees. In general, the review of activities by various nuclear power plant licensees related to PSV and PORV performance revealed reliance on EPRI, Wyle, and valve vendor testing to provide support for the performance of these valves under various service conditions. Specific certification for flow capacity of these valves for water discharge in accordance with the ASME BPV Code and National Board was not identified in the review of various justifications prepared by nuclear power plant licensees.
In evaluating the historical documents for Byron and Braidwood, the Panel found it challenging to determine specifically how the licensee resolved the concern raised in NSAL-93-013 in its analyses and plant operations. While the record does not currently support a compliance backfit in this case, if (as recommended by the Panel) the NRC staff undertakes a generic review of licensees' treatment of the potential for pressurizer valve damage following water discharge, it may be appropriate to consider what actions have been taken, how operating experience with water discharge has been considered, and how analysis assumptions are considered in operational practices (including inservice testing) at each plant.
In evaluating the historical documents for Byron and Braidwood, the Panel found it challenging to determine specifically how the licensee resolved the concern raised in NSAL-93-013 in its analyses and plant operations. While the record does not currently support a compliance backfit in this case, if (as recommended by the Panel) the NRC staff undertakes a generic review of licensees' treatment of the potential for pressurizer valve damage following water discharge, it may be appropriate to consider what actions have been taken, how operating experience with water discharge has been considered, and how analysis assumptions are considered in operational practices (including inservice testing) at each plant.
APPENDIX D: REFERENCES
APPENDIX D: REFERENCES
: 1. AEC 1970: Atomic Energy Commission (AEC), "Backfitting of Production and Utilization Facilities; Construction Permits and Operating Licenses," Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, published March 31 , 1970 (as amended).
: 1. AEC 1970: Atomic Energy Commission (AEC), "Backfitting of Production and Utilization Facilities; Construction Permits and Operating Licenses," Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, published March 31 , 1970 (as amended).
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: 9. ComEd 2000a: Commonwealth Edison Company, letter from RM. Krich to U.S. NRC, "Request for a License Amendment to Permit Uprated Power Operations at Byron and Braidwood Stations," dated July 5, 2000. ADAMS Accession No. ML003730608.
: 9. ComEd 2000a: Commonwealth Edison Company, letter from RM. Krich to U.S. NRC, "Request for a License Amendment to Permit Uprated Power Operations at Byron and Braidwood Stations," dated July 5, 2000. ADAMS Accession No. ML003730608.
: 10. ComEd 2000b: Commonwealth Edison Company, letter from RM. Krich to U.S. NRG, "Response to Request for Additional Information Regarding the License Amendment Request to Permit Uprated Power Operations at Byron and Braidwood Stations," dated November 27, 2000. ADAMS Accession No. ML003772461.
: 10. ComEd 2000b: Commonwealth Edison Company, letter from RM. Krich to U.S. NRG, "Response to Request for Additional Information Regarding the License Amendment Request to Permit Uprated Power Operations at Byron and Braidwood Stations," dated November 27, 2000. ADAMS Accession No. ML003772461.
11 . Consumers 1982: Consumers Energy Company, letter from D. P. Hoffman to Harold R. Denton, U.S. NRC, "Safety and Relief Valve Test Report for the EPRI PWR Safety and Relief Valve Test Program," dated April 1, 1982. (This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8207160337 and microfiche location 13866:001 - 13869: 106.]
11 . Consumers 1982: Consumers Energy Company, letter from D. P. Hoffman to Harold R. Denton, U.S. NRC, "Safety and Relief Valve Test Report for the EPRI PWR Safety and Relief Valve Test Program," dated April 1, 1982. (This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8207160337 and microfiche location 13866:001 - 13869: 106.]
: 12. CP&L 2000: Carolina Power & light Company, letter from James Scarola to U.S. NRC, "License Amendhlent Application Power Uprate," dated December 14, 2000. ADAMS Accession No. ML003780761.
: 12. CP&L 2000: Carolina Power & light Company, letter from James Scarola to U.S. NRC, "License Amendhlent Application Power Uprate," dated December 14, 2000. ADAMS Accession No. ML003780761.
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(TAC Nos. M77332, M77333, M77334, M77335, M77402, M77403, M77404, and M77405)," dated November 18, 1991. ADAMS Accession No. ML020860105.
(TAC Nos. M77332, M77333, M77334, M77335, M77402, M77403, M77404, and M77405)," dated November 18, 1991. ADAMS Accession No. ML020860105.
: 56. NRC 1994a: U.S. NRC, SECY-94-084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs," dated March 28, 1994. ADAMS Accession No. ML003708068.
: 56. NRC 1994a: U.S. NRC, SECY-94-084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs," dated March 28, 1994. ADAMS Accession No. ML003708068.
: 57. NRC 1994b: U.S. NRC, SRM-SECY-94-084, "SECY-94-084 - Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive
: 57. NRC 1994b: U.S. NRC, SRM-SECY-94-084, "SECY-94-084 - Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs and COMSECY-94-024 - Implementation of Design Certification and Light-Water Reactor Design Issues," dated June 30, 1994. ADAMS Accession No. ML003708098.
 
Plant Designs and COMSECY-94-024 - Implementation of Design Certification and Light-Water Reactor Design Issues," dated June 30, 1994. ADAMS Accession No. ML003708098.
: 58. NRC 1996: U.S. NRC, NUREG-0800, SRP Section 15.5.1 - 15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," draft Revision 2, dated April 1996. ADAMS Accession No. ML052070725.
: 58. NRC 1996: U.S. NRC, NUREG-0800, SRP Section 15.5.1 - 15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," draft Revision 2, dated April 1996. ADAMS Accession No. ML052070725.
: 59. NRC 1997: U.S. NRC, letter from Leonard N. Olshan, NRC, to Leon R. Eliason, Public Service Electric & Gas Company, "Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. M97827 and M97828}," dated June 4, 1997. ADAMS Accession No. ML011720397.
: 59. NRC 1997: U.S. NRC, letter from Leonard N. Olshan, NRC, to Leon R. Eliason, Public Service Electric & Gas Company, "Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. M97827 and M97828}," dated June 4, 1997. ADAMS Accession No. ML011720397.
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: 77. NRC 2012a: U.S. NRC, letter from Jason C. Paige to Mano Nazar, Florida Power and Light Company, "Turkey Point Units 3 and 4- Issuance of Amendments Regarding Extended Power Uprate (TAC Nos. ME4907 and ME4908)," dated June 15, 2012.
: 77. NRC 2012a: U.S. NRC, letter from Jason C. Paige to Mano Nazar, Florida Power and Light Company, "Turkey Point Units 3 and 4- Issuance of Amendments Regarding Extended Power Uprate (TAC Nos. ME4907 and ME4908)," dated June 15, 2012.
ADAMS Accession No. ML11293A359.
ADAMS Accession No. ML11293A359.
: 78. NRC 2012b: U.S. NRC, letter from Tracy J. Orf to Mano Nazar, Florida Power and Light Company, "St. Lucie Plant, Unit 2 - Issuance of Amendment Regarding Extended Power
: 78. NRC 2012b: U.S. NRC, letter from Tracy J. Orf to Mano Nazar, Florida Power and Light Company, "St. Lucie Plant, Unit 2 - Issuance of Amendment Regarding Extended Power Uprate (TAC No. ME5843)," dated September 24, 2012. ADAMS Accession No. ML12268A132.
 
Uprate (TAC No. ME5843)," dated September 24, 2012. ADAMS Accession No. ML12268A132.
: 79. NRC 2013: U.S. NRC, Management Directive 8.4, "Management of Facility-Specific Backfitting and Information Collection," dated October 9, 2013. ADAMS Accession No. ML12059A460.
: 79. NRC 2013: U.S. NRC, Management Directive 8.4, "Management of Facility-Specific Backfitting and Information Collection," dated October 9, 2013. ADAMS Accession No. ML12059A460.
: 80. NRC 2014a: U.S. NRC, letter from Joel S. Wiebe to Michael J. Pacilio, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. MF2418, MF2419, MF2420, and MF2421)," dated February 7, 2014. ADAMS Accession No. ML 13281AOOO.
: 80. NRC 2014a: U.S. NRC, letter from Joel S. Wiebe to Michael J. Pacilio, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. MF2418, MF2419, MF2420, and MF2421)," dated February 7, 2014. ADAMS Accession No. ML 13281AOOO.
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102. WOG 1982: Westinghouse Owners Group (WOG), letter from Oliver D. Kingsley, Alabama Power Company, to Harold R. Denton, U.S. NRC, "NUREG-0737, Item 11.0.1,
102. WOG 1982: Westinghouse Owners Group (WOG), letter from Oliver D. Kingsley, Alabama Power Company, to Harold R. Denton, U.S. NRC, "NUREG-0737, Item 11.0.1,
     'Pressurizer Safety Valve Operability,'" dated July 27, 1982. Forwards Westinghouse WCAP-10105, "Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program," dated June 1982. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8208190307 and microfiche location 14387:189-301 .]
     'Pressurizer Safety Valve Operability,'" dated July 27, 1982. Forwards Westinghouse WCAP-10105, "Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program," dated June 1982. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8208190307 and microfiche location 14387:189-301 .]
 
~PPENDIX lcr3a1E: LIST OF ABBREVIATIONS From:                          Spencer, Michael Sent:                          Monday, July 18, 2016 9:14 AM To:                            Holahan, Gary; West, Steven; Scarbrough, Thomas; Clark, Theresa
~PPENDIX lcr3a1E: LIST OF ABBREVIATIONS
 
From:                          Spencer, Michael Sent:                          Monday, July 18, 2016 9:14 AM To:                            Holahan, Gary; West, Steven; Scarbrough, Thomas; Clark, Theresa


==Subject:==
==Subject:==
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1
1


the Back.fit Rule (1OCFR50.109), related court actions, and Commission and staff guidance on application of the Backfit Rule; licensing activities on Byron and Braidwood (License Amendment Requests, License Amendments, staff Requests for Additional Information, licensee responses, meeting summaries, staff Safety Evaluation Reports, and Updates of the Final Safety Analysis Report UFSAR)) aver the period of 1997 to the present; the NRC guidance relevant to the analysis of inadvertent operation of the ECCS (Standard Review Plan (SRP) sections 15.0, 15.5.1, and 15.6.1) over the period of 1981 to the present; the Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013, June 30, 1993) and its supplement (NSAL-93-13 Suppl. 1, October 28, 1994)(Adams XXXXXXXXXXX);
the Back.fit Rule (10CFR50.109), related court actions, and Commission and staff guidance on application of the Backfit Rule; licensing activities on Byron and Braidwood (License Amendment Requests, License Amendments, staff Requests for Additional Information, licensee responses, meeting summaries, staff Safety Evaluation Reports, and Updates of the Final Safety Analysis Report UFSAR)) aver the period of 1997 to the present; the NRC guidance relevant to the analysis of inadvertent operation of the ECCS (Standard Review Plan (SRP) sections 15.0, 15.5.1, and 15.6.1) over the period of 1981 to the present; the Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013, June 30, 1993) and its supplement (NSAL-93-13 Suppl. 1, October 28, 1994)(Adams XXXXXXXXXXX);
actions taken by other licensees in response to Westinghouse NSAL-93-013; the history of NRC and industry activities related to Power Operated Relief Valves (PORV), their Block Valves, and Safety Valves (including Three Mile Island Action Plan (NU REG 0737) items 11.D.1, 11.0.3,11.G.1, 11.K.3, and Generic Letter 89-10 and supplements), and related Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREGXXXX .... ).
actions taken by other licensees in response to Westinghouse NSAL-93-013; the history of NRC and industry activities related to Power Operated Relief Valves (PORV), their Block Valves, and Safety Valves (including Three Mile Island Action Plan (NU REG 0737) items 11.D.1, 11.0.3,11.G.1, 11.K.3, and Generic Letter 89-10 and supplements), and related Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREGXXXX .... ).
ln addition to the document review, the panel had the benefit of meetings with the Office of Nuclear Reactor Regulation, the Office of the General Council, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon and NEI declined offers for a public meeting, but indicated a willingness and interest in providing information if the panel identified the need.
ln addition to the document review, the panel had the benefit of meetings with the Office of Nuclear Reactor Regulation, the Office of the General Council, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon and NEI declined offers for a public meeting, but indicated a willingness and interest in providing information if the panel identified the need.
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ln the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal fact?r in initiating a backfit." Id. at 38102. The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic]
ln the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal fact?r in initiating a backfit." Id. at 38102. The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic]
regulations." Id. at 38103. 3
regulations." Id. at 38103. 3
*************************************************************************************
{In addition to the above, I found the following potentially useful statements from NRC guidance documents on backfitting. I am not sure how much of this will be relevant or where it should be included if relevant, but I am providing it for your information and for potential future use in the document.}
{In addition to the above, I found the following potentially useful statements from NRC guidance documents on backfitting. I am not sure how much of this will be relevant or where it should be included if relevant, but I am providing it for your information and for potential future use in the document.}
NUREG-1409. Backfitting Guidelines, 1990 To be a backfit, "a new or revised staff position or requirement must be involved, that is, there must be a change in content or applicability of the previously applicable regulatory staff position (in the direction of increased safety requirements} .... "
NUREG-1409. Backfitting Guidelines, 1990 To be a backfit, "a new or revised staff position or requirement must be involved, that is, there must be a change in content or applicability of the previously applicable regulatory staff position (in the direction of increased safety requirements} .... "

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Text

From: Clark, Theresa Sent: Thursday, Sept ember 29, 2016 4:46 PM To: McCree, Victor; Johnson, Michael Cc: Inverso, Tara; Bowen, Jeremy

Subject:

messages for 10/12 NSIAC event Attachments: VM M - NSIAC discussion key messages 10-12-16.docx Hi Vic! Attached are some key messages and background to prepare for your remarks at the October 12 NSIAC event.

{Thanks to the offices for providing quick reviews/input.) Please let me know if you'd like more/different information, or if you need a formal speech prepared. Thanks so much !

Theresa Valentine Clark Execut ive Technical Assistant (Reactors)

U.S. Nuclear Regulat ory Commission Theresa .Clark@nrc.gov I 301-415-4048 I 0 -4H10 From: Mccree, Victor Sent: Tuesday, September 20, 2016 7:06 AM To: Dean, Bill <Bill.Dea n@nrc.gov>; Jo hnson, M ichael <Michael.Johnson@nrc.gov>; Ho lian, Brian

<Brian.Holian@nrc.gov>; Uhle, Jennifer <Jennifer.U hle@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

RE: October 12 NSIAC Reception and Dinner Outstanding, thanks Bill! I plan to have Theresa C lark coordinate with your guys, as necessary, to collect the right key messages on the seven topics highlighted below.

Vic From: Dean, Bill Sent: M onday, September 19, 2016 10:21 PM To: Mccree, Vict or <Vict or.M cCree@nrc.gov>; Johnson, M ichael <M ichael.Johnson@nrc.gov>; Holia n, Brian

<Brian.Holian@nrc.gov>; Uhle, Jennifer <Jennifer. Uhle@nrc.gov>

Subject:

Re: October 12 NSIAC Reception and Dinner V ic I do not have an NSIAC meeting on my calendar.

Non-Responsive Record

That's a start.

On: 19 September2016 17:01 , "McCree, Victor" <Victor.McCree@nrc.gov> wrote:

Mike/Bill/Brian/Jennifer, I accepted Tony's invitation below. I'd appreciate your suggestions for any timely topics to include in my after-dinner remarks with the CNOs.

Are you scheduled to meet with the CNOs for this meeting? If so I'd be happy to tee up a few items for you to drive lhome the next day .. . or dive deeper if you're not set to meet with them until much later.

Vic From: PIETRANGELO, Tony [1]

Sent: Wednesday, September 07, 2016 3:58 PM To: McCree, Victor <Victor.McCree@nrc.gov>

Subject:

[External_Sender] October 12 NSIAC Reception and Dinner Hi Vic, I hope all is well. I am writing to inquire about your availability to attend the October 12 NSIAC reception and dinner and to provide some after-dinner remarks. The event is from 6:00 to 9:00 pm in downtown DC near NEl's offices (probably at t he Marriott Metro Center). I hope you will be able to join us.

Thanks.

Best, Tony Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer Nuclear Energy Institute 1201 F Street NW, Suite 1100 Washington, DC 20004 (O} 202-739-8081 (M~(b)(6) I Email: arp@nei.org TAKE THE NE/ FUTURE OF ENERGY QUIZ, www.NEl.org/whynuclear FOLLOW USON 2

7i

  • ic message transmission contains information /ram the Nuclear Energy Institute, Inc. The information is intended solely for the vse of the addressee and its vse b any other person Is no e not the inte nded recipient, yov have received this commvnlcotlon In error, ond any review, use, dis ,s ribution of the contents of this communication is strictly pro , celved this electronic transmission in error, pleas r ,mmedlotely by telephone or by electronic moil and permanently delete the original message. IRS Circvlr" 2 e with requirements imposed by the IRS ond other taxing authorities, we inform you that any tax advice contained in this cu ing any ottachmen s written to be used, and cannot be used, for the purpose of(/) avoiding penalties tho on any taxpayer or (ii) promoting, marketing or recommending to another party a otter addressed herein.

Sent throuch www.intcrmedia.com 3

Victor M . McCree - NSIAC Remarks- October 12, 2016 KEY MESSAGES Non-Responsive Record BACKGROUND Stakeholder Meeting Follow-Up

  • Non-Responsive Record Project AIM Non-Responsive Record 1

Licensing Action Initiatives Non-Responsive Record Byron/Braidwood Backfit Appeal Decision

  • The EDO concluded that the NRC staffs position in the October 2015 backfit issued to Byron and Braidwood related to pressurizer valve performance was a new or modified interpretation of what constitutes compliance and did not provide a basis for a compliance backfit.
  • This decision was communicated on 9/15/16 in publicly available letters to Exelon and NEI, as well as in a memo to the staff tha,t also requested preparation of a plan to reevaluate the generic implications of the technical issue.
  • In a separate activity, CRGR is implementing an EDO tasking to eva luate guidance, training, and knowledge management on backfitting. Multiple industry representatives participated in a 9/13/16 public meeting held by CRGR.

2

Subsequent License Renewal

  • Non-Responsive Record Operator Licensing Non-Responsive Record Digital l&C Action Plan Non-Responsive Record 50.46c Rulemakin~

Non-Responsive Record 3

NUREG-2180 / Very Early Warning Fire Detection Systems Non-Responsive Record Force on Force Efficiencies/Guidance Non-Responsive Record Cybersecurlty at Nuclear Power Plants Non-Responsive Record Advanced Reactor Preparations

  • Non-Responsive Record 4

Ne.ru.........,~....IL!.:~.!.:£!.in:..:.e..~A~t~iv~it~i~e~s~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Non-Responsive Record 5

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From: Clark, Theresa Sent: Tuesday, October 04, 2016 2:20 PM To: Rihm, Roger

Subject:

RE: messages for 10/12 NSIAC event I'm sure he appreciated it! And I hope you don't mind the toe-stepping too much. I do try to avoid it © .

From: Rihm, Roger Sent: 1iuesday, October 04, 2016 2:19 PM To: Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

RE: messages for 10/12 NSIAC event Thanks. So it seems Vic thought he t asked us with different t hings, but there definit ely was some overlap. Oh, well. I guess better to have more than less. Hope the trip has been going well. See you soon!

From: Clark, Theresa Sent: Tuesday, October 04, 2016 12:40 PM To: Rihm, Roger <Roger.Rihm@nrc.gov>

Subject:

Fwd: messages for 10/12 NSIAC event See email chain below and latest copy of background attached.

Original Message --------

From: "Clark, Theresa" <Theresa.Clark@nrc.gov>

Date: Thu, September 29, 2016 10:46 PM +0200 To: "McCree, Victor" <Victor.McCree@ nrc.gov>, "Johnson, Michael" <Michael.Johnson@nrc.gov>

CC: "Inverso, Tara" <Tara.Inverso@nrc.gov>, "Bowen, Jeremy" <Jeremy.Bowen@nrc.gov>

Subject:

messages for 10/ 12 NSIAC event Hi Vic! Attached are some key messages and background to prepare for your remarks at the October 12 NSIAC event.

(Thanks to the offices for providing quick reviews/input.) Please let me know if you'd like more/different information, or if you need a formal speech prepa red. Thanks so much!

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0-4H10 From: McCree, Victor Sent: Tuesday, September 20, 2016 7:06 AM To: Dean, Bill <Bill.Oea n@nrc.gov>; Johnson, Michael <Michael.Johnson@nrc.gov>; Holian, Brian

<Brian.Holian@nrc.gov>; Uhle, Jennifer <Jennifer.Uhle@nrc.gov>; Clark, Theresa <Theresa.Clark@ nrc.gov>

Subject:

RE: October 12 NSIAC Reception and Dinner Outstanding, thanks Bill! I plan to have Theresa Clark coordinate with your guys, as necessary, to collect the right key messages on the seven topics highlighted below.

Vic From: Dean, Bill Sent: Monday, September 19, 2016 10:21 PM To: Mccree, Victor <Victor.McCree@nrc.gov>; Johnson, Michael <Michael.Johnson@nrc.gov>; Holian, Brian

<Brian.Holian@nrc.gov>; Uhle, Jennifer <Jennifer.Uhle@nrc.gov>

Subject:

Re: October 12 NSIAC Reception and Dinner Vic I do not have an NSIAC meetirn.! on mv calendar.

Non-Responsive Record That's a start.

On: 19 September 2016 17:01, "McCree, Victor" <Victor.McCree@nrc.gov> wrote:

Mike/Bill/Brian/Jennifer, I accepted Tony's invitation below. I'd appreciate your suggestions for any timely topics to include in my after-dinner remarks with the CNOs.

Are you scheduled to meet with the CNOs for this meeting? If so I'd be happy to tee up a few items for you to drive home the next day ... or dive deeper if you're not set to meet with them until much later.

Vic From: PIETRANGELO, Tony rg

Sent: Wednesday, September 07, 2016 3:58 PM To: Mccree, Victor <Victor.McCree@nrc.gov>

Subject:

[External_Sender] October 12 NSIAC Reception and Dinner Hi Vic, I hope all is well. I am writing to inquire about your availability to attend the October 12 NSIAC reception and dinner and to provide some after-dinner remarks. The event is from 6:00 to 9:00 pm in downtown DC near NEl' s offices (probably at the Marriott Metro Center). I hope you will be able to join us.

Thanks.

Best, Tony 2

Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer Nuclear Energy Institute 1201 F Street NW, Suite 1100 Washington, DC 20004 (O) 202-739-8081 (M) (b)(6)

Ema, : a._. .aa.r ......~"""'""-=

TAKE THE NE/ FUTUREOF ENERGY au,z, www.NEl.org/whynuclear FOLLOW US ON T ronic message transmission contains Information from the Nuclear Energy Institute, Inc. The Information is intended solely for the use of the addressee and Its any ather person, ou are not the intended recipient, ycu hove received this comm<1nication in error, ond any review II mg or distribution of the contents of this communication Is strict y p ve received this electronic transmission In e sender immediately by telephone or by electronic moil and permanently delete the original message. IRS Cirw or compliance with requirements imposed by the IRS and other taxing avthorities, we Inform you that any tax advice co un cation {Including ony attoc m or written to be used, arid cannot be used. for the purpose of (I) ovoid/ y e imposed on ony taxpayer or (ii) promoting, marketing or recommending to another par matter addressed

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Sent t hrough www.intermedia.com 3

From: Mccree, Victor Sent: Thu rsday, September 29, 2016 5:21 PM To: Clark, Theresa; Johnson, Michael Cc: Inverso, Tara; Bowen, Jeremy

Subject:

Re: messages for 10/12 NSIAC event This is exactly what I need. Thanks Theresa!

Vic

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On: 29 September 2016 22:46, "Clark, Theresa" <Theresa.Clark@nrc.gov> wrote:

Hi Vic! Attached are some key messages and background to prepare for your remarks at the October 12 NSIAC event.

(Thanks to the offices for providing quick reviews/input.} Please let me know if you'd like more/different information, or if you need a formal speech prepared. Thanks so m uch!

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0 -4H10 From: Mccree, Victor Sent: Tuesday, September 20, 2016 7:06 AM To: Dean, Bill <Bill.Dean@nrc.gov>; Johnson, M ichael <M ichael.Johnson@nrc.gov>; Holian, Brian

<Brian .Holian@nrc.gov>; Uhle, Jennifer <Jennifer.Uhle@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

RE: October 12 NSIAC Reception and Dinner Outstanding, thanks Bill! I plan to h ave Theresa Clark coordinate with your guys, as necessary, to collect the right key messages on the seven topics highlighted below.

Vic From: Dean, Bill Sent: Monday, September 19, 2016 10:21 PM To: Mccree, Victor <Victor.McCree@nrc.gov>; Johnson, M ichael <Michael.Johnson@nrc.gov>; Holian, Brian

<Brian .Holian@nrc.gov>; Uhle, Jennifer <Jennifer .Uhle@nrc.gov>

Subject:

Re: October 12 NSIAC Reception and Dinner Vic I do not have an NSIAC meeting on my calendar.

Non-Responsive Record 1

Non-Responsive Record That's a start.

On: 19 September 2016 17:01 , "McCree, Victor" <Victor.McCree@nrc.gov> wrote:

Mike/Bill/Brian/Jennifer, I accepted Tony's invitation below. I'd appreciate your suggestions for any timely topics to include in my after-dinner remarks with the C NOs.

Are you scheduled to meet with the CNOs for this meeting? If so I'd be happy to tee up a few items for you to drive home the next day ... or dive deeper if you're not set to meet with them until much later.

Vic From: PIETRANGELO, Tony [2]

Sent: Wednesday, September 07, 2016 3:58 PM To: Mccree, Victor <Victor.McCree@nrc.gov>

Subject:

[External_Sender] October 12 NSIAC Reception and Dinner Hi Vic, I hope al l is well. I am writing to inquire about your availability to attend the October 12 NSIAC reception and dinner and to provide some after-dinner remarks. The event is from 6:00 to 9:00 pm in downtown DC near NEl's offices (probably at the Marriott Metro Center). I hope you will be able t o join us.

Thanks.

Best, Tony Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer Nuclear Energy Institute 1201 F Street NW, Suite 1100 Washington, DC 20004 (OJ 202-739-8081 (M) l(b)(6) I Email: arp@nei.org 2

TAKE THE NEI FUTURE OF ENERGY QUIZ, WWW. NEl .org/whynuclea r FOLLOW USON

/ectronic message transmission contains Information from the Nuclear Energy Institute, Inc. The information is intended solely for the use of the r,ddressee and its us ony other per . I ou are not the intended recipient, you have received this communicr,tlon In error, and any review, use, d r istribution of the conte11ts af this communication is s r c u have received this electronic transmission in error, I er Immediately by telephone or by electronic mail and permanently delete the orlginal message. IRS ,r  : r: nee with requirements Imposed by the IRS and other taxing authorities, we inform you that any tax advice contained in thl me uding any o 'n ended or written to be used, and cannot be used, for the purpose of (i) avoiding penalties that n any taxpayer or (ii) promoting, marketing or recommending to an sactlon or matter addressed herein.

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From: Clark, Theresa Sent: Sunday, September 11, 2016 8:34 PM To: McCree, Victor; Holahan, Gary; Johnson, Michael; Tracy, Glenn

Subject:

RE: Exelon Backfit Appeal key messages Attachments: EDO EXELON BACKFIT APPEAL DECISION - Comm Message Map rl.docx Thanks, Vic! I t hink t his will be super helpful. Please note that I correct ed two tiny errors in the attached. I tracked changes, but you can print wit hout markup or accept changes before printing, and it will just show the final.

Thanks, Theresa From: Mccree, Victor Sent: Sunday, September 11, 2016 7:06 PM To: Holahan, Gary <Gary.Holahan@nrc.gov>; Johnson, Michael <M ichael.Johnson@nrc.gov>; Tracy, Glenn

<G lenn.Tracy@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

RE: Exelon Backfit Appeal key messages See the attached message map (attached) for use in communicating the subject decision internally.

[Gary] note that I edited the key message for clarity and succinctness.

Vic From: Holahan, Gary Sent: !Friday, September 09, 2016 1:19 PM To: Mccree, Victor <Victor.McCree@nrc.gov>

Cc: Johnson, Michael <M ichael.Johnson@nrc.gov>; Tracy, Glenn <G lenn.Tracy@nrc.gov>; Clark, Theresa

<:rheresa.Clark@nrc.gov>

Subject:

Exelon Backfit Appeal key messages

Vic, Some thoughts on messages associated with the Exelon backfit appeal decision ...

Facts:

The staff issued a compliance backfit letter to Exelon (October 9, 2015) on the issue of pressurized overfill and safety valve performance during Condition II events (ANS Condition II categorization as frequent events).

Exelon twice appealed the compliance backfit (once at the office level, then at the EOO level) as inappropriate since, in their view, the staff failed to identify any error or omission that make the previously approved analysis incorrect.

The EOO established a panel of senior staff and managers to review the Exelon backfit appeal.

The backfit appeal panel recommended, and the EDO supported, a reversal of the compliance backfit, agreeing w ith the Exelon appeal.

Key Messages:

  • - , . * ..* ,.,. - - N T< .,...~'"*'*,._..,. --~f'--~r--~ . .,,... , _ *.,., .....r......*"'!""-*.,..,*,.,.,._,...,. ~ ...... ...... .._..~,,-_..... ....,,.... ..-,..! ' - - . ' ' * ,,,.,, _.,.,..._..,,.,...,,., .......,-,.***~,,.,,J*,,..,,,,... **** ,.,-,,..

~ -: -**-*- ,** *v, *,.v*,,_.,,~ , ~,...., .,._. ,__ ...

The staff takes its responsibility for assuring safety very seriously; and pursues backfits when it concludes they are appropriate to assure or enhance safety. The staff endeavors to perform thorough, technically sound, and legally well-founded reviews in all cases.

On complex technical and legal matters there can be differing views either within the staff, or with licensees and other stakeholders. The staff considers resolution of comments and alternative views, including appeals, important and deserving of serious attention. The backfit panel and the EDO have given this issue the expected, thorough and serious attention.

In this case the Panel concluded, and the EDO agreed, that the NRC staff's position on valve qualification in the Backfit safety evaluation is a new or modified interpretation of what constitutes compliance in addressing potential Pressurizer Safety Valve failures following water discharge. Although this new staff position represents a well-intentioned and conservative approach that could provide additional safety margin, it does not provide a basis for a compliance backfit.

Consistent with its commitment to be an effective, efficient, and predictable regulator; and in a proactive approach to important safety and legal matters, on June 2, 2016 the EDO tasked the Committee to Review Generic Requirements (CRGR) with undertaking a review of NRC implementation of agency backfitting and finality guidance. This effort includes an assessment of the clarity and effectiveness of backfitting requirements, guidance and criteria . It also includes an assessment of staff training and knowledge management.

The staff holds itself to high technical, safety, and legal standards and is committed to excellence in fulfilling its responsibilities. The staff is also committed to learn from experience, just as it expects of its licensees.

Gary 2

Communications Message Map: EDO EXELON BACKFIT APPEAL OECI SION: SEPTEMBER 12, 2016 KEY MESSAGES NRC ACTIONS/ACTIVITIES BACKGROUND On June 22, 2016, the EDO estabiished a Ba ckfit Appeal The NRC staff issued a compliance Review Panel (Panel) of senior staff and man agers to backfit letter to Exelon (October 9, 2015) review the Exelon backfit appeal. on the issue of pressurizera overfill and safety valve performance during On August 24, 2016, the Panel recommendec , and the Condition 11 events (ANS Condition II EDO supported, a reversal of the compliance backfit, categorization as frequent events).

agreeing with the Exelon appeal.

Exelon twice appealed the compliance

_ _ _ _....., backfit (once at the office level, then at NEXT ST!;PS the EDO level) claiming it was

- - - - - l inappropriate since, in their view, the staff d failed to identify any error or omission mber 13, that make the previously approved analysis incorrect.

ector, NRR, The NRC regulation for backfitting eptember 13, ( 10 CFR 50.109) indicates that "the compliance exception is intended to address situations where the licensee see (Exelon) has failed to meet known and established

' constitutes compliance in addressing of this decision. Date: September 13, 2016 (pm) standards of the Commission because of potential pressurizer safety valve failures omission or mistake of fact.. .. new or following water discharge, and did not OPA, in coordination with OEDO, will issue a news release modified interpretations of what provide a basis for a compliance backfit. (preferably) or BLOG post as an alternative, announcing constitutes compliance would not fall this decision: September 13, 2016 (pm) . within the exception .... "

Previously (on June ~2:. 2016) the EDO tasked the Committee to Review Generic The EDO will send a letter to NEI in response to its earlier

~equirements (CRGR) to review NRC ;I* letter supporting Exelon's backfit appeal.

implementation of agency backfitting and Date: September 14, 2016 finality guidance. This effort, which includes an assessment of the clarity and effectiveness of backfitting requirements, guidance and criteria, staff training and knowledge management, will incorporate insights from the EDO's

decision on this matter.

From: Holahan, Gary Sent: Monday, September 12, 2016 4:48 PM To: McCree, Victor; Clark, Theresa Cc: Johnson, Michael; Tracy, Glenn; Lewis, Robert

Subject:

Exelon decisioin Message Map Attachments: EDO EXELON BACKFIT APPEAL DEOSION - Comm Message Map rev 2 2016 09 12.docx FYI

... latest version (rev 2) 1

Communications Message Map: EDO EXELON BACKFIT APPEAL OECISJOt,i:-- SEPTEMBER 12, 2016 KEY MESSAGES NRC ACTIONS/ACTIVITIES BACKGROUND

  • --------------------+----------------------

. The staff takes its responsibility for On June 22, 2016, the EDO established a Backfit Appeal The NRG staff issued a compliance assuring safety very seriously; and Review Panel (Panel) of senior staff and managers to backfit letter to E)(elon (October 9, 2015) pursues technically sound, and legally review the Exelon backfit appeal. on the issue of pressurizer overfill and well-founded backfits when it concludes safety valve performance during they are needed to assure or enhance On August 24, 2016, the Panel recommended, and the Condition II events (ANS Condition II safety_ EDO supported, a reversal of the compliance backfit, categorization as frequent events).

. agreeing with the Exelon appeal.

On complex technical and legal matters Exelon twice appealed the compliance there can be differing views either within the staff, or with licensees stakeholders. The NRC used and other its formal NEXT STEPS

1 Ibackfit (once at the office level, then at the EDO level) claiming it was inappropriate since, in their view, the staff backfit review process to ensure this The EDO will verbally inform the Chairman and failed to identify any error or omission issue received appropriate consideration. Commissioners of this decision. Date: September 13, that make the previously approved 2016 (am) analysis incorrect.

In this case, the EDD concluded that the NRC staff's position on valve The EDO will send a memorandum to the Director, NRR, The NRC regulation for backfitting qualification, valve performance, and the formally notifylng him of this decision. Date: September 13, a (1 CFR 50.109) indicates that "the application of the single failure criterion 2016 (am) compliance exception is intended to in the backfit safety evaluation was a address situations where the licensee  !

new or modified interpretation of what The EDO will send a letter informing the licensee (Exelon) has failed lo meet known and established constitutes compliance in addressing of this decision. Date: September 13, 2016 (pm) standards of the Commission because of potential pressurizer safety valve failures omission or mistake of facL.new or following water discharge, and did not The EDO will call Exelon to inform them of his decision. modified interpretations of what provide a basis for a compliance backfit. Date: September 13, 2016 (pm} constitutes compliance would not fall within the exception .... "

Previously (on June 9, 2016} the EDO The EDO will send a letter to NEI in response to its earlier tasked the Committee to Review Generic letter supporting Exelon's backfit appeal.
  • Requirements (CRGR) to review NRC Date: September 14, 2016 (pm) implementation of agency backfitting and finality guidance. This effort, which I The EDO will Call NEI to inform them of his decision. Date:

includes an assessment of the clarity September 13, 2016 (pm) and effectiveness of backfitting requirements, guidance and criteria, staff I OPA, in coordination with OEDO, will issue a news release training and knowledge management, , (preferably) or BLOG post as an alternative, announcing will incorporate insights from the EDO's this decision: September 13, 2016 (pm) decision on this matter.

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From: West, Steven Sent: Tuesday, December 06, 2016 8:53 AM To: Johnson, Michael Cc: Inverso, Tara; Clark, Theresa; Holian, Brian; Bowen, Jeremy

Subject:

Backfitting presentation Attachments: 20161206 Focus on Backfitting with comments - Steve Westpptx

Mike, My slides for the backfitting presentation I'm delivering in Region 3 tomorrow are attached for your information and use. I'm still noodling my talking notes. The presentation slides may include points you can use in Region
4. For the Exelon appeal, which is Region 3's main interest," I drew heavily from Theresa's "Safety Spotlight." Have a safe trip.

Steve Steven West, Deputy Director Office of Nuclear Security and Incident Response U.S. Nuclear Regulatory Commissio n 301-287-3734 Steven.West@nrc.gov 1

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What reasons? l--lnHfd ~mi:~ NHrlm ll,~_ul*HJI Y Cttmm11\ltlil A'fJlitt*ti,w ~fJpk f11iil. t//1 Em,,,,,,,,m~11t

  • To ensure adequate protection.
  • A substantial increase in protection to public health and safety is cost justified.
  • To bring a facility into compliance.

\_ --/ U.S.NRC Why focus on this now? W1iml 4!:llf\ Nudur ll':flHl~ro1:y f:ommioilrn Pffllfeti,w Jtgp/e aniltlJtP.Hv111mment

  • Higher than normal interest by internal and external stakeholders.
  • Concerns about over use of the compliance exception.
  • Questions about consistency.
  • Questions about the use of qualitative factors.

<< J

What is CRGR's role? \.llllfHf ~Im! N11l-lfa* R:~1111l:1ro1_r f'nmmi,~1011 Pffllfffi,W PefJ/Jfl (f11ti 'fbi! /S1UJiffJHHliHl Ensure that certain proposed generic backfits are appropriately justified based on the backfit provisions of the applicable NRC regulations and the Commission's backfit policy.

What is the EDO having the ,-/ U.S.NRC cRGR dQ? YllllHI ~Im\ NHllt'~[ ntij\lli!Hlf)'

Prtrtttm11 Jtopk anil lh1A'lw1HJnmem f:t1mm1~mlii

  • Assess backfit requirements, guidance, criteria and procedures.
  • Assess knowledge management on backfitting.
  • Provide any findings with corrective actions and milestones.

Wh at ' s t he EXe IOn ba Ckt*It a bOut -"r ~-,)u.S.NRC

\Jniml Rrm~ Ntttlm lt!'~Hl~Hlfr fnmn1i1~1on

~tttliJll PetJpk a1,tl the P.lu1iPtJ11ment

  • In 2015 staff took positions different from those it had taken in 2001 and 2004.
  • Staff concluded that the 2001 and 2004 positions were in error.
  • Staff used the error as a basis for backfitting under the compliance exception.
  • Staff directed Exelon to resolve the non-compliance.

\--/ u.S.NRC What about the appeal? \-illHHI ~lam NHfl~ar ke1wl,tro1v h1mm1,imn Pffltttli,W ~pit tlml tlJtEHvironmf!nt

  • Exelon appealed the staff's backfit decision.
  • Exelon took the unusual step of appealing th~

appeal.

  • EDO appeal review panel concluded that the compliance exception was not justified.
  • 2015 positions used to justify the backfit differed from those taken in 2001 and 2004.
  • 2001 and 2004 positions were not in error.
  • 2015 staff positions were not a basis for a compliance backfit.
  • Staff should also address any generic aspects of the issue.
  • Staff should also re-evaluate relevant generic communications.

Let's discuss why this W111ml !jWl'\ N11~lm lteijHl!lltlf; f:,imm1\~lnfl ProtettiHp lwplt, anti t-~ lfnv1f¥JmHt!Hl matter is important?

  • Consistent application of backfit rule is expected and necessary.
  • Appeal provided opportunity for real time assessment of staff handing of backfit issue.
  • Appeal review yielded insights on differences between plant-specific and generic reviews.
  • Appeal outcome validated certain aspects of stakeholder concerns.
  • CRGR tasking is uncovering areas for improvement.
  • Reached out to stakeholders including a public meeting. A second meeting is being considered.
  • CRGR completing assessment of adequacy of requirements, guidance and training by office.
  • Coordinating with offices and regions.
  • Considering OGC input.
  • Deliverable to EDO.
  • Communicating results to stakeholders.
  • Implementation.
  • Future effectiveness review.

What are the key messages ,--/u.S.NRC 1ln11fd liwf~ Nuflm Reijnl~rm_r Cn111m1~11011

/JfflteftitfX ftopk from the appeal review panel? ffHff llJt lfiivlffffl1H~Ht

  • Consistent with NRC mission and values, it is the staff's right and responsibility to raise safety concerns.
  • It is the agency's responsibility to evaluate and document the resolution of such concerns.
  • Evaluation of issues should usually consider the licensing basis, precedents, and safety significance.
  • Compliance backfits are justified for failure to meet "known and established" NRC standards at time of

How can I learn more? 1/nlfHI ijm..~ Nudm ll~1l1nnr (.;omm1.-11rn Pffllf>ffiHJt l¥fl/JIP ami tb~ E,,fJirtmmfnt

  • ilearn training: Backfitting and Issue Finality (Web-Based) (1D_212144)

ML16236A202 and ML16236A208 (includes many additional references)

  • * *** **-***-* ..... *.** **~ -*- .. ** - .,,.* , *.u ~ ~-* ..*~-;*,~,..... .. - .....- .- .............._._'f'<...'Y""- - - -- -~--.....,..,....-............~-- - ......... ,

From: Clark, Theresa Sent: Thursday, September 15, 2016 9:54 AM To: Valliere, Nanette; Ruesch, Eric; Castleman, Patrick; Frazier, Alan; Krsek, Robert Cc: Lewis, Robert; Rasouli, Houman; Inverso, Tara; Bowen, Jeremy; Holahan, Gary

Subject:

FYI: backfit appeal documents signed Good morning, all!

This morning, Vic signed the three documents associated with the Byron/Braidwood backfit appeal. They are being processed now, and we expect that they (along with the panel documents referenced within) will be made publicly available in ADAMS later today. Please let me know if you have any questions. Thanks I

  • letter responding to E><elon: ML16243A067 1The 3 records are all publicly available in ADAMS.
  • Letter responding to NEI: Ml16246A150
  • Memo to NRR: ML16246A247 Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0-16E22 1

. ~------~-:::-::"::--,:-,-::,------,------====:c-~==--::-::--::---:--c,---~ ~

Communications Message Map: EDO EXELON BACKFIT APPEAL DECISION: SEPTEMBER 12, 2016 KEY MESSAGES NRC ACTIONSIACTMTIES BACKGROUND

,-.---------------*-*---+------------------------+-----------------

The staff takes its responsibility for On June 22, 2016, the EDO established a Backfit Appeal The NRC staff issued a compliance assuring safety very seriously; and Review Panel (Panel) of senior staff and managers to backfit letter to Exelon (October 9, 2015) pursues technically sound, and legally review the Exelon backfit appeal. on the issue of pressurizer overfill and well-founded backfrts when it concludes safety valve performance during they are needed to assure or enhance On August 24, 2016, the Panel recommended, and the Condition II events (ANS Condition II safety. EDO supported, a reversal of the compliance backfit, categorization as frequent events).

agreeing with the Exelon appeal.

On complex technical and legal matters Exelon twice appealed the compliance there can be differing views either within 1 - - - - - - - - - - - . , * - - - - - - - - - - - - - - - 1 backfit (once at the office level, then at the staff, or with licensees and other NEXT STEPS the EDO level) claiming it was stakeholders. The NRC used its formal 1 - - - - - - - - - - - - - - - - - - - - - - - - - - 1 inappropriate since, in their view, the staff backfit review process to ensure this The EDO will verbally inform the Chairman and failed to identify any error or omission issue received appropriate consideration. Commissioners of this decision. Date: September 13, that make the previously approved 2016 (am) analysis incorrect.

In this case, the EDO concluded that the NRC staff's position on valve The EDO will send a memorandum to the Director, NRR, The NRC regulation for backfitting qualification, valve performance, and the formally notifying him of this decision. Date: September 13, ( 10 CF R 50. 109) indicates th at "the application of the single failure criterion 2016 {am) compliance ~xception is intended to in the backfit safety evaluation was a address situations where the licensee new or modified interpretation of what The EDO will send a letter informing the licensee (Exelon) has failed to meet known and established constitutes compliance in addressing of this decision. Date: September 13, 2016 (pm) standards of the Commission because of*

potential pressurizer safety valve failures omission or mistake of fact .... new or

. following water discharge, and did not  ; The EDO will call Exelon to inform them of his decision . modified interpretations of what provide a basis for a compliance backfit. Date: September 13, 2016 (pm) constitutes compliance would not fall within the exception .... n  ;

Previously (on June 9, 2016) the EDO The EDO will send a letter to NEI in response to its earlier tasked the Committee to Review Generic letter supporting Exelon's backfit appeal.

Requirements (CRGR) to review NRC Date: September 14, 2016 (pm) implementation of agency backfitting and finality guidance. This effort, which The EDO will Call NEI to inform them of his decision. Date:

includes an assessment of the clarity September 13, 2D16 (pm) and effectiveness of backfitting requirements, guidance and criteria, staff OPA, in coordination with OEDO, will issue a news release training and knowledge management, (preferably) or BLOG post as an alternative, announcing will incorporate insights from the EDO's this decision: September 13, 2016 (pm) decision on this matter.

In addition to the performance of pressurizer valves [MAs11for water relief, the Panel reoognizes

~the NRC staff-MS raised other issues in its October 2015Backfit ibetter in support of its backfit decision for Braidwood and Byron. The staff emphasized some of those issues in its comments on the preliminary Panel findings. The Panel summarizes its consideratlon of those issues as follows:

The staff stategs that ANS-51.1/N18.2-1973.A.merican {MAs21Nuslear Sos1ety (ANS) Standard ANS N 18.2 1973, "(:IJuslear Safety Criteria for the Design of Stationary Pressurize Water Reactor Plants,tt defines the categories of design basis transients and accidents based on an anticipated frequency of occurrence (annually for Condition II events).:., The staff also stated that it was a and the long-standing NRC position that escalation from one condition to another is not acceptable. The staff considers ANS-511/N18.2-1973i\Na ~J18.2 to constitute a known and established standard that has been reflected in NRC guidance documents and in the licensing basis of each U.S. nuclear power plant. This ANS standard is referenced for use in several Braidwood/Byron UFSAR Chapter 15 accident analyses. The Panel agrees that the non-escalation prohibition is an established standard applicable to Bryon and Braidwood, but did not identify historical evidence that implementation of this standard requires the Braidwood/Byron licensee to assume that its pressurizer valves will fail open under water relief conditions, to apply the single failure criterion to PSV failure in these circumstances. or to impose ASME requirements for certification or qualification of PSVs for water relief.

The staff raises concerns that extended steam and water relief through the pressurizer valves would result in the failure of the pressurizer relief tank rupture disk, would require repair of the damaged PSVs, and might cause an extended time period for the return to service of the nuclear power plant. The Panel does not consider the time period necessary for the licensee to perform radioactive clean-up activities in the containment building, to inspect and conduct any necessary repairs 10 the PSVs, or to prepare for plant startup, to constitute issues that support a compliance backfit imposed by the NRC. The NRC staff and inspectors would verify that these activities are conducted appropriately to protect the public health and safety prior to plant restart. The Backfit Letter states that UFSAR Section 15.5.1.3 "impliefs)" that the plant will return to operation in a "short period," but the Panel sees no support for a timing requirement in UFSAR Section 15.5.1.3. Also. the Panel has not identified a regulatory interest in limiting the time needed for the plant to return to operation The staff considers ~ h e known and established standard in question r:equirement into be the NUREG-0737 (Item 11.D.I) standard for licensees and applicants to conduct testing to qualify the RCS relief and safety valves under expected operating conditions for design-basis transients and accidents to be the known and established standard in question as a regulatory requirememt in r:esponse to the stuck open PORV at TMI 2 in 1979, As discussed above and in Appendix B to this report, the NRG accepted the EPRl testing to satisfy NUREG-0737 (Item II. D.1} for Braidwood and Byron in an SE forwarded by letters in 1988 and 1990. Therefore, the Panel considers this known and established standard referenced by the staff to have been met for Braidwood and Byron.

The staff stategs that an omission or mistake of fact occurred when the licensee failed to acknowledge that the EPRI testing program did not evaluate liquid discharge from the pressurizer valves during extended high pressure safety injection far Braidwood and Byron. As discussed in Appendix B to this report, the NRC SE on the Braidwood/Byron response to NUREG-0737 (Item 11.D.1) discussed the evaluation of the PSVs and PORVs for feedline break accidents including water relief, and found that the performance of the PSVs and PORVs was acceptable based on the EPRI tests. Therefore, the Panel does not agreeconsieler that aH

omission or R1istako of fact occurred by the licensee's reference to the EPRI testing program was an omission or mistake of fact.

In support of its backfit, the staff considers that flow capacity certification in accordance with the ASME BPV Code and inservice testing throughout the service life in accordance with the ASME OM Code to be necessary to satisfy the standard for water relief capability of pressurizer valves.

The staff points to its review of several nuclear power plants where water relief was not considered acceptable, or different actions were required. As discussed in Appendix C to this report, the staff required additional actions for some licensees to support reliance on the PORVs for water relief and to avoid water relief through the PSVs. The staff also has allowed some licensees to rely only on EPRI testing without significant additional activities. The Panel did not identify instances where the staff imposed certification by the ASME BPV Code and historical OM Code testing, or required alternatives to the ASME BPV or OM Codes, in the examples of NRC staff review of water relief capability for pressurizer valves.

The staff has also identified specific ASME Code provisions that purportedly support its position that ASME requirements apply to qualification of pressurizer valves for water relief. But the staff did not provide evidence that these provisions have consistently been interpreted as the staff is now interpreting them_ Given the NRC's treatment of Item II. D.1 in NUREG-0737 and the NRG staff's historical licensing practice, the Panel concludes that the staff's current applicationiAtereFetatieF(ST3J of the ASME Code is not supported by the historical record.

In light of the wide range of NRC staff positions during the review of pressurizer valve capability since the TMl-2 accident. the Panel agrees that the staff could have pursued insisted on additional actions by the Braidwood/Byron licensee to improve the reliability of the PSVs or PORVs for water relief, or to avoid water relief through the PSVs by PORV improvements, as part of the 2001 power uprate, or the 2004 PSV setpoint amendment, er tt:19 2Q14 FAeaswrement uncertainty capture upFateEW.~41. The staff may have been able to justify additional actions, but they determined that it was not necessary. The 2001 uprate staff reviewers used their expert engineering judgement to determine wl;ietl::ier it was nottesl=lnisally necessary to assume that the PSVs or PORVs would stick open with water relief based on EPRI test information, licensee supplemental information, and their own technical experience.

The staff raises a concern that the NRC review of the 2004 setpoint amendment for Byron and Braidwood did not re-review the qualification of the PSVs. The staff notes that if the 2001 SE had not found liquid discharge through the PSVs to be acceptable, it is unlikely that the staff would have roachea an approval conclusion in the 2004 SEapproved the 2004 setpoint amendment. In Appendix C to this report, the Panel summarizes the 2004 SE review of the PSV water relief capability. The Panel recognizes that a staff review may rely on a previous more extensive review to determine the acceptability of a similar request. The Panel does not consider th§is review approach used in 2004 to challenge the adequacy of the 2001 review.

5 ADDITIONAL PANEL THOUGHTS In addition to the specific finding relating to the backfit appeal, the Panel believes it is important to acknowledge, and for the NRC staff and licensee to appreciate, that water relief through an PSV not specifically designed for such service is undesirable and should be minimized or

avoided as a matter of conservative engineering and prudent operations. This is reinforced by the information provided in Westinghouse NSAL-93-013 and its supplement, and the actions by various licensees in response to the NSAL, as well as the limited scope of the EPRI testing conducted over 30 years ago.

Operator training, and Emergency Operating Procedures to terminate the event before pressurizer filling, as well as the use of PORVs rather than reliance on PSVs, are clearly preferred and prudent measures, whether they form the facilities' UFSAR licensing basis and are assumed in the accident analyses or not.

The PSVs in question were designed for steam service. Steam relief is their normal service condition; and applies to their certification. The Panel supports the previous staff determinations for Braidwood/Byron and certain other plants that PSVs experiencing water relief during an abnormal or accident condition need not be assumed to fail since if there W!li& a reasonable and technically well-informed engineering judgement to the contrary. However, the Panel also considers the actions by various licensees aREI aGGepted ey the sta# to improve the reliability and performance of the PORVs to avoid water relief through the PSVs to be prudent in light of their design specifications.

The Panel considered but could not determine the extent to which the Braidwood/Byron licensee addressed crediting water relief through the PSVs, PORVs, or PORV block valves in the Braidwood and Byron 1ST Programs. The Panel recognizes that the difference between the intended use of these valves for overpressure protection and their infrequent response to plant events might be considered in implementing appropriate 1ST activities.

The Panel notes that water relief through various pressurizer valves is not a new issue because water relief has always been credited (by the Braidwood/Byron licensee and other licensees) for the UFSAR Feedwater System Pipe Break analysis {UFSAR Section 15.2.8{MAS5J).

Tl=le PaRel Believes tl=lese tl=le~gl=lts te ee weftl';i 1Nl=lile ier seRsieteratieR ~Y tl=le f)JRS staff aAS l!l~aid\f.*88Gll8yr8R liC8R888.

T. Scarbrough July 125, 2016 Revision 1 Pressurizer Valve Timeline March 1979: TMl-2 Accident with stuck open pressurizer PORV.

July 1979: NUREG-0578 with TMl-2 Lesson Learned.

November 1980: NUREG-0737 includes letter to all licensees and applicants with TMI requirements, including Item It. D.1, "Performance Testing of Boiling-Water Reactor and Pressurized-Water Reactor Relief and Safety Valves" [NRC referenced EPRI testing program].

July 1981: SRP (Rev. 1) Chapters 15.5.1-15.5.2, and 15.6.1 provide general guidance for staff review of IOECCS, eves malfunction, and IOPRV April 1982: D. Hoffman (Consumers) to H. Denton summarizes EPRI program with Crosby report indicating EPRI results applicable 10 Crosby valves.

July 1982: 0. Kingsley to H. Denton submits WCAP-10105 with review of EPRI test data.

December 1982: EPRI NP-2628-SR summarizes EPRI safety and relief valve tests.

January 1983: EPRI NP-2770-LD describes testing of PWR safety and relief valves in response to NUREG-0578 and NUREG-0737.

1985-87: Byron and Braidwood Operating Licenses issued.

Braidwood NUREG-1002 (Nov. 1983) discusses overall response to TMI action items.

Braidwood NUREG-1002 S1 (Sept. 1986) states applicant's approach acceptable based on preliminary review.

Byron NUREG-0876 S5 (Oct. 1984) includes similar discussion.

Byron NUREG-0876 SB (March 1987) states Item 11.D.1 is closed based on S5.

August 1988: NRC letter with TER closing NUREG-0737, Item 11.0.1, for Byron.

May 1990: NRC letter with TER closing NUREG-0737, Item 11.D.1, for Braidwood.

June 1993 and October 1994: Westinghouse issued NSAL-93-013 and supplement in response to potentially non-conservative assumptions in IOECCS analysis.

June 1997: Salem TS changes to ensure automatic PORV capability (including circuitry modifications, air supply capability, Wyle endurance tests, reactor operator performance, and TS changes) in response to NSAL-93-013 to avoid PSV water relief.

June 1998: NRC issues Millstone Unit 3 TS amendment for PORVs. Revised TS Bases states PORV qualified for water relief. Prevent water relief from PSVs for which qualification not demonstrated. Licensee references EPRI and manufacturer tests to support PORV water relief.

GL 89-10 specified for PORV block valve capability.

SeptemberMay 2000: NRC issues Callaway proposed TS amendment to modify PSV setpoint that prevents water relief. PORVs indicated to be capable of water relief in Callaway request.

November 2000: ComEd responded to RAI on Byron/Braidwood Stretch Uprate Request by discussing EPRI testing program.

January 2001: Exelon responded to supplement RAI with more detail on EPRI testing program.

May 2001: NRC issues Byron/Braidwood Stretch Power Uprate with EPRI NP-2770-LD.

December 2002: Byron/Braidwood UFSAR {Rev. 9) Chapter 15.5.1 includes PSV water relief, and references lNEL 1988 report and L. Olshan August 1988 SER.

December 2003: NRR Review Standard for Extended Power Uprates (RS-001, Rev. 0) states in Item 8 on page 7 that pressurizer level should not be allowed to reach a pressurizer water-solid condition.

August 2004: NRC issues Byron/Braidwood PSV setpoint amendment that accepts water relief from PSVs with reference to 2001 Stretch Power Uprate.

December 2005: RIS 2005-29 indicates NRC staff disagreement with NSAL provisions.

March 2007: Revision 2 to SRP Chapters 15.5.1-15.5.2 and 15.6.1 provide details on evaluating these events, including PSVs and PORVs fail if relieve water without being qualified.

June 2012: NRC issues Turkey Point EPU with SER indicating action taken to avoid filling pressurizer, and PSV or PORV water relief. Block valve used to isolate open PORV. No discussion of NSAL.

September 2012: NRC issues St. Lu~ie Unit 2 EPU with SER stating that filling pressurizer avoided so PSVand PORV will not stick open. Block valve used to isolate PORV. No discussion of NSAL.

February 2014: NRC issues Byron/Braidwood MUR power uprate with IOECCS out of scope.

December 2014: Byron/Braidwood UFSAR (Rev. 15) Chapter 15.5.1 states that IOECCS does not progress beyond IOPSRV because the 3 PSVs will reclose and leak less than 1 PSV.

October 2015: NRC issues backfit for Byron/Braidwood where the staff determines that Byron/Braidwood are not in compliance with GDC 15, 21, and 29, 10 CFR 50.34(b), and plant-specific design bases.

December 2015: Exelon submits appeal of backfit.

May 2016: NRG responds to backfit appeal.

June 2016: Exelon submits appeal of backfit.

T. Scarbrough July 5, 2016 Pressurizer Valve Timeline March 1979: TMl-2 Accident with stuck open pressurizer PORV.

July 1979: NUREG-0578 with TMl-2 Lesson Learned.

November 1980: NUREG-0737 includes letter to all licensees and applicants with TMI requirements, including Item 11.D.1, "Performance Testing of Boiling.Water Reactor and Pressurized-Water Reactor Relief and Safety Valves" [NRC referenced EPRI testing program].

July 1981: SRP (Rev. 1) Chapters 15.5.1-15.5.2, and 15.6.1 provide general guidance for staff review of IOECCS, eves malfunction, and IOPRV April 1982: D. Hoffman (Consumers) to H. Denton summarizes EPRI program with Crosby report indicating EPRI results applicable to Crosby valves.

July 1982: 0. Kingsley to H. Denton submits WCAP-10105 with review of EPRI test data.

December 1982: EPRI NP-2628-SR summarizes EPRI safety and relief valve tests.

January 1983: EPRI NP-2770-LD describes testing of PWR safety and relief valves in response to NUREG-0578 and NUREG-0737.

1985-87: Byron and Braidwood Operating Licenses issued.

Braidwood NUREG-1002 (Nov. 1983) discusses overall response to TMI action items.

Braidwood NUREG-1002 S1 (Sept. 1986) states applicant's approach acceptable based on preliminary review.

Byron NUREG-0876 S5 (Oct. 1984) includes similar discussion.

Byron NUREG-0876 S8 (March 1987) states Item 11.D.1 is closed based on S5.

August 1988: NRC letter with TER closing NUREG-0737, Item 11.D.1, for Byron.

May 1990: NRC letter with TER closing NUREG-0737, Item 11.D.1, for Braidwood.

June 1993 and October 1994: Westinghouse issued NSAL-93-013 and supplement in response to potentially non~conservative assumptions in IOECCS analysis.

June 1997: Salem TS changes to ensure automatic PORV capability (including circuitry modifications, air supply capability, Wyle endurance tests, reactor operator performance, and TS changes) in response to NSAL-93-013 to avoid PSV water relief.

June 1998: NRC issues Millstone Unit 3 TS amendment for PORVs. Revised TS Bases states PORV qualified for water relief. Prevent water relief from PSVs for which qualification not demonstrated. Licensee references EPRl and manufacturer tests to support PORV water relief.

GL 89-10 specified for PORV block valve capability.

May 2000: Callaway proposed TS amendment to modify PSV setpoint that prevents water relief. PO RVs indicated to be capable of water relief.

November 2000: ComEd responded to RAI on Byron/Braidwood Stretch Uprate Request by discussing EPRI testing program.

January 2001: Exelon responded to supplement RAI with more detail on EPRI testing program.

May 2001: NRC issues Byron/Braidwood Stretch Power Uprate with EPRI NP-2770-LD.

December 2002: Byron/Braidwood UFSAR (Rev. 9) Chapter 15.5.1 includes PSV water relief, and references INEL 1988 report and L. Olshan August 1988 SER.

August 2004: NRC issues Byron/Braidwood PSV setpoint amendment that accepts water relief from PSVs with reference to 2001 Stretch Power Uprate.

December 2005: RIS 2005-29 indicates NRC staff disagreement with NSAL provisions, March 2007: Revision 2 to SRP Chapters 15.5.1-15.5.2 and 15.6.1 provide details on evaluating these events, including PSVs and PORVs fail if relieve water without being qualified.

June 2012: NRC issues Turkey Point EPU with SER indicating action taken to avoid filling pressurizer, and PSV or PORV water relief. Block valve used to isolate open PORV.

September 2012: NRC issues St. Lucie Unit 2 EPU with SER stating that filling pressurizer avoided so PSV and PORV will not stick open. Block valve used to isolate PORV.

February 2014: NRC issues Byron/Braidwood MUR power uprate with IOECCS out of scope.

December 2014: Byron/Braidwood UFSAR (Rev. 15) Chapter 15.5.1 states that IOECCS does not progress beyond lOPSRV because the 3 PSVs will reclose and leak less than 1 PSV.

December 2015: Exelon submits appeal of backfit.

May 2016: NRC responds to backfit appeal.

June 2016: Exelon submits appeal of backfit.

T. Scarbrough July 17, 2016 Questions on Byron/Braidwood Backfit Appeal NRR Questions

1. In that Volume 6 of EPRI NP-2770-LD cannot be located in the NRG record system, does the staff have a copy that can be reviewed and entered into ADAMS?
2. Does the staff have the design and procurement specifications for the Byron/Braidwood pressurizer safety valves (PSVs) and power operated relief valves (PORVs)?
3. What was the basis for determining that the concern regarding reliance on water relief by PSVs and PORVs was outside the scope of the NRG safety evaluation in February 2014 for the Byron/Braidwood MUR power uprate request?
4. Did the staff evaluate the Byron/Braidwood response to NUREG-0737, Item 11.D.1, and make a determination as ta its acceptability in preparing the Byron/Braidwood backfit evaluation dated October 9, 2015?
5. Did the staff evaluate (with assistance from Region Ill) the operating experience evaluation by the Byron/Braidwood licensee in response to Westinghouse NSAL-93-013 and its Supplement 1?
6. Did the staff evaluate any follow-up request by the Byron /Braidwood licensee following its withdrawal of the May 29, 1998, TS amendment request for reliance on the PORVs in response ta inadvertent operation of the emergency core cooling system (IOECCS)?
7. Does the staff have information related to operating experience or test results where PSVs or PORVs failed in the open position following relief of water in support of this assumption in the backflt technical evaluation? For example, did the staff evaluate the study described in EPRI TR-1011047 (August 2004), "Probability of Safety Valve Failure-to-Reseat Following Steam and Liquid Relief- Quantitative Expert Elicitation," regarding the potential increase in failure rates following steam and liquid relief, including its Appendix Bon safety valve test data evaluation? Also, did the staff evaluate NUREG!CR-7037 (March 2011 ), "Industry Performance of Relief Valves at U.S. Commercial Nuclear Power Plants through 2007," as part of its review? With respect to Pressurizer PORVs, the report found four separate liquid relief events at four PWR plants. The report estimated 698 total demands on these PORVs during their liquid relief events with no failures to close. The report also summarized test data from EPIX for three valve types. The report indicated 2 failures of PORVs ta reclose during 2070 demands, but did not specify liquid or steam service for the EPIX test information. With respect to PSVs, the report indicated 2 failures out of 4 total demands following plant scrams, but did not indicate liquid or steam service. The report summarized EPIX test data for PSVs as no failures to reclose during 1805 demands.
8. Did the staff discuss with the Byron/Braidwood licensee its safety concerns with water relief through the PSVs and PORVs, and the possible upgrade of the PORVs to avoid water relief through the PSVs (similar to implemented at other nuclear power plants, such as Salem in 1997 and Millstone Unit 3 in 1998), rather than ASME BPV Code qualification of the PSVs specified in the backfit safety evaluation?
9. Did the staff discuss with the applicable valve manufacturers whether the PSVs used in nuclear power plants can be qualified to the ASME BPV Code for water relief in that NB-7171, Safety Valves, in Article NB-7000, Overpressure Protection, specifies that safety valves are qualified for steam service or air and gas service, whereas NB-7172, Safety Relief Valves, indicates that these valves may be qualified for steam, air and gas, and liquid service?

1O. Did the staff verify that the Byron/Braidwood licensee had continued implementation of the GL 89-1 O and GL 96-05 provisions for the PORV block valves to maintain their capability to close under all design-basis event conditions in light of the latest UFSAR revision?

11. Did the staff discuss with the Byron/Braidwood licensee the application of staggered setpoints for the PORV and each PSV to allow one valve to provide relief to avoid challenging the other valves?
12. While RIS 2005-29 states that closing the PORV block valve is not acceptable in isolating an open PORV, did the staff review other NRC documents (such as St. Lucie Unit 2 EPU SER, dated September 24, 2012, and Turkey Point EPU SER, dated June 15, 2012) to verify that the staff has always taken this position?
13. Did the staff hold a public meeting with the Westinghouse Owners* Group discuss an appropriate response to NSAL-93-013 and its Supplement 1, and whether all Westinghouse-designed PWRs had implemented an acceptable response in light of the statement in the backfit technical evaluation that recommendations in NSAL-93-013 were unacceptable?
14. Did the staff consider issuing a generic letter under 10 CFR 50.54(f) to determine the response of all PWR licensees to the safety concerns regarding the perfom,ance of PSVs and PORVs under liquid conditions as identified in NSAL-93-013?
15. In preparing the response to the backfit appeal dated May 3, 2016, what was the basis for stating that "the NRC staff had some awareness of an approach inconsistent with the requirements discussed here" in light of the NRC review of the EPRI tests in several safety evaluations (such as the August 1988 NRC letter with INEL TER closing NUREG-0737, Item 11.D.1 for Byron; May 1990 NRG letter with INEL TER closing NUREG-0737, Item 11.D.1, for Braidwood; June 1998 NRC safety evaluation for Millstone Unit 3 TS amendment that referenced EPRI testing program; and May 2001 Byron/Braidwood Stretch Power Uprate that referenced EPRI testing program)?

Byron/Braidwood Questions

1. Did the licensee submit a copy of Volume 6 of EPRI NP-2770-LD on the docket in that it discusses the EPRI testing in its submittals?
2. Are the design and procurement specifications for the Byron/Braidwood pressurizer safety valves (PSVs) and PORVs available for NRC review?
3. What 1ST Program tests are conducted for the PORVs, PORV block valves, and PSVs to demonstrate their operational readiness with the reliance on these valves in the accident analyses (for example, tested for both steam and liquid service)?
4. Has the Byron/Braidwood licensee continued implementation of the GL 89-10 and GL 96-05 provisions for the PORV block valves to maintain their capability to close under all design-basis event conditions in light of the latest UFSAR revision?
5. What were the design-basis performance requirements for the PSVs and PORVs regarding steam and liquid service when the NRC staff reviewed the Byron/Braidwood operating license applications using the NRC Standard Review Plan (Revision 1, dated July 1981 ),

Chapters 15. 5.1-15.5.2 and 15.6.1? Did the licensee submit documentation on the docket specifying reliance on liquid service for the PSVs and PORVs?

6. What were the licensee's actions to address the safety concerns raised in Westinghouse NSAL-93-013 for Byron/Braidwood? Where is this documented for NRC review?
7. In light of the limited testing performed by EPRI for the types of PSVs and PORVs used at Byron/Braidwood, what is the licensee's basis for determining that these valves are "qualified" for liquid service where they have not been subject to ASME BPV Code or National Board qualification testing for such service?
8. Did Westinghouse and the applicable valve manufacturers agree with the decision by the licensee to assume that the PSVs and PORVs were "qualified" for liquid service based on the limited EPRI testing?
9. Did the licensee evaluate more recent EPRI studies that discuss the potential far failure of PSVs during liquid service based on unstable test results during the EPRI testing in the 1980s? See EPRI TR-1011047 (August 2004), "Probability of Safety Valve Failure-to-Reseat Following Steam and Liquid Relief- Quantitative Expert Elicitation," that states in Appendix B that "[b]ecause these valves are not designed for liquid flow, and because EPRI tests with subcooled liquid led to unstable conditions more often than not, the likelihood of PSV failure during an SBO [station blackout] accident would be quite high."
10. Did the licensee resolve the concerns to be addressed in its proposed license amendment request (LAR) dated May 29, 1998, to upgrade the PORVs that was withdrawn on July 16, 1999?
11. In the May 29, 1998, LAR, the licensee stated that "the PSRVs have not been qualified to reseat after passing subcooled liquid." This statement was made with knowledge of the EPRI testing conducted in the 1980s. However, the Byron/Braidwood UFSAR (Revision 9)

in December 2002 included reliance on water relief through the PSVs. What was the licensee's basis for changing its technical evaluation of the capability of the PSVs to perform under liquid service?

12. What is the current safety-related status of the PORV and its circuitry, including its single failure reliability?
13. Could PORV availability be used to avoid water relief through the Byron/Braidwood PSVs in response to design-basis events?

From: Mccree, Victor Sent: Sunday, September 11, 2016 6:14 PM To: Clark, Theresa; Holahan. Gary Cc: Johnson, Michael; Tracy, Glenn

Subject:

RE: backfit documents Attachments: EDO Memo to NRR FINAL3.docx Please incorporate my final edits (attached). Thanks again!

Vic From: Clark, Theresa Sent: Saturday, September 10, 2016 9:32 PM To: Mccree, Victor <Victor.McCree@nrc.gov>; Holahan, Gary <Gary.Holahan@nrc.gov>

Subject:

RE: backfit documents Hi Vici I worked with Tom on an updated response to question 2.c that is much stronger/more specific. It along with the rest of your reclama response are now included in the enclosure to the memo to NRR, both in ADAMS and attached . The tracked version includes the new response alongside the old, as well as the light edits that I made (again, mostly acronym definitions).

Please let me know if you'd like any further changes to this or the other documents and I can have them ready by the time we need them Monday. Thanks!

View ADAMS P8 Properties ML16246A247 Open ADAMS P8 Document (Appeal of Backfit Imposed in Braidwood and Byron Stations {To: William Dean. From: Victor McCreel}

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0-16E22 From: Mccree, Victor Sent: !Friday, September 09, 2016 4:46 PM To: Clark, Theresa <Theresa.Clark@nrc.gov>; Holahan, Gary <Gary.Holahan@nrc.gov>

Subject:

Re: backfit documents Theresa/Gary, See attached edits to the reclama response. I need a stronger response to question 2.c. Please work either Tom and let me know what you propose.

Thanks, Vic

MEMORANDUM TO: William M. Dean, Director Office of Nuclear Reactor Regulation FROM: Victor M. McCree Executive Director for Operations

SUBJECT:

RESULT OF APPEAL TO THE EXECUTIVE DIRECTOR FOR OPERATIONS OF BACKFIT IMPOSED ON BYRON AND BRAIDWOOD STATIONS REGARDING COMPLIANCE WITH 10 CFR 50.34(b), GDC 15, GDC 21 , GDC 29, AND THE LICENSING BASIS As you are aware, on June 22, 2016, I established a Backfit Appeal Review Panel (Panel) in accordance with Management Directive (MD} 8.4, "Management of Facility-specific Backfitting and Information Collection," to review the subject appeal and to provide me with recommendations (Agencywide Documents Access and Management System (ADAMS}

Accession No. ML16173A311). On August 24, 2016, the Panel transmitted the results of its review to me (ADAMS Accession No. ML16236A202}. The memorandum from the Panel responding to my tasking , recommended that the 2015 compliance backfit be withdrawn, and included the Panel's report and the basis for this recommendation (ADAMS Accession No. ML16236A208}.

I have reviewed the Panel's report, its recommendations , and its responses to the questions I posed when establishing the panel. In addition, I met with you on September 12, 2016, to discuss my decision and assure that it reflects the thorough, technically sound, and legally well-founded consideration that this matter merits. Our discussion included my response to the additional perspectives you provided to me in your email dated September 2, 2016, which is enclosed, for reference.

As we discussed, the central question in the a1:31:3eal p,Eanel's review was whether an adequate basis exists for backfitting using the compliance exception in Title 10 of the Code of Federal Regulations (1 OCFR}, Section 50.109(a)4(i) to address potential pressurizer safety valve failures following water discharge. With regard to compliance, the 1985 statement of considerations for 10 CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact.. .. new or modified interpretations of what constitutes compliance would not fall within the exception ... ." In answering this question, the Panel focused on the following three related technical and regulatory positions for the pressurizer safety valves (PSVs} described in the staff's October 9, 2015, safety evaluation CONTACT: Gary M. Holahan, OEDO (301) 415-1765

W. Dean imposing the back.fit (ADAMS Accession No. ML14225A871, referred to as the Backfit SE), as well as the staff's May 3, 2016, response (ADAMS Accession No. ML16095A204) to the backfit appeal by Exelon Generation Company, LLC (the licensee):

1. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code} water qualification (certification) documentation is required if a PSV is to be assumed to reclose after passing water.
2. Water discharge through a steam-qualified PSV will cause the valve to stick in its fully open position.
3. PSVs are subject to the single-failure criterion.

As the Panel noted in its report, it is important to acknowledge that the PSVs in question were designed for steam service and that water discharge through such valves is undesirable and should be minimized or avoided as a matter of conservative engineering and prudent operations. This perspective is reinforced by several industry positions and testing, as well as operator training and control room procedures intended to terminate a potential pressurizer overfill event before filling the pressurizer~. For these reasons, the staff's position described in the NRC's backfit imposition letter and its response to the backfit appeal, represents a well-intentioned and conservative approach that could provide additional safety margin. However, based on my review of the relevant documents and discussions, I agree with the Panel's conclusions and support its recommendations. In particular, I agree with the Panel's assessment of the three relevant technical and regulatory positions.

First. regarding ASME Code water qualification (or certification), when considered in the context of the Byron and Braidwood licensing basis, valve "qualification* implies a general demonstration of capability, such as through the Electric Power Research Institute testing conducted in response to Three Mile Island (TMI) Action Plan Item 11.D.1, not ASME BPV Code certification. Thus, when preparing the safety evaluations associated with two license amendments in 2001 and 2004 (referred to as the Uprate SE and the Setpoint SE), the NRC staff exercised reasonable and well-informed engineering judgment to conclude that the PSVs were unlikely to stick in the fully open position. The NRC staff's determination that ASME BPV Code certification is necessary for PSVs first appears in the Backfit SE and is not addressed in any of the final NRC requirements or guidance documents reviewed by the Panel. As such, the NRC staff's position on valve qualification in the Backfit SE represents a new or modified interpretation of what constitutes compliance in addressing potential PSV failures following water discharge.

Second, regarding PSV failure following water discharge, the standard in place in 2001 and 2004, and at present, is simply that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-infom,ed staff engineering judgment. Without the presumption of PSV failure to reseat, the concerns in the Backfit SE related to event classification, event escalation, and compliance with 10 CFR 50.34(b) and General Design Criteria 15, 21, and 29 are no longer at issue.

Third, the determination that application of the single failure criterion is necessary first appears in the draft Revision 1 to Regulatory Issue Summary 2005-29. This position, which is still under

W. Dean development, is not included in any final NRC requirement or guidance document reviewed by the panel.

In sum, neither of the three positions were "known and established standards of the Commission" when the NRC issued the Uprate SE and the Setpoint SE in license amendments for Byron and Braidwood in 2001 and 2004, respectively, for determining when it was appropriate to assume a failure of a PSV to reseat. Based on the Panel's review, they were not "known and established standards of the Commission" in 2005 (when RIS 2005-29 was issued),

in 2006 (when the Beaver Valley extended power uprate was approved}, in 2007 (when Revision 2 to Standard Review Plan Sections 15.5.1 -15.5.2 was issued), nor are they "established standards of the Commission" at present.

As a result, I do not support the use of the compliance exception to impose the subject backfit.

agree with the Panel's assessment that the current licensing basis for Byron and Braidwood complies with the applicable regulations and provides adequate protection of public health and safety. I have responded directly to the licensee with my decision on its appeal.

The Panel's report also identifies two issues that warrant further NRC consideration. The report reveals the need ta assess the treatment of the underlying technical issue described in the 1993 Westinghouse Nuclear Safety Advisory Letter {NSAL-93-013) on PSV performance after water discharge at pressurized water reactors. In addition, given the decision communicated herein, the positions included in RIS 2005-29, as well as its proposed Revision 1, should be

{re)assessed through the appropriate generic process to ensure they receive appropriate backfit consideration. You are requested to inform me within 30 days of your plan to respond to these issues.

As you are also aware, I have recently directed the U.S. Nuclear Regulatory Commission (NRC)

Committee to Review Generic Requirements (CRGR) to assess the adequacy and currency of existing NRC requirements, guidance, criteria, procedures, and training on the subject of backfitting (ADAMS Accession No. ML16133A575). The Panel members haves already been in contact with the CRGR to share insights and perspectives from this review. I believe that the CRGR evaluation of our implementation of the backfit process presents us with a timely opportunity to further enhance our regulatory process.

Finally, I recognize that the technical and regulatory positions used in the staff's decision-making involved careful, thorough, and technically solid considerations, reflecting their commitment to ensuring safety. Knowing that our people take seriously the responsibility for assuring public health and safety and are willing to pursue backfits, when appropriate, to assure or enhance safety is key to successfully fulfilling our mission. Although expected, I also sincerely appreciate the cooperation and respect evidenced by both your staff and the Panel members as the Panel evaluated the merits of the licensee's appeal of this technically complex and difficult regulatory issue. Their open, constructive and collegial interactions reflected the

w. Dean best of our agency values and contributed to what I consider to be a sound final regulatory decision.

Enclosure:

As stated cc: Chairman Burns Commissioner Svinicki Commissioner Baran

W. Dean and difficult regulatory issue. Their collegial interactions reflected the best of our agency values and contributed to what I consider to be a sound regulatory decision.

Enclosure:

As stated cc: Chairman Burns Commissioner Svinicki Commissioner Baran DISTRIBUTION: OED0-16-00584 RidsNrrMailCenter RidsNrrDorllpl3-2 AGody, Region II RidsOgcMailCenter RidsNrrPMByron AGendelman, OGC RidsResMailCenter RidsNrrPMBraidwood GMizuno, OGC RidsNroMailCenter RidsNrrDss RCorreia, RES RidsNmssMailCenter RidsNrrDe KSWest, NSIR RidsRgn1 MailCenter RidsNrrDpr MBailey, NSIR RidsRgn2MailCenter RidsNrrDorl TScarbrough, NRO RidsRgn3MailCenter AGarmoe, NRR MASpencer, OGC RidsRgn4MailCenter TKeene, NRR. TClark, OEDO ADAMS Accession No.: ML16246A247 ED0-002 OFFICE OEDO OEDO OEDO NAME GHolahan MJohnson VMcCree DATE 09/ /16 09/ /16 09/ /16 OFFICAL RECORD COPY

RESPONSE BY EXECUTIVE DIRECTOR FOR OPERATIONS (EOO) TO ADDITIONAL PERSPECTIVES ON BACKFIT APPEAL REVIEW PANEL FINDINGS PROVIDED IN SEPTEMBER 2, 2016, EMAIL FROM THE DIRECTOR OF THE OFFICE OF NUCLEAR REACTOR REGULATION (NRR)

1. NRR appreciates the panel's efforts. However, NRR believes that the panel's perspectives do not provide sufficient basis to overturn the backfit.

Response: Based on my review, the Panel's perspectives provide a sound basis for supporting the licensee's appeal of the compliance exception backfit. The concerns listed below do not address the specific Panel finding that is a primary basis for overturning the backfit. In particular, the NRC has previously accepted water qualification of pressurizer safety valves (PSVs) and power-operated relief valves (PORVs) based on Electric Power Research Institute (EPRI), Wyle, or vendor testing for nuclear power plants (beyond Byron and Braidwood) as part of Three Mile Island (TMI) action items, Chapter 15 accident analyses, and other evaluations. In those evaluations, the NRC did not require American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code) certification for water service.

2. NRR Concerns
a. The panel has narrowly focused its review on the water qualification question.

NRR maintains that the original backfit documents numerous issues with the licensing basis for Byron and Braidwood that have not been addressed in the panel's assessment Response: In the report, I find that the Panel adequately addressed the issues identified as important by NRR in its comments on the preliminary findings. Although it is not clear which specific additional issues are of concern to NRR. I find that the most salient positions and issues associated with the compliance exception backfit question have been appropriately considered in this decision.

b. With regard to the PSV water qualification question, the panel's position is reliant on its interpretation of the 1977 Information SECY [SECY-77-439, "Single Failure Criterion," dated August 17, 1977]. The panel has provided select quotes from that SECY that it believes supports its position. NRR believes that when the entire SECY is reviewed it becomes clear that the SECY was simply documenting current practices in 1977, some of which were still being researched, and does not provide a "known and established standard." The staff contends that if the 1977 SECY had been Intended to provide the "known and established standard" it would have been included in subsequent updates to regulations, regulatory guides, and SRPs [Standard Review Plans] over the following nearly 40 years. It has not Response: In its report, the Panel indicates that it addressed SECY-77-439 in response to NRR's assertion that Exelon had not satisfied the "single failure assumption" for the PSVs at Byron and Braidwood. The intent regarding NRR's reference to a "known and established standard" in this comment on SECY-77-439 is not clear. On page 21 , the report states that the Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that failures of PSVs to reclose need not be assumed to occur following water discharge if the likelihood is sufficiently small, ENCLOSURE

based on well-informed staff engineering judgment. This is the "known and established standard" of interest for water qualification of PSVs with respect to the backfit.

c. In numerous places the panel quotes documents that it interprets as describing the treatment of check valves as analogous to PSVs. The panel did not find any definitive documentation that demonstrates that the agency concluded that PSVs are analogous to check valves and, as such, should be considered passive components. This appears to be the panel's judgement, not an NRC position.

NRR disagrees with the panel's interpretation and has historically treated PSVs as active components, including designating them as such during license renewal.

PSVs are designed to perform a specific RCS overpressure protection safety function critical to protecting one of the key defense-in-depth barriers to protect public health and safety from the release of radioactive materials. The staff believes the panel's comparison is inappropriate and establishes a very concerning precedent.

Response: The Panel's discussion of check valves in relation to PSVs was provided for context given that PSVs were not explicitly discussed in documents describing passive failures and the application of the single failure criterion (e.g., SECY-77-439; SECY 084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs,~ dated March 28, 1994; and SECY-05-0138, "Risk-Informed and Performance-Based Alternatives to the Single-Failure Criterion,"

dated August 2, 2005). The Panel did not intend to establish precedent, but rather summarize past practice that could be used to consider PSVs. The Panel explicitly acknowledged that PSVs may be considered "active" for various regulatory applications (e.g., inservice testing). As indicated in the comment, license renewal aging management programs are another regulatory application where active and passive components are distinguished, and PSVs may be considered active for purposes of some such programs.

To apply the "single failure assumption" to the PSVs at Byron and Braidwood as NRR asserts, however, the PSVs would need to have been treated as active components in the accident analysis to which single failures are applied, and the Panel did not find evidence that this has been the case. The principal discussion of potential PSV failures in the accident analyses occurs only in the evaluation of an inadvertent opening of a PSV. In addition, other analyses of overpressure events (e.g, loss of load, loss of feedwater) do not apply the single failure criterion to cause a PSV to stick open (i.e., fail to reseat) when opening on steam flow. In addition, the feedwater system pipe break analysis does not apply the single failure criterion to cause a PSV to stick open either during steam discharge or during water discharge. Furthermore, the Panel reviewed documentation for several other Westinghouse-designed plants showing consistent treatment of the PSVs (i.e., without single failure assumptions) in the accident analyses.

d. On page 13, the panel acknowledges the Byron/Braidwood licensing basis as categorizing the PSVs and PORVs as active components. However, the panel, given its reliance on treating PSVs akin to check valves, establishes a new and different position in its own summary when it determines these valves should be treated as passive components for the purposes of considering the single failure criterion.

Response: In this section and in Section 3.2 of the report, the Panel indicates that the failure of a "simple check valve" is most similar, from a mechanical perspective, to the PSV failure addressed in the October 9, 2015, Backfit SE. The discussion also indicates that PSVs (like check valves) could be considered passive or active depending on the specific evaluation, such as for design, accident analysis, or inservice testing.

The report includes the following quote from SECY-94-084: "... the NRC staff normally treats check valves, except those in containment isolation systems. as passive devices during transients or design-basis accidents: Thus, the Panel is not establishing a new position, but rather summarizing past NRC practice.

e. Regarding ASME, [Title 10 of the Code of Federal Regulations (10 CFR), Section]

50.55a requires nuclear power plants to be initially designed and constructed [in accordance with) ASME [BPV Code], Section Ill and to be tested throughout their service life [in accordance with] ASME OM Code [Operation and Maintenance of Nuclear Power Plants]. These codes comprise the qualification standards for ASME Class 1 safety valves such as the pressurizer PSVs with which licensees are required to comply unless alternatives have been authorized by the staff [in accordance with] 10 CFR 50.55a.

Response: ASME BPV Code does indeed provide certification requirements for safety and relief valves for their intended design function. However, as noted in the Panel's report, since the TMl-2 accident, the NRC has accepted qualification of safety and relief valves based on EPRI, Wyle, and vendor testing to demonstrate that these valves will not stick open during water discharge as part of TMI action items, Chapter 15 accident analyses, or other evaluations for numerous nuclear power plants. Therefore, a longstanding NRC precedent for the acceptance of water qualification of safety and relief valves through such testing has been established in those evaluations. Based on infonnation provided by NRR and a sampling review by the Panel, the NRG has not required safety or relief valves to be certified by the ASME BPV Code for water service when referenced Chapter 15 accident analyses. A requirement at this time that all safety and relief valves be certified for water service in accordance with the ASME BPV Code for every reference to safety and relief valves not sticking open upon water discharge for all TMI action items, Chapter 15 accident analyses, and other evaluations would constitute a significant regulatory action and warrant a different decision-making process

f. The panel asserts in its summary that the valves in question were water qualified due to the licensee's reliance on them to pass water during feedline break events.

The panel does not appear to acknowledge that feedline breaks are Condition IV events, similar to [loss-of-coolant accidents], which are never expected to occur in the lifetime of the facilities and therefore, given their lower probability of occurrence, are permitted to have more significant consequences. The EPRI testing demonstrated acceptable performance under conditions anticipated during these Condition IV events (higher temperature fluid - 650°F), while the EPRI test at the more likely Condition II inadvertent mass addition event conditions (lower temperature fluid -sso*F) was terminated early due to valve chatter on opening. The summary of the EPRI testing indicated that for subcooled water conditions valve chatter and resultant valve damage was generally observed.

Response: In evaluating information provided by NRR associated with water testing, NRR maintained that the EPRI testing did not address water discharge for the Byron and Braidwood PSVs. The current question recognizes the EPRI testing, but asserts that the water was not at an acceptable temperature, or the water relief might result in valve chatter or damage. Based on the Panel's document review and discussions, the NRC staff approval of the amendments during the 2001 and 2004 reviews evaluated the EPRI testing, and were indeed aware of the test, its results and water temperatures.

3. Path Forward
a. If the EDO supports the original backfit, NRR agrees with the panel that risk insights are important considerations in determining how reasonable assurance of compliance can be demonstrated. However, as acknowledged by the panel, consistent with [Regulatory Guide (RG)] 1.174, risk insights must include consideration of defense-in-depth and safety margins. If a PSV were to stick open or significantly leak at Bryon and Braidwood during a licensing basis Condition II event, which is anticipated to occur on an annual frequency, the licensee has not yet demonstrated adequate defense-in-depth. NRR is open to considering risk-informed licensing basis changes, or potential plant modifications, that appropriately consider all 5 elements of RG 1.174.

Response: I support the recommendations of the Panel. However, I agree that the safety significance of the potential for the PSVs to stick open should be considered as part of a generic resolution of this issue for all pressurized-water reactors.

b. If the ECO supports the Backfit Panel's conclusion, NRR requests that the EDO allow the staff to independently assess what path forward is appropriate given the positions documented in the panel's report and EDO's decision. In particular, NRR has concerns regarding the recommendations on page 3 of the report that need to be further considered before determining what future course of action is most appropriate.

Response: I agree. The report reveals the need to assess the treatment of the underlying technical issue described in the 1993 Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013) on PSV performance after water discharge at pressurized-water reactors. In addition, given the decision communicated herein, the positions included in Regulatory Issue Summary 2005-29, as well as its proposed Revision 1, should be (re)assessed through the appropriate generic process to ensure they receive appropriate backfit consideration. The Director of NRR should inform me within 30 days of the plan to respond to these issues.

From: McCree, Victor Sent: Sunday, September 11, 2016 7:06 PM To: Holahan, Gary; Johnson, Michael; Tracy, Glenn; Clark, Theresa

Subject:

RE: Exelon Backfit Appeal key messages Attachments: EDO EXELON BACKFIT APPEAL DECISION - Comm Message Map.docx See the attached message map (attached) for use in communicating the subject decision internally.

[Gary] note that I edited the key message for clarity and succinctness.

Vic From: Holahan, Gary Sent: Friday, September 09, 2016 1:19 PM To: Mccree, Victor <Victor.McCree@nrc.gov>

Cc: Johnson, Michael <Michael.Johnson@nrc.gov>; Tracy, Glenn <Glenn.Tracy@nrc.gov>; Clark, Theresa

<Theresa.Clark@nrc.gov>

Subject:

Exelon Backfit Appeal key messages

Vic, Some thoughts on messages associated with the Exelon backfit appeal decision ...

Facts:

The staff issued a compliance backfit letter to Exelon (October 9 , 2015) on the issue of pressurized overfill and safety valve performance during Condition II events (ANS Condition II categorization as frequent events).

Exelon twice appealed the compliance backfit (once at the office level, then at the EDO level) as inappropriate since, in their view, the staff failed to identify any error or omission that make the previously approved analysis incorrect.

The EDO established a panel of senior staff and managers to review the Exelon backfit appeal.

The backfit appeal panel recommended , and the EDO supported, a reversal of the compliance backfit, agreeing with the Exelon appeal.

Key Messages:

The staff takes its responsibility for assuring safety very seriously; and pursues backfits when it concludes they are appropriate to assure or enhance safety. The staff endeavors to perform thorough, technically sound, and legally well.founded reviews in all cases.

On complex technical and legal matters there can be differing views either within the staff, or with licensees and other stakeholders. The staff considers resolution of comments and alternative views, including appeals, important and deserving of serious attention. The backfit panel and the EDO have given this issue the expected, thorough and serious attention.

In this case the Panel concluded, and the EDO agreed, that the NRC staffs position on valve qualification in the Backfit safety evaluation is a new or modified interpretation of what constitutes compliance in addressing potential Pressurizer Safety Valve failures following water discharge. Although this new staff position

represents a well-intentioned and conservative approach that could provide additional safety margin, it does not provide a basis for a compliance backfit.

Consistent with its commitment to be an effective, efficient, and predictable regulator; and in a proactive approach to important safety and legal matters, on June 2, 2016 the EDO tasked the Committee to Review Generic Requirements (CRGR) with undertaking a review of NRC implementation of agency backfitting and finality guidance. This effort includes an assessment of the clarity and effectiveness of backfitting requirements, guidance and criteria. It also includes an assessment of staff training and knowledge management.

The staff holds itself to high technical, safety, and legal standards and is committed to excellence in fulfilling its responsibilities. The staff is also committed to learn from experience, just as it expects of its licensees.

Gary 2

From: Clark, Theresa Sent: Friday, September 09, 2016 5:11 PM To: Scarbrough, Thomas Cc: Holahan, Gary

Subject:

REQUEST: backfit response item

Tom, We are in the process of responding to NRR on the backfit, and Vic is using some of the responses that you prepared to NRR's comments. For one item, he would like a stronger response. I have attempted below to write something (Updated Respon se), but would like your check/insights. Please respond when you have a chance and I will send back to Vic. We would like to get this memo finalized on Monday if possible (I' ll probably be working on it over the weekend but am not asking you to). Thank you so much! !
c. In numerous places the panel quotes documents that it interprets as describing the treatment of check valves as analogous to PSVs. The panel did not find any definitive documentation that demonstrates that the agency concluded that PSVs are analogous to check valves and, as such, should be considered passive components. This appears to be the panel's judgement, not an NRC position. NRR disagrees with the panel's interpretation and has historically treated PSVs as active components, including designating them as such during license renewal. PSVs are designed to perform a specific RCS overpressure protection safety function critical to protecting one of the key defense-in-depth barriers to protect public health and safety from the release of radioactive materials. The staff believes the panel's comparison is inappropriate and establishes a very concerning precedent.

Previous Response: The Panel's provided a discussion of PSVs in comparison to check valves referenced in SECY-77-439 to provide context for the discussion of the NRR assertion that Exelon did not satisfy the "single failure assumptionn for the PSVs in Byron and Braidwood. The Panel indicates that components, such as PSVs or check valves, may be treated as passive or active depending on the specific evaluation.

Updated Response: The Panel's discussion of check valves in relation to PSVs was provided for context given that documents describing passive failures and the application of the single failure criterion (e.g., SECY-77-439, SECY-94-084, and SECY-05-0138) did not provide explicit discussion of PSVs. The Panel did not intend to establish precedent, but rather summarize past practice that could be used to consider PSVs. The Panel explicitly acknowledged that PSVs may be considered "activen for various regulatory applications, including inservice testing and the definition of license renewal aging management programs. To, apply the "single failure assumption" to the PSVs at Byron and Braidwood as NRR asserts, the PSVs would need to have been treated as active components in the accident analysis to which single failures are applied, and the Panel did not find evidence that this had been the case.

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0 -16E22 1

From: McCree, Victor Sent: Friday, September 09, 2016 11:02 AM To: Clark, Theresa; Johnson, Michael Cc: Holahan, Gary; Lewis, Robert; Tracy, Glenn

Subject:

Re: backfit documents Ok, got it, thanks Theresa!

Vic On: 09 September 2016 10:48, "Clark, Theresa" <Theresa.Clark@nrc.gov> wrote:

Vic, Per your request, attached are the current Word versions of the three documents we are working related to the backfit decision. ADAMS links are also provided below for completeness. I would be happy t o assist in incor oratin our changes into the paper/ADAMS packages. Feel free to email or call me at 301-415-4048 (office) or (b)(6l (cell).

Please note that I am also working three related matters (on which you may have views):

  • Need for NLO on the letter to Exelon - per chat with Gary, I have a question in to Margie (who is in a meeting currently) to get her view
  • Whether the memo to NRR should be public - my recommendation is yes, since there were panel recommendations regarding NIRR activities and there could be questions
  • Coordination of a press release (if we do one) as there*are several process steps needed if it were to be released concurrently M ike-FYI, as you concurred in an earlier version of the letter to Exelon.

Each of these records is publicly

  • Letter responding to Exelon available in ADAMS under t he View ADAMS P8 Properties ML16243A067 specified accession number.

Open ADAMS P8 Document (09/XX/16 Letter to Exelon from Victor M ccree)

  • Letter responding to NEI (which had sent a letter in support of Exelon)

View ADAMS P8 Properties ML16246A150 Open ADAMS P8 Document (09/XX/16 NEI Comments in Support of Exelon Generation Company Second Level Appeal (To: Anthony Pietrangelo. From: Victor M ccree))

From: Victor McCree})

Thanks!

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@nrc.gov I 301-415-4048 I 0 -16E22 1

From: Rihm, Roger Sent: Wednesday, August 31, 2016 2:47 PM To: Holahan, Gary

Subject:

Exelon Backfit One Pager Attachments: NRR_ Exelon Backfit Appeal.docx Importance: High Here you go - thank you!

1

Exelon Appeal of Compliance Backfit Key Messages

  • NRC imposed a backfit using the compliance exception on Exelon's Braidwood and Byron plants by letter dated October 9, 2015. Exelon appealed the compliance backfit claiming it did not meet the requirements of the compliance exception and a backfit analysis should be conducted.
  • Something about the appeal Facts
  • The Backfit Rule for power reactors allows the imposition of new regulatory requirements after prior NRC approval (e.g., issuance of a license), if an analysis is prepared demonstrating that the backfit involves a substantial increase in protection to safety or security and that the costs are justified by this increase in protection.
  • However, when the NRC demonstrates in a documented evaluation that a proposed backfit involves adequate protection or compliance with an established NRC requirement or licensee commitment, the NRC does not need to prepare a backfit analysis. The compliance exception can be used when NRC approved something that should not have been approved as a result of omitted information or a mistake of fact.

Compliance Backfit Imposed on the Braidwood and Byron Plants

  • Certain events can occur that result in the reactor coolant system over-filling with water, which pushes pressurizer safety valves (PSVs) open. Once the excess water addition is stopped, the PSVs must re-close to prevent an uncontrolled leak.
  • Failure of a PSV to reclose would result in a Condition II event becoming a Condition Ill event (small break LOCA), which is prohibited by the Braidwood and Byron UFSARs.
  • During review of a recent power uprate request, the NRG staff determined that Exelon did not provide sufficient information to shaw that the PSVs would re-close after relieving water and, thus, compliance with the UFSAR
  • In 2001 and 2004 the staff approved license amendment requests that predicted the PSVs would relieve water and re-close, even though the staff determined in hindsight that Exelon did not, at the time, provide sufficient information to show that the PSVs would re-close.
  • The staff issued a backfit in October since Exelon's safety analyses predict the PSVs will relieve water in certain events and Exelon had not shown that the PSVs would re-close to prevent escalation from a Condition II to Condition Ill event.

Exelon's Appeal of the Compliance Backfit

  • Exelon appealed the staff's use of the compliance exception to the backfit rule stating the NRC needed to perform a backfit analysis because the NRG had not shown the 2001 and 2004 approvals to be a result of omission or a mistake of fact.
  • The NRR Office Director appointed an independent three member backfit review panel to review Exelon's backfit appeal. The panel reviewed documentation, interviewed staff, and held a public meeting with the licensee. The NRR Office Director accepted the panel's recommendation to deny the appeal, which was communicated to the licensee by letter dated May 3. 2016.
  • The licensee then appealed the decision to the EDO and a second independent review panel was created, which recommended granting the appeal.
  • While non-escalation from a Category II to a Category Ill was acknowledged to be a known and established standard, and part of the Braidwood and Byron licensing basis, the EDO panel did not believe there was a known and established standard for qualification of pressurizer safety valves that relieve water.

From: West, Steven Sent: Thursday, August 25, 2016 5:44 PM To: Holahan, Gary; Clark, Theresa Cc: Hackett, Edwin

Subject:

RE: Exelon Backfit Appeal Panel Report In casual conversation with Brian McDermott today, he indicated that he doesn't expect NRR to have any comments on the report. He said NRR's focus is on understanding why NRR's answer differed from the review panel's answer (I gave him my views) and how to avoid this again in the future.

Do either of you see any way that Vic will issue his decision before the CRGR public meeting on September 131h? Having the report (or at least the decision) publically available before the meeting could be a game changer for the meeting, especially if Vic accepts our recommendations. Allowing the standard 2 weeks for OGC review could make it do*able, but very tight. Given the decision, I almost think a meeting on September 13th would be a waste of time if we can't talk about the Exelon appeal and the generic insights it provided as a case study.

During a CRGR business meeting this afternoon, Len Wert wondered if the agency should revisit the Hatch appeal decision in light of the outcome of the Exelon appeal decision. I told him that, while the appeal review panel did not do the same deep dive on Hatch, like Byron and Braidwood, we did review the Hatch decision and convinced ourselves that the Byron and Braidwood decision should not change as a result of the Hatch decision.

Steve Steven West, Deputy Director Office of Nuclear Security and Incident Response U.S. Nuclear Regulatory Commission 301-287-3734 Steven.West@nrc.gov From: Holahan, Gary Sent: Wednesday, August 24, 2016 12:49 PM To: Dean, Bill <Bill.Dean@nrc.gov>; McDermott, Brian <Brian.McDermott@nrc.gov>; Evans, Michele

<Michele.Evans@nrc.gov>; McGinty, Tim <Tim.McGinty@nrc.gov>; Ll!lbinski, John <John.Lubinski@nrc.gov>; Mccree, Victor <Victor.McCree@nrc.gov>; Johnson, Michael <Michael.Johnson@nrc.gov>; West, Steven

<Steven.West@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>; Scarbrough, Thomas

<Thomas.Scarbrough@nrc.gov>; Spencer, Michael <Michael.Spencer@nrc.gov>

Subject:

FW: Exelon Backfit Appeal Panel Report

... full ADAMS Accession numbers Package: ML16236A198 These records are publicly available in ADAMS.

Memo: ML16236A202; Enclosure ML16236A208

From: Holahan, Gary Sent: Wednesday, August 24, 2016 12:31 PM To; Dean, Bill <Bill.Dean@nrc.gov>; McDermott, Brian <Brian.McDermott@nrc.gov>; Evans, Michele

<Michele.Evans@nrc.gov>; McGinty, Tim <Tim.McGinty@nrc.gov>; Lubinski, John <John.Lubinski@nrc.gov>

Cc: Mccree, Victor <Victor.McCree@nrc.gov>; Johnson, Michael <Michael.Johnson@nrc.gov>; West, Steven

<Steven.West@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>; Scarbrough, Thomas

<Thomas.Scarbrough@nrc.gov>; Spencer, Michael <Michael.Spencer@nrc.gov>

Subject:

Exelon Backfit Appeal Panel Report

NRR, The Exelon backfit appeal panel delivered its report to the EDO and DEDO this morning (ML16236A202 and ML 16236A20). The panel reviewed the NRR response to the panel's preliminal)' findings, but could not agree with the NRR positions. The report therefore recommends to the EDO that he support the Exelon appeal. The report will be distributed today at the EDO's request.

The EDO will make his final decision after studying the report and considering any feedback from NRR and other stakeholders.

The panel is available to discuss the report with you and respond to your questions, Gary 2

From: West, Steven Sent: Tuesday, August 23, 2016 7:34 AM To: Spencer, Michael; Holahan, Gary; Scarbrough, Thomas; Clark, Theresa

Subject:

RE: Backfit Appeal Panel Report (MASTER) - 2016-08-22 R2 - MAS Attachments: Backfit Appeal Panel Report (MASTER) - 2016-08-22 R2 - MAS Steve's comments.docx

Theresa, In this version , I've responded to some of the previous comments and made a substantive (and I think simplifying) change to our response to Question 4.

Steve Steven West, Deputy Director Office of Nuclear Security and Incident Response U.S. Nuclear Regulatory Commission 301-287-3734 Steven.West@nrc.gov From: Spencer, Michael Sent: Monday, August 22, 2016 6:19 PM To: Holahan, Gary <Gary.Holahan@nrc.gov>; West, Steven <Steven.West@nrc.gov>; Scarbrough, Thomas

<Thomas.Scarbrough@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

Backfit Appeal Panel Report (MASTER) - 2016-08-22 R2 - MAS All, attached are my comments on the* document Theresa emailed out 20 minutes ago. I incorporated all of Steve's/Tom's/Theresa's edits, so any edits in the attached are mine.

Michael

Report of the Backf it Appeal Review Panel Chartered by the Execu tive Direc tor for Opera tions to Evaluate the June 2016 Exelo n Backf it Appeal Gary M. Holahan K. Steven West Thomas G. Scarbro ugh Michael A. Spence r Theresa V. Clark August XX, 2016 ADAMS Accession No. MLXXXXXXXXX

TABLE OF CONTENTS 1 Background .................................................................................................................. 1 1.1 Conduct of the Panel's Review ............................................................................. .......... 2 1.2 Proposed Compliance Backfit and Exelon Appeals ........................................................ 3 1.3 Backfit Rule and the Compliance Exception ................................................................... 5 1.4 A Brief History of Pressurizer Valve Issues ................................................................... I6 1.5 History and Review of Westinghouse NSAL and Related Activities ................................ 8 2 Summary of the Appeal Review Panel Findings ........................................................ 8 3 Discussion .................................................................................................................... 9 3.1 General Design Criteria (1971) .................................................................................... 1OQ.

3.2 Commission Paper on Single Failure ( 1977)................................................................ 109 3.3 TMI Action Plan Item 11.D.1 (1980) ................................................................................ 10 3.4 NRC Closure of TMI Action Plan Item 11.D.1 for Byron and Braidwood (1988-1990) ..... 11 3.5 Westinghouse NSAL-93-013 and Supplement 1 (1993-1 994) ....................................1144 3.6 Commission Paper on Passive Plant Designs (1994) ................................. .................. 12 3.7 Draft Standard Review Plan Revision (1996) .................................................... ............ 13 3.8 Power Uprate Reviews and License Amendments (2001-2006) ................................ 144-3 3.9 RIS 2005-29 (2005) and Proposed Draft Revision 1 to RIS 2005-29 (2015) .............. 15~

3.10 SECY-05-0138 (2005) .................................................................................................. 15 3.11 Standard Review Plan Revision (2007) ......................................................................1§:+a 3.12 Backfit Letter and Subsequent Backfit Appeals (2015-2016) ........................................ 16 4 Response to the EDO Questions ............................................................................2049 4.1 Were the approvals based on a mistake? If so, what was the mistake and what are the implications for Braidwood and Byron? ................ ...................................................... 20+9 4.2 What is the known and established standard for water qualification of PSVs? ............ £1~

4. 3 What is the known and established standard for progression of postulated events between categories of severity? ................................................................................ 21~

4.4 Does the current licensing basis for Braidwood and Byron comply with the applicable regulations? Is it adequate to provide protection to public health and safety? .......... .... 21 4.5 Given that Exelon suggests that the NRC pursue a cost-justified substantial safety enhancement backfit, what is the contribution to overall plant risk of the current configuration at Braidwood and Byron? ............... ......................................................... 21 5 Summary and Conclusions ....................................................................................2224 6 Additional Panel Thoughts .....................................................................................242-3 Appendix A: History of the Backfit Rule and the Compliance Exception ....................... 262i Appendix B: Qualification of Pressure Relief Valves in Nuclear Power Plants in Response to the TMl-2 Accident .............................................................................................. 3130 Appendix C: Concerns regarding Performance of Pressurizer Valves under Water Flow Conditions................................................................................................................3736 Appendix D: References .....................................................................................................484+

Appendix E: List of Abbreviations ..................................................................................... 586+

-i -

1 BACKGROUND On June 22, 2016, 1 in accordance with NRC Management Directive (MD) 8.4,2 the NRC Executive Director for Operations (EDO) established a Backfit Appeal Review Panel (Panel) to review the appeal by Exelon Generation Company, LLC (Exelon or the licensee) of the U.S.

Nuclear Regulatory Commission (NRC) staffs determination that a backfrt is necessary at Byron Station, Units 1 and 2 (Byron) and Braidwood Station, Units 1 and 2, as well as the NRC staffs application of the compliance backfit exception provided in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, "Backfitting."

This backfit determination is documented in an October 9, 2015, letter (referred to as the Backfit Letter). 3 The letter describes the NRC staff's review of licensing basis documents for Byron and Braidwood. The NRC staff determined that Byron and Braidwood were not in compliance with the plant-specific design bases and several NRC regulations:

  • GDC 21, "Protection system reliability and testability"
  • Paragraph (b) of 10 CFR 50.34, "Contents of applications; technical information" Specifically, the NRC staff determined that Byron and Braidwood do not comply with provisions in American Nuclear Society (ANS) Standard 51.1/N18.2-1 9734 for ensuring that ANS Condition II events5 do not progress to more serious ANS Condition Ill events following water discharge6 through certain valves. The NRC staff acknowledged that the NRC staff position differed from a previous staff position documented in a May 4, 2001, safety evaluation (SE) supporting a stretch power uprate (referred to as the Uprate SE). 7 However, the NRC staff determined that the backfitting was justified under the compliance exception in 10 CFR 50.109(a)(4)(i). The NRC staff directed the licensee to take action to resolve the non-compliance.

On December 8, 2015, the licensee appealed the NRC staffs decision to the Director of the Office of Nuclear Reactor Regulation (NRR), stating its disagreement with the NRC's conclusion that the compliance exception to the backfit rule applied in this case, while noting that the NRC 1

NRC 2016e (Author and year citations in footnotes refer to the designation of references in Appendix D to this report.)

2 NRC 2013 3

NRC 2015b - referred to as the Backfit Letter in the remainder of the report 4

ANS 1973 5

Specifically, inadvertent operation of the emergency core cooling system, malfunction of the chemical and volume control system, and inadvertent opening of a pressurizer safety or relief valve.

6 For consistency in this report, the Panel uses the phrase "water discharge" rather than "water relief' or "liquid discharge" (except in direct quotes), as this is the phrase used in the Westinghouse documents that raised the issue addressed in this report.

7 NRC 2001 b - referred to as the Uprate SE in the remainder of the report staff had twice approved the underlying analysis. 8 The approvals referenced by the licensee were an August 26, 2004, license amendment associated with pressurizer safety valve (PSV) setpoints9 and the above-referenced Uprate SE. In a letter dated May 3, 2016, the NRC responded to the licensee's appeal and reaffirmed its decision that the backfit per the compliance exception provisions of 10 CFR 50.109(a)(4){i) is appropriate. 10 On June 2, 2016, the licensee again appealed the NRG staffs decision, this time to the ED0. 11 The purpose of this report by the Backfit Appeal Review Panel is to provide information and recommendations to support the EDO's decision on the appeal.

1.1 Conduct of the Panel's Review In order to establish a technically sound, well informed, and legally defensible basis for its recommendations, the Backfit Appeal Review Panel undertook a review of the relevant documents in this case. This included the licensee and NRG staff letters mentioned above; the Uprate SE and the Setpoint SE; and a June 16, 2016, letter from the Nuclear Energy Institute (NEl)12 supporting the EDO Appeal. The Panel also reviewed many other related documents, which fall into five broad categories:

  • Docketed communications for Byron and Braidwood from 1982 to the present, including license amendment requests (LARs) by the licensee, NRG-issued license amendments, NRC requests for additional information (RAls), licensee responses, meeting summaries, NRC SEs, and the licensee's Updated Final Safety Analysis Report (UFSAR}13
  • NRG guidance relevant to the analysis of inadvertent operation of the emergency core cooling system (IOECCS) events over the period of 1981 to the present, including Standard Review Plan (SRP) Section 15.0, Sections 15.5.1 - 15.5.2, and Section 15.6.1 14
  • Westinghouse Nuclear Safety Advisory Letter (NSAL) 93-01315 and its Supplement 116 ,

as well as docketed communications regarding actions taken by other licensees in response to Westinghouse NSAL-93-013

  • The history of NRC and industry activities related to power operated relief valves (PORVs), their block valves, and PSVs (including Three Mile Island (TMI) Action Plan s Exelon 2015 - referred to as the NRR Appeal in the remainder of the report 9 NRG 2004b - referred to as the Setpoint SE in the remainder of the report 10 NRG 2016d - referred to as NRR Appeal Decision in the remainder of the report 11 Exelon 2016a - referred to as EDO Appeal in the remainder of the report 12 NEI 2016 13 Exelon 2002 and Exelon 201 4 (The Panel reviewed other revisions as well, but they are not included in Appendix Das they are not referenced in this report.)

14 NRG 1981 a, NRC 1981b, NRG 1981c, NRC 2007a, NRC 2007b, and NRC 2007c 1s Westinghouse 1993 1s Westinghouse 1994 Items 11.D.1, 11.0.3, 11.G.1, and 11.K.3 as documented in NUREG-073717, as well as Generic Letter 89-1018 and its supplements), Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREG/CR-703719)

In addition to the document review, the Panel had the benefit of meetings with NRR (both the Division of Safety Systems and the Division of Engineering), the Office of the General Counsel, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon (Bradley Fewell, Senior Vice President of Regulatory Affairs) and NEI (Tony Pietrangelo, Senior Vice President and Chief Nuclear Officer) declined offers for a public meeting, but indicated a willingness to provide information if the Panel identified the need. The Panel did not Identify a need for additional information from either Exelon or NEI to complete the review documented in this report.

At the request of the Panel, the Office of Nuclear Regulatory Research (RES) conducted risk analyses using the NRC's Standardized Plant Analysis Risk model for Byron Unit 1.20 These analyses informed the Panel's response to the question from the EDO regarding the risk significance of the relevant accident sequences.

[Given (CT1Jcsw2Jthat the Backfit Rule creates a structured process for changes to previous NRG staff positions-In effect, placing the burden of proof on the NRG staff-the Panel determined that this level of historical review and staff interaction was necessary to provide context for consideration of the validity of the backfit.

1.2 Proposed Compliance Backfit and Exelon Appeals In the Backfit Letter, the NRC staff informed Exelon that it had determined that Byron and Braidwood are not in compliance with GDGs 15, 21, and 29; 10 GFR 50.34(b); and the plant-specific design bases that were expected to demonstrate there will be no progression of ANS Condition II events to ANS Condition Ill events. The NRC staff stated that based on its review of Byron and Braidwood UFSAR Sections 15.5.1, 15.5.2, and 15.6.1, the UFSAR predicts water discharge through a valve that is not "qualified" for water discharge. Therefore, the NRG staff concluded that the UFSAR does not contain analyses that demonstrate that the plants' structures, systems, and components (SSGs) meet the design criteria for ANS Condition II events as stated In Byron and Braidwood UFSAR Section 15.0.1.2. Based on the SE attached to its letter,21 the NRG staff found that the licensee must take action to resolve the non-compliance.

The Backfit SE addressed three accident analyses in Chapter 15 of the Byron and Braidwood UFSAR: (1) IOECGS; (2) chemical and volume control system (CVCS) malfunction that increases reactor coolant inventory; and (3) inadvertent opening of a pressurizer safety or relief valve (IOPORV). The NRG staff noted that each ANS Condition II event must be shown to meet the following:

17 NRC 1980c- referred to as the TMI Action Plan In the remainder of the report; lessons learned from TMI were also presented in NUREG-0578 (NRC 1979a), NUREG-0585 (NRC 1979b), and NUREG-0660 (NRC 1980a) 18 NRC 1989 19 NRC 2011 20 NRC 2016f 21 Referred to as the Backfit SE in the remainder of the report.

1. no fuel damage,
2. no overpressure of the reactor coolant system (RCS) or main steam system, and
3. no progression into an event of a more serious category without another independent fault.

Regarding an IOECCS, the NRC staff stated in Section 3.1.2.1 of the Backfit SE that use of the block valve to isolate a stuck-open PORV was unacceptable. The NRC staff stated that Westinghous~ [1cr3i sw41recommended this approach in 1993, and that the NRC staff rejected this approach in 2005 (RIS 2005-2922).

In Section 3.1.2.4 of the Backfit SE, the NRC staff stated that the Byron and Braidwood IOECCS analysis depended on water discharge through the PSVs. The NRC staff faulted the licensee for "not appl[ying] the single-failure assumption" and stated that the following information was necessary to support water qualification of the PSVs:

1. In accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), Section Ill, provide the original Overpressure Protection Report defining operating conditions and required relief capacities, and manufacturer's certification and test results
2. In accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), provide inservice test history for PSVs, including water and steam tests, or provide correlation test for alternative test fluid.

Regarding a eves malfunction, the NRC staff stated in Section 3.2 of the Backfit SE that the licensee had not provided an analysis for the eves malfunction that increases reactor coolant inventory that demonstrated the plants' ability to meet the requirements of an ANS Condition II event.

Regarding an IOPORV, the NRC staff stated in Section 3 .3 of the Backfit SE that the licensee had not provided an analysis for the IOPORV that extends long enough into the transient to demonstrate the event would not transition from an ANS Condition II event to an ANS Condition Ill event.

In the Backfit SE, the NRC staff referenced Millstone23 and Callaway24 license amendments as examples of licensees upgrading PORVs for water discharge; a Beaver Valley extended power uprate (EPU) license amendment25 as an example of qualifying PORVs for water discharge; and Turkey Point26 and St. Lucie Unit 227 EPU amendments as additional precedent in support of the backfit decision.

22 NRC 2005b 23 NRC 1998 24 NRC 2000 25 NRC 2006 26 NRC 2012a 27 NRC 2012b In the NRR Appeal, Exelon asserted that the NRC had not justified invoking the compliance exception to the backfit rule. Exelon stated that the NRC approved its IOECCS analysis in both the Uprate SE and the Setpoint SE.

!In ~crsnMAS6Jthe !NRR Appeal Decisiori[SW7], the NRC staff stated that the previous NRC approvals in 2001 and 2004 were inconsistent with the Agency's general position on the known and established standard at issue- in this case, the progression of ANS Condition II events to higher level events. The NRC staff stated that the fact that the NRC staff were aware of references to EPRI reports on the ability of these non-water qualified PSVs to reseat in certain circumstances was not sufficient to support the licensee's position on the compliance backfit.

In the EDO Appeal, Exelon stated that the NRC had misidentified the "known and established standard" at issue as the prohibition of ANS Condition II events progressing to ANS Condition Ill events. Exelon asserted that the standard in question concerns what is necessary to ,,qualify

valves for water discharge. Exelon contended that this standard was the EPRI testing and analysis, and that the NRC agreed that Byron and Braidwood met this standard. Exelon also contended that the change in NRC staff position on prior approvals was not a mistake of fact, but rather a new or modified interpretation of compliance with NRC requirements, for which use of the compliance exception provided for in the Backfit Rule was not appropriate.

1.3 Backfit Rule and the Compliance Exception Backfltting is defined by 10 CFR 50.109(a) as:

... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position Interpreting the Commission's regulations that is either new or different from a previously applicable staff position ... !.1crairMAS9HSW1o1 Unless one of three specified exceptions apply, the NRC may impose a backfit only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50. 109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection."

Section 50.109(a)(4) sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2) and (a)(3). The first exception, the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee." The second and third exceptions relate to actions necessary to ensure adequate protection or to actions that involve defining or redefining adequate protection.

The Commission explained its intended application of the compliance exception in the Statements of Consideration (SOC) accompanying the 1985 final rule amending 10 CFR 50.109:28 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.

In the same SOC, the Commission acknowledged that staff interpretations of rules are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."20 By its terms, the compliance exception applies to actions necessary for compliance with rules, licenses, and orders, or for conformance with written connmitrnents. 30 Also, the Commission explicitly acknowledged the importance of staff interpretations of rules in the regulatory process.

Thus, the Panel understands the term "known and established standard" to include standards established in rules, licenses, orders, and written commitments, and NRC interpretations of rules. Some standards may be broad-based, while others may apply only to a limited number of plants. As stated in NUREG-1409, "[i]nformal or formal communications to one licensee are not official positions to all licensees .... Orders, licenses, and written commitments are applicable only to a particular licensee."

The failure to meet a known and established standard is grounds for a compliance backfit if this failure is due to "omission or mistake of fact." Thus, if a licensee obtains NRC approval of an alternative to a specific standard set forth in guidance, that standard and guidance could not be used to support a compliance backfit unless the NRC's approval of the alternative was based on an omission or mistake of fact. "Known and established standards" are to be distinguished from "new or modified interpretations of what constitutes compliance," which do not fall within the compliance exception. The Panel understands the term "new or modified interpretations" to include situations where the, NRC staff has, in effect, "changed its mind" on how to interpret the language of a requirement or on how much assurance is necessary to conclude that the requirement is met. Levels of assurance might be established in terms such as acceptable probabilities or consequences, conservative assumptions, or sufficient margin.

Additional background information on the Backfit Rule and the compliance exception is provided in Appendix A to this report.

2s NRC 1985, at 38103 20 NRC 1985, at 38102. The 1985 backfit rule was vacated by a Federal court on grounds unrelated to the compliance backfit exception. See Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 824 F .2d 108, 119-20 (1987). In 1988, the Commission amended the backfit rule (NRC 1988b) to address

, the court's concerns, but did not change the 1985 rule's compliance exception provision. Thus, the quoted statements from the 1985 rule are the applicable expression of Commission intent regarding compliance backfits.

30 NUREG-1409 (NRC 1990c) defines written commitments broadly to include the "final safety analysis report, licensee event reports, and docketed correspondence, including responses to NRC bulletins, generic letters, inspection reports, or notices of violation and confirmatory action letters."

1.4 A Brief History of Pressurizer Valve Issues Appendix B to this report provides a summary of the NRC and industry's testing, evaluation, and other consideration of PORVs and PSVs since the TMI Unit 2 (TMl-2) accident in 1979. This historical review provides context for discussion of valve "qualification" in the Backfit SE. It also provides the basis for the Panel's conclusions regarding the "known and established standard" for "qualification" in the context of TMI Action Plan Item 11.D.1 and subsequent activities, as well as how it should be interpreted in the Byron and Braidwood licensing basis.

In light of the NRC staff's assertion that the licensee had not applied the "single-failure assumption" as noted above, the Panel also considered the applicability of the single failure criterion to PSVs. The Panel expended considerable effort in searching for an answer to what appears to be a simple question: "Are PSVs active components subject to the single failure criterion, or are they passive components exempt from the single failure criterion?" NRR staff have taken the position that PSVs have consistently been treated as active components.

In the Panel's evaluation of the treatment of PSV failure potential (Section 3 below), a1cr 1111sw121 historical perspective is provided. In general, the Panel found that the classification of a component as "active" or "passive" depends on its design, application, and function. For example, passive components almost always do not need external power; usually do not need an external actuator (e.g., signal)31; sometimes do not involve any mechanical motion (e.g.,

movement of a valve disc)32; and sometimes do not involve any motion, either fluid or mechanical (e.g., piping). While it does not represent formal NRC guidance, additional views on passive components are included in International Atomic Energy Agency (IAEA) TECDOC-1624.33 This document states that "[s]afety related terms such as passive and inherent safety have been widely used, particularly with respect to advanced nuclear plants, generally without definition and sometimes with definitions inconsistent with each other." This guidance further defines four levels of "passivity" to "help eliminate confusion and misuse of the terms by members of the nuclear community." In addition, SECY-05-013834 also acknowledged and discussed inconsistencies in the use and application of the term "passive." Additional consideration of this topic by the Panel is documented in Section 3. 10 below.

The introduction to the GDCs and the related footnote define the applicability of the single failure criterion in terms of electrical versus fluid systems, and active versus passive components. Neither the GDCs nor NRC guidance define which characteristics of passive components are necessary to make a component exempt from the single failure criterion. Some examples are clear: pipes are passive components and pumps and motor-operated valves that operate to perform their safety functions are active components. As discussed in Section 3.6 31 For example, SECY-77-439 (NRC 1977) states: "Examples [of passive failures in fluid systems] include the failure of a simple check valve to move to its correct position when required, the leakage of fluid from failed components, such as pipes and valves- particularly through a failed seal at a valve or pump-or line blockage. Motor-operated valves which have the source of power locked out are allowed to be treated as passive components."

32 For example, NUREG-1800 (NRC 2001c) states that "'[p]assive' structures and components, for the purpose of the license renewal rule, are those that perform an intended function ... without moving parts or without a change in configuration or properties ... 'passive' may also be interpreted to include structures and components that do not display 'a change of state.'"

33 IAEA 2009 34 NRC 2005a below, check valves might be classified as active or passive components depending on certain specific considerations.

With respect to PSVs, the ASME BPV Code applicable to Byron and Braidwood includes requirements for overpressure protection that relate to the single failure criterion through several specific design and construction requirements. As a result, the PSVs are conservatively sized with sufficient margin to accommodate a single failure although the single failure criterion is almost never explicitly discussed or applied in accident analyses. The Byron and Braidwood UFSAR states that "adequate overpressurization protection is provided by the three installed safety valves." Neither the UFSAR system descriptions nor the safety analyses provide detailed discussions of potential PSV failures or their consequences. The principal discussion of potential PSV failures in the accident analyses occurs in the evaluation of an inadvertent opening of a PSV in UFSAR Section 15.6.1.

Most relevant for the current issue, the Byron and Braidwood UFSAR analyses of overpressure events (e.g., loss of load, loss of feedwater) do not apply the single failure criterion to cause a PSV to stick open (i.e., fail to reseat) when opening on steam flow. In addition, the UFSAR Feedwater System Pipe Break analysis (Chapter 15.2.8) does not apply the single failure criterion to cause a PSV to stick open either during steam discharge or during water discharge.

A survey of other Westinghouse-designed plants showed that this treatment of PSV valve performance during anticipated operational occurrences (AOOs, similar to ANS Condition II events) and postulated accidents (similar to ANS Condition IV events) has been consistent and without any identified exceptions.35 1.5 History and Review of Westinghouse NSAL and Related Activities Appendix C to this report provides the Panel's review of the issues identified by Westinghouse in NSAL-93-13 and its Supplement 1, how various licensees responded to these issues, and how the NRC was Involved In reviewing and approving these actions. This review provides the basis for the Panel's conclusions related to the approach taken by Byron and Braidwood to address these issues in their licensing basis, as well as on the "known and established standard" for event escalation from ANS Condition II to ANS Condition Ill, referred to hereafter as the "non-escalation position."

2

SUMMARY

OF THE APPEAL REVIEW PANEL FINDINGS For the reasons provided in Section 3, the Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Panel also concluded that, in preparing the Uprate SE and the Setpoint SE, the NRC staff exercised reasonable and well-informed engineering judgment when the NRC staff concluded that the PSVs were unlikely to stick open.

The non-escalation position does not establish specific standards for valve qualification, so the non-escalation position, standing alone, provides no basis for rejecting the licensee's reliance on EPRI valve testing . Moreover, the Panel found that no mistake or error occurred in the licensee's or previous staff's reliance on the EPRI testing program that included an evaluation of 35 Examples include Watts Bar (NRC 1982 and TVA 1983), North Anna (NRC 1976), and AP1000 (Westinghouse 2011 ).

water discharge through pressurizer valves. 36 Therefore, the Panel also concluded that the NRC staff's position on valve qualification in the Backfit SE is a new or modified interpretation of what constitutes compliance.

The Panel also concluded that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. The Panel included additional information about this finding in Section 6 and Appendices B and C below.

3 DISCUSSION The compliance exception to the Backfit Rule is intended to address failures to meet known and established Commission standards because of omission or mistake of fact. New or modified interpretations of what constitutes compliance do not fall within the exception. The Panel reviewed and evaluated the information referenced in this report to determine if, in 2001 and 2004, there was a known and established standard of the Commission relating to the potential for PSVs to fail following water discharge during IOECCS events.

In addition, the Panel considered the issue of "known and established standards of the Commission" as it relates to "event escalation." The NRR Appeal Decision stated that the Backfit SE "showed that the approvals at issue for Braidwood and Byron were inconsistent with the Agency's general position on the known and established standard at issue, in this case the progression of [ANS] Condition II events." The Panel recognizes that the non-escalation position, although not included in NRC regulations, is widely referenced in reactor licensing bases as an approach for addressing AOOs and postulated accidents as articulated in the GDCs. The non-escalation position is incorporated in Section 15.0.1.2 of the Byron and Braidwood UFSAR as "By definition, these faults (or events) do not propagate to cause a more serious fault, i.e., [ANS]

Condition Ill or IV events."

Exelon and the Panel agree that the non-escalation position is now, and was in 2001 and 2004, a part of the licensing basis of both Byron and Braidwood. In addition, the Panel supports the NRC staff's view that non-escalation (from ANS Condition II to ANS Condition Ill or IV) is a known and established standard applicable to Byron and Braidwood. However, the Panel also agrees with Exelon that the fundamental issue is not the non-escalation position, as the NRC staff contends, but rather the appropriate standard for PSV water discharge. In the absence of a PSV failure to reseat, the concerns articulated by the NRC staff in the backfit related to event classification, event escalation, and compliance with 10 CFR 50.34(b) and GDCs 15, 21, and 29 would no longer be at issue.

The Panel's evaluation of the treatment of PSV failure potential includes an assessment of multiple relevant references, which are discussed chronologically in the sections that follow.

36 "Pressurizer valves" is used in this report to refer to either PORVs or PSVs when discussing issues common to both types of valves.

3.1 General Design Criteria (1971)

In 1971, the Atomic Energy Commission published the GDCs, which had been under development since 1965.37 The introduction to 10 CFR Part 50, Appendix A addresses "Single Failure" in the section on Definitions and Explanations. The paragraph on single failures includes a footnote stating: "The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development" (emphasis added}.

3.2 Commission Paper on Single Failure (1977)

In response to several staff concerns and differing views on the subject of application of the single failure criterion, the Acting Director of NRR issued SECY-77-439 "[t]o inform the Commission of the present status and future use of the Single Failure Criterion as a tool in the reactor safety process."38 In part, that paper addressed the application of the single failure criterion to passive components in fluid systems, stating that "[a]pplication of the (single failure]

concept is complicated by tl:1e interrelationships between the various fluid and electrical systems and their supporting auxiliaries in a nuclear power plant. Furthermore, there is a need to stipulate the events and associated assumptions which must be considered during application of the Single Failure Criterion."

SECY-77-439 specifically spoke to how "additional passive failures"-that is, failures in addition to the initiating event-had been and should be addressed, stating (with emphases added):

During subsequent years [since the single failure footnote quoted above was published] staff assum ptions regarding the nature of passive failures which should be considered have not been com pletely consistent and there has been some disagreement. However, on the basis of the licensing review experience accumulated in the period since 1969, it has been judged in most instances that the probability of most types of passive failures in fluid systems is sufficiently small that they need not be assumed in addition to the initiating failure in application of the Si111gle Failure Criterion to assure safety of a nuclear power plant.

Furthermore, SECY-77-439 provides definitions and examples for distinguishing between active and passive failures. Among these examples, SECY-77-439 cites "the failure of a simple check valve to move to its correct position when required" as a passive failure. Of the examples cited in SECY-77-439, the check valve example is most similar from a mechanical perspective to the PSV failure addressed in the Backfit SE, as explained below in the discussion of SECY-94-084.

SECY-77-439 also stresses the use of engineering j udgment relating to the probability of component failure and does not suggest that valve "certification" or "qualification" in accordance with ASME standards should be invoked as the basis for such decisions.

3.3 TMI Action Plan Item 11.D.1 (1980)

As an element of the TMI Action Plan, the NRC staff required licensees to address the capability of relief and safety valves to perform their intended functions without failure. Specifically, 37 AEC 1971 38 NRC 1977 Item 11.0.1 states that "[pJressurized-water reactor [PWR] and boiling-water reactor [BWR]

licensees and applicants shall conduct testing to qualify the [RCS] relief and safety valves under expected operating conditions for design-basis transients and accidents." With reference to planned EPRI testing and other generic industry test programs, NUREG-0737 specified provisions for then-operating nuclear power plants and applicants for operating licenses and holders of construction perm its to address the TM I Action Plan items, including Item 11. D .1.

NUREG-0737 stated, for the performance testing of relief and safety valves for Item 11.D.1 , that

"[t]he testing should demonstrate that the valves will open and reclose under the expected flow conditions."

Although limited in scope, the EPRI test results did not identify any generic issues with PSVs or PORVs sticking open following water discharge. The NRC staff approvals summarized below show that the word "qualify" in this TMI Action Plan item was not intended to refer to ASME valve certification or qualification. Instead, "qualify" was used in a less formal sense to refer to a reasonable judgment that the valve would open to relieve pressure and then reliably reseat. As referenced in NUREG-0737, the EPRI test program was the widely used approach to address TMI Action Plan Item 11.D.1 at PWR nuclear power plants. The Westinghouse Owners Group submitted WCAP-10105 to the NRC in 1982 to demonstrate the acceptability of the EPRI testing program for PSVs and PORVs in Westinghouse-designed PWRs.39 3.4 NRC Closure of TMI Action Plan Item 11.D.1 for Byron and Braidwood (1988-1990)

A 1988 letter from the NRC staff to the licensee for Byron found the licensee's reliance on EPRI testing of PSVs to be acceptable. 40 The 1988 SE states that the test program was designed "[t]o reconfirm the integrity of the overpressure protection system and thereby assure that the

[GDCs] are met." As discussed in Appendix B to this report, the 1988 SE described the NRC staffs evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs. Based on the NRC staff and contractor review, the 1988 SE found that the performance of the PSVs and PORVs was acceptable based on the EPRI tests.

For the specific extended high pressure injection event, the 1988 SE states that water discharge through the PSVs and PORVs could be disregarded because of the long time available for operator action. However, the SE addressed water discharge through the PSVs and PORVs as part of the feedwater line break evaluation.

In the cover letter for the 1988 SE, the NRC staff states that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The 1988 SE contains no reference to or suggestion of a need for certification of these valves in accordance with the ASME BPV Code for water discharge capability. In 1990, the NRC staff also found the use of the EPRI test program similarly acceptable for Braidwood.4 1 39 WOG 1982 40 N RC 1988c, referred to as the 1988 SE 41 NRG 1990a 3.5 Westinghouse NSAL-93-013 and Supplement 1 (1993-1994)

In 1993, Westinghouse sent NSAL-93-013 to operating nuclear power plants in response to its discovery that potentially non-conservative assumptions had been used in the licensing analysis of the IOECCS event. Westinghouse recommended that licensees determine if their pressurizer safety relief valves (PSRVs)42 "are capable of closing following discharge of subcooled water.tt Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse also noted that "licensees may have qualified these valves in compliance to NUREG-0737, Item 11.D. 1." If the PSRVs were not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees reevaluate the IOECCS event with three possible options of ( 1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the accident.

Later, in Supplement 1 to NSAL-93-013, Westinghouse alerted licensees to potential reduced time for operator action if a positive displacement pump (a typical component of the CVCS) were in service, and to the need to qualify the PSRVs and the piping downstream of the PSRVs and PORVs if water discharge from the pressurizer is predicted.

Some licensees submitted license amendments that involved improvements to the PORVs and their circuitry to avoid water discharge through the PSVs (e.g., Salem43, Millstone44, Callaway45 ,

and Diablo Canyon46 ). The NRG staff review and approval of those proposed improvements relied on engineering judgment relative to the various test information and PORV circuitry upgrades described by individual licensees. The licensee for Byron and Braidwood submitted an LAR for similar PORV improvements,47 but that request was later withdrawn.48 As indicated below, the Panel's sampling review found at least two plants, in addition to Byron and Braidwood, that chose to address this issue by crediting the capability of PSVs to relieve water, based on the EPRI testing performed in response to TMI Action Plan Item 11.D.1.

3.6 Commission Paper on Passive Plant Designs (1994)

In 1994, in preparation for the design certification reviews of passive reactor designs (e.g., the Westinghouse Advanced Passive 1000 (AP1000) and the General Electric Economic Simplified Boiling-Water Reactor (ESBWR)), the NRC staff presented nine issues to the Commission for policy decisions. 49 Although PSV categorization and performance requirements were not explicitly addressed, the paper does include an issue on "Definition of Passive Failure" and an 42 Westinghouse used the term PSRVs. The specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code, and by the licensee. This difference in terminology is not significant to any of the findings or conclusions in this report.

43 NRC 1997 44 NRC 1998 45 NRC 2000 46 NRC 2004a 47 ComEd 1998 48 ComEd 1999 4 9 NRC 1994a extensive discussion on whether check valves are passive or active components and how they should be addressed in current plants and future passive designs.

SEeY-94-084 recognized the GDes and SEeY-77-439 as establishing long-standing requirements and guidance in this area. The paper acknowledge~(CT13JISW14J that the industry (including EPRI documents and ANSI/ANS 58.950) have been inconsistent with respect to check valve failures, sometimes considering them as "active failures" and sometimes as "passive failures." In SEeY-77-439, however, the NRe staff stated that the failure of a simple check valve to move to its correct position when required was a "passive failure." In addition, SECY-94-084 states that "[i]n licensing reviews, however, only on a long-term basis [e.g., long-term recirculation cooling following a loss of coolant accident (LOCA)] does the NRe staff consider passive failures in fluid systems as potential accident initiators in addition to initiating events."

The paper also states that "[f]or current plants, the NRe staff normally treats check valves, except for those in containment isolation systems, as passive devices during transients or design-basis accidents."

Furthermore, SEeY-94-084 states that "[r]edefining check valves as active components, subject to consideration for single active failures would cause these valves to be evaluated in a more stringent manner than that used in previous licensing reviews" (emphasis added). The NRe staff then recommended (and the Commission agreed51) that the NRe staff should "maintain the current licensing practice for passive component failures on the passive [advanced light water reactor] ALWR designs, and to redefine check valves, except for those whose proper function can be demonstrated and documented, in the passive safety systems as active components subject to single failure consideration." Therefore, the NRe's position on check valves was changed only for passive ALWR designs going forward.

The Panel considered the opening function of check valves and PSVs to be similar in that they both open through the motion of the valve disk under differential pressure with no external signal or motive power. The Panel also recognized that the ambiguity with respect to "passive" versus "active" component definitions and nomenclature exists for safety valves. In addition, the passive or active classification of check valves or safety valves may differ based on design considerations, inservice testing, or accident analyses. For example, the PSVs and PORVs, as well as numerous check valves, are classified as active components in the Byron and Braidwood inservice testing programs. However, for purposes of applying the single failure criterion in the GDe context, the Panel concluded that it is appropriate to consider the potential failure of a PSV following water discharge as a passive failure (consistent with the treatment of check valve failures for the operating fleet), provided the licensee or applicant qualifies the performance of the PSV in an acceptable mannertsM15J. !In tcn s1the case of Byron and Braidwood, the NRe staff accepted the EPRI testing associated with TMI Action Plan Item 11.D.1 to provide this qualification.

3.7 Draft Standard Review Plan Revision (1996)

The 1996 draft revision to SRP Sections 15.5.1 - 15.5.2 on IOEees and eves malfunctions includes extensive updates to the 1981 revision, but neither version includes any discussion, 50 ANS 1981 51 NRC 1994b criteria, or guidance on applying ASME Code requirements to PSVs or on applying the single failure criterion or any other failure assumption to PSVs.52 3.8 Power Uprate Reviews and License Amendments (2001-2006)

As part of the 2001 power uprate review for Byron and Braidwood, the NRC staff approved the analysis of an IOECCS (UFSAR Section 15.5.1) that included pressurizer filling, PSV water discharge, ECCS termination, and PSV closure. In the Backfit SE, the NRC staff indicated that the 2001 license amendment was predicated on the NRC's mistaken (unsubstantiated) belief that the valves were ASME-qualified (certified). However, the Panel's review of the SE and associated RAls showed that, in 2001, the NRC staff was well aware of the nature of the EPRI testing that the licensee relied on. The Panel did not find any evidence that the licensee claimed or the NRC staff believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows*that the NRC staff thoroughly considered the testing conducted on valves of the type installed at the plants and applied well-informed and reasoned technical judgment in reaching its conclusion that the EPRI testing provided appropriate qualification.

The Panel confirmed its conclusions and understanding about the 2001 NRC staff review via discussions with the individual who was the responsible Section Chief in the Reactor Systems Branch at the time. He informed the Panel that the 2001 license amendment was based on the exercise of staff engineering judgment and that there was no discussion of ASME BPV Code certification or qualification of valves. In addition, the Panel found that the NRC approved power uprates for other nuclear power plants that included comparable staff evaluations of water discharge through PORVs or PSVs based on test information provided by individual licensees.

For example, in 2001, the NRC granted a power uprate for Shearon Harris that included the operability of PORVs and PSVs during the discharge of subcooled water, referencing TMI Action Plan Item 11.D.1.53 As. noted above, in 2006, the NRC also granted a power *u prate for Beaver Valley. The SE for this Beaver Valley amendment referred to RIS 2005-29 and indicated that there was reasonable assurance that the PSVs would adequately discharge water and reseat following a spurious safety injection actuation, based on the EPRI test data from 1981 and an evaluation of the temperature of the liquid being discharged.

During the NRC evaluations of license amendments since the TMl-2 accident, the NRC staff has specified in some SEs that a PORVor PSV would be assumed to stick open if it was not qualified for liquid service. To address this concern, the NRC staff reviewed and accepted a variety of test information (including EPRI, Wyle, and vendor testing) submitted by individual licensees to demonstrate the capability of PORVs or PSVs to reseat following water discharge.

In the sample of SEs it reviewed, the Panel did not find a specific requirement for the PORVs or PSVs to be certified under the ASME BPV Code as capable of reclosing ~fter rcn71water discharge.

In 2004, the NRC issued license amendments for Byron and Braidwood granting an adjustment to the PSV setpoints. In an RAI, the NRC staff requested that the licensee perform a quantitative analysis regarding the number of opening cycles during which the PSV would be expected to pass water and the temperature of the water being discharged. In the Setpoint SE, the NRC staff concluded that the analysis was acceptable for assuring that the PSVs would remain operable following a spurious safety injection event.

52 NRC 1996 53 NRC 2001d 3.9 RIS 2005-29 (2005), and Proposed Draft Revision 1 to RIS 2005-29 (2015)

In 2005, the NRC staff issued RIS 2005-29 "to notify licensees of a concern identified during recent reviews of power uprate [LARs]." The RIS addressed the manner in which some licensees acted in response to NSAL-93-013. The RIS was issued at the division level in NRR and does not include a record of office-level concurrence. The RIS was not reviewed by CRGR.

The Panel requested information on the basis for the CRGR's decision not to review the proposed RIS before it was issued, but the CRGR staff could not find any related documentation. It appears to the Panel that the CRGR may not have reviewed the RIS because of assertions in the RIS such as these:

  • "This RIS requires no action or written response and, therefore, is not a backfit under 10 CFR 50.109. Consequently, the NRC staff did not perform a backfit analysis."
  • "This RIS is informational and pertains to a NRC staff position that does not depart from current regulatory requirements and practice."

A key statement in RIS 2005-29 is the following (with emphasis added):

The NRC staffs position is noted in the power uprate review standard, as follows:

"For the [IOECCS] and [CVCS] malfunctions that increase reactor coolant inventory events: (a)i non-safety-grade pressure-operated relief valves should not be credited for event mitigation and (b) pressurizer level should not be allowed to reach a pressurizer water-solid condition.".

However, the NRC staff review standard cited in the RIS (RS-001) is explicitly limited to EPU reviews, and states ing that "{ijlhe staff does not intend to impose the criteria and/or guidance in this review standard on plants whose design bases do, not include these criteria and/or guidance. No backfitting is intended or approved in connection with the issuance of this review standard."54 This intent of RS-001 to define and clarify the scope of EPU reviews, but not impose new requirements or new interpretations of requirements, was confirmed by the Panel in discussions with the manager responsible for developing and issuing RS-001. Therefore, contrary to the RIS statement, neither RS-001 nor RIS 2005-29 documented "known and established standards of the Commission" applicable to Byron and Braidwood.

The Panel also notes that neither RIS 2005-29 nor its draft Revision 1,55 which is currently under development, discuss water discharge certification requirements in accordance with the ASME BPV Code. In fact, as stated above, the NRC issued a 2006 power uprate amendment for Beaver Valley in which the SE cited RIS 2005-29 and yet relied on the EPRI testing data to address the concern.

3.10 SECY-05-0138 (2005)

SECY-05-0138 presents a comprehensive history of the application of the single failure criterion, including extensive discussion of the treatment of passive components in fluid 54 NRC 2003 55 NRC 2015a systems.56 The paper enclosed a July 2005 draft of an NRC staff technical report on the single failure criterion. Section 4.2.2 of this report acknowledges that "[o]ne particular issue identified in this project is the continued existence of the footnote to the definition of single failure in 10CFR

[Part] 50 Appendix A stating that the regulatory position on considering passive failures in fluid systems is under development." In Section 2.5.3, the draft report quotes from SECY-77-439 (discussed above) and recognizes that in current practice, as in 1977, "[p]assive failures in fluid systems are generally excluded from single-failure assessments."

SECY-05-0138 and the accompanying draft report present three alternatives for using a risk-informed and performance-based approach to address the single failure issue. The draft report clarifies that all of the alternatives "could include developing a position on single passive failures in fluid systems to replace the footnote now in 10 CFR Part 50 Appendix A definitions."

These documents make it clear that, with few exceptions, neither the NRC staff nor the Commission has established specific requirements relating to the treatment of passive component failures in fluid systems. The Panel believes the existence of this Commission paper, contemporaneous with discussions on potential PSV failures (e.g., RIS 2005-29), makes it clear that no specific "known and established standards" on PSV failures had been developed between 1977 and the time of the Byron and Braidwood license amendments in 2001 and 2004.

3.11 Standard Review Plan Revision (2007)

Revision 2 to SRP Sections 15.5.1 - 15.5.2 states:

If the plant is equipped with PORVs that are ( 1) safety-related equipment and (2) qualified for water relief, then they may be assumed to reseat properly after having relieved water. The [PSVs], too, may be assumed to reseat properly after having relieved water; but only if such valves have been qualified for water relief.

However, this section does not reference ASME BPV Code requirements for safety valve certification.

3.12 Backfit Letter and Subsequent Backfit Appeals (2015-2016)

The Backfit SE is predicated on the following positions:

  • "water relief through a valve that is not qualified for water relief will cause that valve to stick in its fully open position" (emphasis added)
  • "the licensee ... has not applied the single-failure assumption" (emphasis added)
  • "nor [has the licensee] provided ASME water qualification documentation for the PSVs

... the ASME ... orig1inal Overpressure Protection Report ... inservice test history ...

including both water and steam tests" (emphasis added)

The Backfit SE contends that an IOECCS would escalate to a more severe event. Such an escalation would be contrary to the Byron and Braidwood licensing basis (i.e., contrary to the ANS non-escalation position) and could be in non-compliance with the GDCs (as included in the 56 NRC 2005a Byron and Braidwood licensing basis) since an IOECCS with a stuck-open valve had not been analyzed and shown to meet the appropriate criteria for an AOO.

Based on its review of all the relevant documents and discussions with the individuals (staff and managers) involved in the original review and the backfit, the Panel has developed an understanding of the regulatory requirements and practices, the potential safety issues, and backfit rule obligations. The Panel has determined that the numerous, complex, and detailed regulatory and technical issues all depend on the answers to two critical questions on valve performance:

  • Must the PSVs in question be assumed to fail given liquid water discharge because of the lack of ASME BPV Code certification for water discharge?
  • Must the PSVs be assumed to fail in accordance with the GDC "single failure" requirements?

In the Backfit SE, the NRC staff indicated that "[o]ne assumption that is particularly important to the non-escalation criteria is that water relief through a valve that is not qualified for water relief will cause that valve to stick in its fully open position" (emphasis added). The Panel concluded that this issue-the treatment of potential valve failure- is not only "particularly important," it is the critical issue upon which the compliance backfit hinges.

Based on the historical evidence, the Panel concluded that there is not now, nor has there been, a known and established Commission standard (1) that PSVs must be assumed to fail following water discharge in the absence of ASME BPV Code certification for water discharge, or (2) that PSVs must be assumed to fail as part of single failure criterion analysis. The NRC staff's determination that ASME BPV Code certification is necessary first appears in the Backfit SE.

The determination that application of the single failure criterion is necessary first appears in the draft Revision 1 to RIS 2005-29. The Panel has not identified these positions being stated in any final NRC requirement or guidance document.

The Panel also concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that failures of PSVs need not be assumed to occur fo,towing water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. In preparing the Uprate SE and the Setpoint SE, the NRC staff exercised reasonable and well-informed engineering judgment when the NRC staff concluded that the PSVs were unlikely to stick open. On the bases of its document reviews and interviews, the Panel concluded that the NRC staff reviewers involved in the 2001 power uprate review were among the most experienced and senior reviewers in their areas of expertise. The NRC staff valve expert involved in the review was the agency's most knowledgeable individual on PSVs and the relevant ASME Code requirements, and was a nationally recognized expert. The Panel did not find any evidence that the NRC staff's issuance of the 2001 or 2004 license amendments was based on an omission or mistake of fact. Rather, the Panel concluded that the current NRC staff positions on valve qualification in the Backfit SE are new or modified interpretations of compliance.

In interactions with the Panel, NRR staff emphasized several issues raised in the Backfit Letter.

The Panel summarizes its consideration of those issues in the following subsections.

3.12.1 Non-Escalation Position and Valve Failure In the Backfit SE, the NRC staff discussed the definition of event conditions in ANS-51 .1/N18.2-1973 and the provision In this standard that events of one condition do not propagate to cause a more serious fault. This position is commonly known as the non-escalation position. In

~nteractions JCT18JlMAs19isw201with the Panel, NRR staff provided several clarifications on this topic, summarized by the Panel as follows:

  • ANS-51.1/N18.2-1973 defines the categories of design basis transients and accidents based on an anticipated frequency of occurrence (annually for ANS Condition II events).
  • It is a long-standing NRC position that escalation from one condition to another is not acceptable.
  • ANS-51.1/N18.2-1973 constitutes a known and established standard that has been reflected in NRC guidance documents and in the licensing basis of each U.S. nuclear power plant.

The Panel confirmed that th is ANS standard is referenced in several places in Chapter 15 of the Byron and Braidwood UFSAR. The Panel agrees that the non-escalation position is an established standard applicable to Byron and Braidwood, but did not identify historical evidence that implementation of this standard requires Exelon to assume that its pressurizer valves will fail open under water discharge conditions, to apply the single failure criterion to PSV failure in these circumstances, or to impose ASME Code requirements for certification, qualification, or testing of PSVs for water discharge.

3.12.2 Non-Escalation Position and Return to Service In the Backfit SE, the NRG staff makes reference to the time it would take to clean up a contaminated containment fiollowing a stuck-open pressurizer valve. In interactions with the Panel, NRR staff re-emphasized concerns that extended steam and water discharge through the pressurizer valves would result in the failure of the pressurizer relief tank rupture disk, would require repair of the damaged PSVs, and might cause an extended time period for the return to service of the nuclear power plant.

The Panel does not consider the time period necessary for the licensee to perform radioactive clean-up activities in the containment building, to inspect and conduct any necessary repairs to the PSVs, or to prepare for plant startup, to constitute issues that support a compliance backfit imposed by the NRG. The NRG staff would verify (e.g., through inspection) that the licensee had conducted these activities appropriately to protect the public health and safety prior to plant restart. The Backfit SE states that UFSAR Section 15.5.1.3 "implie[s]" that the plant will return to operation in a "short period," but the Panel found no eases basis for a timing requirement in UFSAR Section 15.5.1 .3. Also, the Panel did not find a regulatory requirement or basis for defining or limiting the time available for the plant to return to operation.

3.12.3 TMI Action Plan Item 11.0.1 and EPRI Testing Although the Backfit Letter and NRR Appeal Decision do not speak explicitly to TMI Action Plan Item 11.0.1, in interactions with the Panel, NRR staff stated that the known and established standard in question is the TMI Action Plan Item 11.D.I standard for licensees and applicants to conduct testing to qualify the RCS relief and safety valves under expected operating conditions for design-basis transients and accidents. As discussed above and in Appendix B to this report, the NRC accepted the EPRI testing to satisfy TMI Action Plan Item 11.D. 1 for Byron and Braidwood in SEs forwarded by letters in 1988 and 1990. Therefore, the Panel concludes that this known and established standard referenced by the NRC staff had been met for Byron and Braidwood.

In interactions with the Panel, the NRR staff further stated that an omission or mistake of fact occurred when the licensee failed to acknowledge that the EPRI testing program did not evaluate water discharge from the pressurizer valves during extended high pressure safety injection for Byron and Braidwood. As discussed in Appendix B to this report, in the 1988 and 1990 SEs for the Byron and Braidwood responses to TMI Action Plan Item 11.D.1, the NRC staff evaluated the capability of the PSVs and PORVs during feedwater line break accidents, including water discharge. In these SEs, the NRC staff found that the performance of the PSVs and PORVs with water discharge was acceptable based on the EPRI tests. Therefore, the Panel also concluded that the licensee's reference to the EPRI testing program was not an omission or a mistake of fact.

3.12.4 ASME Code Certification In the Backfit SE, the NRC staff stated that certain ASME Code information would be necessary to support water qualification of the PSVs. In interactions with the Panel, NRR staff stated that, to satisfy the standard for water discharge capability of pressurizer valves, it would be necessary to conduct flow capacity certification in accordance with the ASME BPV Code and inservice testing throughout the service life in accordance with the ASME OM Code. The NRR staff referenced certain licensing actions in which water discharge was not considered acceptable, or different actions were required. 57 As discussed in Appendix C to this report, the NRC staff required additional actions for some licensees to support reliance on the PORVs for water discharge and to avoid water discharge through the PSVs. The Panel found, however, that the NRC staff also allowed some licensees to rely only on EPRI testing without significant additional activities. The Panel did not identify instances where the NRC staff imposed certification by the ASME BPV Code and testing in accordance with the OM Code, or required alternatives to the ASME BPV or OM Codes, in the examples of NRC staff review of water discharge capability for pressurizer valves.

The NRR staff also identified for the Panel specific ASME Code provisions that it viewed as supporting its position that ASME Code requirements apply to qualification of pressurizer valves for water discharge. The NRR staff, however, did not provide evidence that the NRC staff has consistently interpreted these provisions as the NRC staff is now interpreting them. Given th~ I rcr21Jrsw221NRC staff's resolution of TMI Action Plan Item 11.D.1 and the 1Jariations in

~MAS231JSW24Jthe NRC staffs licensing practices, the Panel concludes that the NRR staff's current application of the ASME Code is not supported by the his torical record.

3.12.5 Conduct of 2001 and 2004 License Amendment Reviews In light of the wide range of positions taken by the NRC staff during its reviews of pressurizer valve capability since the TMl-2 accident, the Panel agrees that, in the course of preparing the 2001 Uprate SE or Setpoint SE, the NRC staff could have considered the need for the licensee for Byron and Braidwood to improve the reliability of the PSVs or PORVs for water discharge or to avoid water discharge through the PSVs by PORV improvements. The NRG staff may have s7 Salem (NRC 1997), Millstone (NRC 1998), and Callaway (NRC 2000) been able to justify additional actions, but they determined that it was not necessary. Instead, the NRC staff reviewers in 2001 used their expert engineering judgement to determine that it was not necessary to assume that the PSVs or PORVs would stick open with water discharge, based on EPRI test informa1ion. licensee supplemental information, and their own technical experience.

In discussions with the Panel, NRR staff raised a concern that the Setpoint SE does not document a re-review of the qualification of the PSVs and noted that if the Uprate SE had not found water discharge through the PSVs to be acceptable, it is unlikely that the NRC staff would 1

have approved this 2004 amendment. In Appendix C to this report, the Panel summarizes the discussion in the Setpoint SE of the PSV water discharge capability. The Panel recognizes that a staff review may rely on a previous more extensive review to determine the acceptability of a similar request. The Panel does not consider the review approach used in 2004 to challenge the acceptability of the 2001 review.

4 RESPONSE TO THE EDO QUESTIONS In establishing the Panel, the EDO asked the Panel to answer five specific questions, as well as evaluating the overall appropriateness of the backfit. The Panel's answers to these questions are provided below.

4.1 Were the approvals based on a mistake? If so, what was the mistake and what are the implications for Braidwood and Byron?

In responding the question, the Panel has considered the differing views of the NRR staff and the licensee on this issue. Those positions are summarized below:

  • In the NRR Appeal Decision, the NRC staff claims that "[t]he NRC erred in approving a sequence of events that allowed the [IOECCS], [CVCS] malfunction, and inadvertent opening of a pressurizer safety or relief valve analyses in the 2001 and 2004 [SEs]" and "the NRC staff understood the PSVs to be qualified for water relief when. in fact, they were not."
  • Exelon claims in the NRR Backfit Appeal that "the compliance exception requires more than simply asserting that the prior staff approvals were wrong- the NRC must demonstrate that the prior approvals were erroneous because of an omission or mistake of fact at the time of the approval. The NRC has not made that case here."

On the basis of its independent review, the Panel concluded that, in 2001 and 2004, the NRC staff did not misunderstand 1he qualification status of the PSVs and that it was not a mistake to undertake a review of or make a technically based safety finding on the likely successful performance of the valves. In the Panel's opinion, the actions of the Reactor Systems Branch in 2001 to reach out to the Division of Engineering's Mechanical Engineering Branch for expert technical review assistance was both appropriate and commendable. ~ fter I 1cr2sisM2sisw27Jconsidering the materials presented by the licensee in support of the 2001 and 2004 requests and discussing the 2001 review with one of the involved managers, the Panel found no indication that the senior reviewer evaluating the topic was misled regarding the qualification status of the PSVs, but rather used his expert judgment in determining the appropriate level of qualification for a technically complex topic for which there was not a single accepted approach. For these reasons, the Panel concluded that the NRC staff reviews and approvals of the 2001 and 2004 license amendments were not based on omissions or mistakes of fact.

4.2 What is the known and established standard for water qualification of PSVs?

The Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Commission has not established a more detailed or prescriptive standard.

4.3 What is the known and established standard for progression of postulated events between categories of severity?

For Byron and Braidwood, the NRG staff and the Panel agreed that the known and established standard for progression of postulated events between categories of severity is the "non-escalation position" specified in ANS-51.1/N18.2-1973. This position, which is included in the Byron and Braidwood UFSAR, requires that events of one condition do not propagate to cause a more serious condition (I.e., from ANS Condition II to ANS Condition Ill or IV}. The Panel concluded that the IOECCS (an AOO per the GDC definition and an ANS Condition II event) would escalate to a more severe event if a PSV were to stick open, or if both a PORV stuck open and its block valve failed to close. Such an escalation would be contrary to the Byron and Braidwood licensing basis (i.e., contrary to the ANS non-escalation position) and could be in non-compliance with the GDC (as included in the Byron and Braidwood licensing basis), since an IOECCS with a stuck-open valve had not been analyzed and shown to meet the appropriate criteria for an AOO. However, this event progression standard does not establish specific standards for valve qualification to determine whether a valve would stick open and cause this escalation. Therefore, the Panel concluded that it Is not the basis for a compliance backflt given the current set of facts. (Additional information about ANS-51.1/N18.2-1973 is included in Section 3.12.1 of this report.)

4.4 Does the current licensing basis for Braidwood and Byron comply with the applicable regulations? Is it adequate to provide protection to public health and safety?

+Re-For the specific technical issue reviewed by the Panel (i.e., blah, blah, blah) the Panel concluded that the current licensing basis for Byron and Braidwood complies with the applicable regulations ~as81-en the Ur SAR analyses, which tho NRG staff found acceptable through a reasonable and technically sound evah,1at-i9'A-\:IS~riate-GGFftmissi0A...saf-ety..standaFEl-&.-

~ e+2a11MAS29NSWJOJ1icensi ng basis has been determined by the NRG staff ta-and provide§ adequate protection to public health and safety.

4.5 Given that Exelon suggests that the NRC pursue a cost-justified substantial safety enhancement backflt, what is the contribution to overall plant risk of the current configuration at Braidwood and Byron?

The Panel requested RES to provide information and insights on the risk significance of the sequence at issue, to assure that the Panel's judgments were being made with a full understanding of their signiffcance, and to assist in responding to the EDO question.

The RES studfBtcn11 suggests that the most significant IOECCS sequence, assuming that all pressurizer overfill events lead to a small LOCA, contributes approximately 1 percent of the total internal event core damage frequency (CDF). In its report, RES estimated that the maximum benefit (CDF reduction) of 1.5E-07 per year would be achieved if the. plants were modified (jbackfitj is perfectl y effective such that pressurizer overfi'lling wasi..§ always prevented.,i If the PSVs are not assumed to always fail following water discharge (consistent with the NRC staff expert judgment in 2001) or if the plants were modified in a different way that did not prevent pressurizer o*.ierfillin~backfit is less than perfectly effective. JMAS32rthe risk-reduction benefit of implementing the backfit would be even smaller.

The Panel is aware of and sensitive to two important issues related to this question. First, NRR, not the Panel, is responsible for any decisions on alternative application of the backfit rule to this issue (through the other categories of adequate protection or cost-justified substantial safety enhancement). Second, the Panel does not wish to imply that the contribution to plant risk" should be seen as the only measure of enhanced safety. The issues of event classification and the non-escalation of events are essentially defense-in-depth concepts. Defense in depth has a recognized role and value in the regulatory process. The Panel is also aware that not every defense-in-depth feature has the same safety significance, and that the estimated risk significance (measured in core damage frequency) is very relevant.

Within the context described above, the Panel concluded that the contribution to overall plant risk is very small.

5

SUMMARY

AND CONCLUSIONS The compliance exception to the Backfit Rule is intended to address failures to meet known and established Commission standards because of omission or mistake of fact. New or modified interpretations of what constitutes compliance do not fall within the exception. Therefore, to address the appeal of the proposed compliance backfit, the Panel focused on determining if this case is most appropriately characterized as one in which the licensee "failed to meet known and established standards of the Commission because of omission or mistake of fact," or rather as a case of a "new or modified interpretations of what constitutes compliance."

The NRC staff's compliance backfit argument depends on two separate determinations:

1. the assumed failure of PSVs to reclose after passing water, and
2. the necessity of preventing "event escalation" (i.e., the position that "an incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently").

For the NRC staff's compliance backfit conclusion to be valid, both of these determinations must meet the above compliance backfit standard by involving failure to meet known and established standards of the Commission.

In the first of these determinations, the NRC staff's compliance backfit is based on the assumption in the Backfit SE that the PSV fails to reclose given the absence of "ASME water 58 NRC 2016f 59 The RES stud y explains that "any practical backfit remedy is not expected to be completely effective.

Therefore. this delta CDF represents the maximum oossible benefit from any backfit plant change."

qualification documentation." As indicated in the Backfit SE, the Uprate SE involved a technical evaluation of safety valve capability and likely performance under water-discharge conditions rather than a simple assumption of a failure. The NRR Appeal Decision indicates that "the 2001 and 2004 [license amendment] approvals occurred because the NRC staff understood the PSVs to be qualified for water relief when, in fact. they were not."

The Panel carefully considered these views and has reviewed the relevant documents including the licensee's responses to the NRC staffs RAls,60 the NRR technical branch's SE input,61 and the Uprate SE. The Panel did not find any evidence that the licensee had claimed or the NRC staff had believed that the valves were "qualified" in an ASME BPV Code certification sense; rather, the record shows thorough consideration of the testing conducted on valves of the type installed at the plant and a well-informed technical judgment that this testing provided appropriate qualification.

On the basis of its independent review. the Panel concluded that the NRC staff who prepared the Uprate SE did not misunderstand the qualification status of the PSVs and that it was not a mistake to undertake a review of or make a technically based safety finding on the likely successful performance of the valves. In the Panel's opinion, the actions of the Reactor Systems Branch in 2001 to reach out to the Division of Engineering's Mechanical Engineerin~

Branch for expert technical review assistance was both appropriate and commendable. !After I (CT33Jconsidering the materials presented by the licensee in support of the requestsLAR and discussing the review with one of the involved managers, the Panel found no indication that the senior reviewer evaluating the topic in 2001 was misled regarding the qualification status of the PSVs, but rather used his expert judgment in determining the appropriate level of qualification for a technically complex topic for which there was not a single accepted approach. For these reasons, the Panel concluded that the NRC staff review documented in the Uprate SE was not based on omissions or mistakes of fact.

The Panel concluded that three related technical and regulatory positions related to the PSVs (separate from the issue of the non-escalation position) underpin the backfit:

1. ASME water qualification (certification) documentation is required if a valve is *to be assumed to reclose after passing water.
2. Water discharge through a steam-qualified valve will cause that valve to stick in its fully open position.
3. PSVs are subject to a single-failure assumption.

In the Panel's view, none of these three positions were "known and established standards of the Commission" in 2001 or 2004 for detennining when it was appropriate to assume a failure of PSVs to reseat. In fact, they*were not "known and established standards of the Commission" in 2005 (when RIS 2005-29 was issued) or 2006 (when the Beaver Valley EPU was approved) or 2007 (when Revision 2 to SRP Sections 15.5.1 -15.5.2 was issued).

Moreover, these positions do not appear to be "established standards of the Commission" at present. The 2007 version of SRP Sections 15.5.1 - 15.5.2 allows credit for PORVs and PSVs if they have been "qualified for water relief." The NRC staff's determination that ASME BPV Code 60 ComEd 2000b, Exelon 2001 61 NRC 2001a

-23

  • certification is necessary first appears in the Backfit SE and is not addressed in any of the final NRC requirements or guidance documents reviewed by the Panel. The determination that application of the single failure criterion is necessary first appears in the draft Revision 1 to RIS 2005-29, which is still under development, and is not included in any final NRC requirement or guidance document reviewed by the panel.

The Panel concluded that the standard in place in 2001 and 2004 and at present is simply that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. In earlier documents addressing this topic, beginning with NUREG-0737, it is the Panel's view that the use of the word "qualified" or "qualificationn implies a general demonstration of capability, such as in the EPRI testing done in response to TMI Action Plan Item 111.D.1. In light of this standard. the Panel concluded that, when preparing the Uprate SE and the Setpoint SE, the NRC staff exercised reasonable and well-informed engineering judgment to conclude that the PSVs were unlikely to stick open.

Overall, the Panel concluded that the NRC staff's position on valve qualification in the Backfit SE is a new or modified interpretation of what constitutes compliance in addressing potential PSV failures following water discharge. Althou91h this new staff position represents a well-intentioned and conservative approach that could provide additional safety margin, the Panel concluded that it does not provide a basis for a compliance backfit.

Finally, in the absence of a PSV failure to reseat, the Panel concluded that the concerns articulated by the NRC staff in the Backfit SE related to event classification, event escalation, and compliance with 10 CFR 50.34(b) and GDCs 15, 21 , and 29 are no longer at issue.

The Panel's findings, therefore, support the Exelon backfit appeal.

6 ADDITIONAL PANEL THOUGHTS In addition to the specific finding relating to the backfit appeal, the Panel believes it is important to acknowledge, and for the NRC staff and licensees to appreciate, that water discharge through a PSV not specifically designed for such service is undesirable and should be minimized or avoided as a matter of conservative engineering and prudent operations. This is reinforced by the information provided in NSAL-93-013 and its Supplement 1, and the actions by various licensees in response to these documents, as well as the limited scope of the EPRI testing conducted over 30 years ago.

Operator training, control room procedures to terminate the event before pressurizer filling, and use of PORVs rather than reliance on PSVs, are clearly preferred and prudent measures, whether they form the facilities' UFSAR licensing basis and are assumed in the accident analyses or not.

The PSVs in question were designed for steam service. Steam relief is their normal service condition and applies to their ASME BPV Code certification. The Panel supports the previous NRC staff determinations for Byron and Braidwood and certain other plants that PSVs experiencing water discharge during an abnormal or accident condition need not be assumed to fail since there was a reasonable and technically well-informed engineering judgement to the contrary. However, the Panel also considers the actions by various licensees to improve the reliability and performance of the PORVs to avoid water discharge through the PSVs to be prudent in light of the design specifications of the PSVs.

The Panel considered but could not determine the extent to which the licensee for Byron and Braidwood addressed crediting water discharge through the PSVs, PORVs, or PORV block valves in the Byron and Braidwood inservice testing programs. The Panel recognizes that the difference between the intended use of these valves for overpressure protection and their infrequent use in response to certain plant events might be considered in implementing appropriate inservice testing activities.

The Panel notes that water discharge through various pressurizer valves is not a new issue because water discharge has always been credited (by the licensee for Byron and Braidwood and other licensees) for the feedwater line break analysis in UFSAR Section 15.2.8.

On the basis of Its review, the Panel also noted that the issue of pressurizer valve performance following water discharge appears to have generic applicability, and is not specific to only Byron and Braidwood. The Panel believes that resolution of this issue would have benefited from consideration of the generic nature of the issue through the appropriate NRC processes. -The Panel included the information it gathered and assessed to reach its conclusion regarding the generic nature of the issue in Appendices Band C of this report. Should the NRC staff undertake a generic look of the issues, it should, among other things, consider the information presented and questions raised in those appendices.- The review should also include a reassessment of the information and staff positions communicated in RIS 2005-29, as well as those included in its proposed Revision 1, which is currently under development, to determine whether or not these documents include new staff positions with the potential for inappropriate or unintended backfitting. As part of any generic assessment, the Panel also recommends that staff determine whether the information in RIS 2005-29 and its proposed Revision 1 should be incorporated into a regulatory guide or another guidance document.

APPENDIX A: HISTORY OF THE BACKFIT RULE AND THE COMPLIANCE EXCEPTION The Backfit Rule Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, "Backfitting," was originally promulgated in 1970.62 Because of perceived deficiencies in the rule, the U.S. Nuclear Regulatory Commission (NRC) substantially revised it in 1985.63 The 1985 rule was challenged in court, and the U.S. Circuit Court for the District of Columbia (D.C. Circuit) vacated this rule in its entirety. The D.C. Circuit took this action because it concluded that the revised rule could be interpreted to allow the NRC to consider costs in defining or redefining what is required for adequate protection of the public health and safety.64 In response, the NRC revised the Backfit Rule in 1988 to remove any implication that costs could be considered in defining or redefining adequate protection.65 The 1988 revisions only differed firom the 1985 rule to the extent necessary to address the court's concerns. The 1988 rule was also challenged in cou rt, but this time the D.C. Circuit upheld the rule.66 In its current form, 10 CFR 50.109(a)(1) defines backfitting as

... the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the im position of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position ....

Unless one of three specifie<:l exceptions apply, the NRC may impose a backfit only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50. 109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection."

Section 50.109(a)(4) sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2) and (a)(3). The first exception, the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee." 10 CFR 50.109(a)(4)(i). The second and third exceptions relate to actions ensuring adequate protection or to actions that involve defining or redefining adequate protection. 10 CFR 50.109(a)(4 )(ii)-(iii).

62 AEC 1970 (Author and year citations in footnotes refer to the designation of references in Appendix D to this report.)

63 NRC 1985 e4 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 824 F.2d 108, 119-20 (1987).

65 NRC 1988b 66 Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 880 F.2d 552 (1989).

Commission Polley The Commission addressed its intended application of the compliance exception in the 1985 rulemaking :67 The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.

In the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit."68 The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic] regulations."69 Backfitting Guidance Extensive information regarding the appropriate implementation of backfitting is provided in NUREG-1409.70 Relevant excerpts from this guidance are provided below.

Applicable Regulatory Staff Positions According to NUREG-1409, to be a backfit, "a new or revised staff position or requirement must be involved, that is, there must be a change in content or applicability of the previously applicable regulatory staff position (in the direction of increased safety requirements) . ... " An applicable regulatory staff position is a requirement or position already specifically imposed on or committed to by a licensee. Examples of applicable regulatory staff positions include:

  • legal requirements, as in explicit regulations, orders, and plant licenses and in amendments, condiUons, and technical specifications
  • written licensee commitments such as those contained in the final safety analysis report, licensee event reports, and docketed correspondence, including responses to NRC bulletins, generic letters, inspection reports, or notices of violation and confirmatory action letters
  • NRC staff positions that are documented explicit interpretations of more general regulations and are contained in documents such as the Standard Review Plan, branch technical positions, regulatory guides, generic letters, and bulletins e1 NRC 1985, at 38103 68 /d. at 38102 69

/d. at 38103. The 1988 rulemaking neither revised the compliance exception as stated in the 1985 rule nor provided additional guidance on its interpretation.

70 NRC 1990c A similar list of examples is provided in Manual Chapter 0514,71 which is also included as Appendix D to NUREG-1409. Manual Chapter 0514 was referenced in the 1988 rulemaking, and a working draft was provided to the Commission for information in SECY-88-102.72 Manual Chapter 0514 provides a definition of "applicable regulatory staff positions" that is slightly more detailed than the definition in NUREG-1409. This definition from Manual Chapter 0514 is quoted below, with additional detail beyond NUREG-1409 emphasized in underlined text.

Applicable regulatory staff positions are those already specifically imposed upon or committed to by a licensee at the time of the identification of a plant-specific backfit, and are of several different types and sources:

a. Legal requirements such as in explicit regulations, orders, plant licenses (amendments, conditions, technical specifications). Note that some regulations have update features built in , as for example, 10 CFR 50.55a, Codes and Standards. Such update requirements are applicable as described in the regulation.
b. Written commitments such as contained in the [Final Safety Analysis Report],

[Licensee Event Reports], and docketed correspondence, including responses to Bulletins, responses to Generic Letters, Confirmatory Action Letters, responses to Inspection Reports, or responses to Notices of Violation.

c. NRC staff positionsZ3. that are documented, approved, explicit interpretations of the more general regulations, and are contained in documents such as the

[Standard Review Plan], Branch Technical Positions, Regulatory Guides, Generic Letters, and Bulletins; and to which a licensee or an applicant has previously committed to or relied uoon. Positions contained in these documents are not considered applicable staff positions to the extent that staff has, In a previous licensing or inspection action, tacitly or explicitly excepted the licensee from part or all of the position.74 How Regulatory Positions are Established NUREG-1409 provides responses to a number of questions regarding backfitting. The following response was given to questions asking, "Is it appropriate for the NRC staff to rely on informal or formal communications to other licensees as official NRC positions? What about NRC tacit approval of documents?"

Informal or formal communications to one licensee are not official positions to all licensees. Section 053 of Manual Chapter 0514 identifies what can be applied as official staff positions in a plant-specific context. They are legal requirements such as contained in explicit regulations, orders, and plant licenses; written commitments such as contained in final safety analysis reports, licenses event reports, and docketed correspondence; and documented, approved explicit interpretations such as contained in the [Standard Review Plan], branch technical 71 NRC 1988c 72 NRC 1988a 73 Requirements may be imposed by rule or order. Staff interpretations such as examples of acceptable ways to meet requirements are not requirements in and of themselves.

74 Imposition of a staff position from which a licensee has previously been excepted is a backfit.

positions, regulatory guides, generic letters, and bulletins. Orders, licenses, and written commitments are applicable only to a particular licensee.

If the NRC staff previously exempted a licensee from a legal requirement or approved position, it is not applicable to that licensee for the purpose of backfit consideration. Explicit exemption would be done formally in writing. The Appendix to NRC Manual Chapter 0514 discusses tacit approval under reanalysis of issues. Two situations are covered. In the first case, staff review of a previously accepted licensee action or program may result in a requested change. This would be classified as a backfit because it represents a change in a previous staff position and would require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed in the backfit rule). In the second case, a licensee submittal committing to a specific course of action that has not received timely NRC staff review is implemented by the licensee. In this case, It is considered that the NRC staff tacitly accepted the licensee's action since timely notice to the contrary was not given. If the NRC staff subsequently adopts a different position and requests a change in the licensee action, this change may be classified as a backfit and thus require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed In the backfit rule}.

NUREG-1409 also addresses a question regarding tacit approvals by an Inspector: "If an inspector has previously accepted (i.e., provided tacit approval of) a licensee's method, does a specific request for change constitute a backfit and if so, is a backfit analysis required?" The response is:

Cases where an inspector provides tacit approval are relatively rare. Simply not challenging a licensee's practice normally would not be considered tacit approval. The only example provided in Manual Chapter 0514 is a case where the NRC has indicated tacit approval by not acting in a reasonable time on a licensee submittal and the licensee has moved ahead to implement the proposal described in the submittal. For the purpose of this question, it would most likely arise in connection with review of a licensee response to an inspection report.

Explicit approval cou ld be provided in an Inspection report that states that a particular approach is acceptable. However, conclusions of that nature are usually made in [safety evaluations] rather than inspection reports.

Comp/lance Backfit Guidance NUREG-1409 gives the following response to the question, "[h]ow does the backfit rule apply to new staff positions that reflect an evolving understanding of technical issues?"

An evolving understanding of issues does not, by itself, define which category fits a particular backfit. Judgment must be applied to the facts of each particular case to determine whether the backfit is for compliance, to provide adequate protection, to redefine adequate protection, or to achieve a cost-justified substantial safety enhancement. For example, with regard to compliance, the 1985 statement of considerations for 10 CFR 50.109 indicates that "the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact....new or modified interpretations of what constitutes compliance would not fall within the exception...."

NUREG-1409 also provides. an example where an evolving understanding of technical issues resulted in a compliance backfit that was apparently justified for at least some licensees. In response to industry claims that Bulletin 88-11 75 lacked any backfitting justification, the NRC staff responded:

Although the justification was not printed in the bulletin, NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," was justified as a backfit. It is an example of a backfit that was determined by the responsible NRC official to be required as a matter of compliance with existing requirements and commitments.

The CRGR reviewed the bulletin and concurred. The regulations currently require licensees to meet the applicable codes of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code. Because of the NRC staffs concern with the integrity of the surge line, licensees were requested to perform their fatigue analysis in accordance with the latest ASME Section Ill requirements that incorporate high cycle fatigue analysis. The justification provided by the NRC staff was that previously unconsidered thermal stratification phenomenon may invalidate the existing analysis performed to confirm the integrity of the surge line.

Subsequently, it was understood that some licensees believed that the NRC staff's rationale was in error because they were not committed to the latest ASME Section Ill requirements by virtue of their license commitment. However, the issue became moot because these licensees undertook the analysis voluntarily in view of the safety importance of the issue and the fact that previous versions of the ASME Code did not completely address the concern.

75 NRC 1988e APPENDIX B: QUALIFICATION OF PRESSURE RELIEF VALVES IN NUCLEAR POWER PLANTS IN RESPONSE TO THE TMl-2 ACCIDENT Byron and Braidwood Design and Code Requirements Nuclear power plants in the United States use various types of pressure relief valves to protect personnel and equipment from overpressure events within reactor fluid systems. Pressure relief valves include safety valves, safety relief valves, and relief valves, with different designs, operating conditions, and requirements. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), Section Ill, Division 1, specifies requirements for the design, operation, installation, and testing of pressure relief valves used for various functions in nuclear power plants.76 For example, the ASME BPV Code (2007 Edition) in Article NB-7000, Overpressure Protection, specifies requirements for several service conditions:

  • steam and air or gas service for safety valves;
  • liquid service for relief valves; and
  • steam, air or gas, and liquid service for pilot operated or power actuated pressure relief valves.

The ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for the preservice and inservice testing (1ST) programs for pressure relief valves in nuclear power plants.

Byron, Units 1 and 2 (Byron) and Braidwood, Units 1 and 2 (Braidwood) are Westinghouse-designed pressurized-water reactors (PWRs) that received their construction permits under Title 10 of the Code of Federal Regulations (1 O CFR), Part 50, in December 1975. The pressurizer for each unit is equipped with three pressuriz,er safety valves (PSVs) and two power-operated relief valves (PORVs). The three PSVs are Crosby Model HP-BP-86, size 6M6 (6-inch), spring-loaded pop type, opened by direct fluid pressure. The PORVs are Copes-Vulcan Model D-100-160 3-inch pneumatic-actuated globe valves that respond to a signal from the pressure sensing system or to manual control. Each PORV can be isolated by a motor-operated block valve.

The ASME BPV Code of record for the design of the PSVs at Byron and Braidwood is the 1971 Edition through the Winter 1972 addenda of the ASME BPV Code, Section Ill. The ASME BPV Code applicable to Byron and Braidwood includes requirements for overpressure protection, including the following:

  • Section NB-7300, "Overpressure Protection Report," in NB-7320(f) requires that the report include the redundancy and independence of the pressure-relief devices and their associated pressure-sensing and controls systems employed to preclude a loss of overpressure protectiion in the event of a failure of any pressure-relief device, or its sensing element, or its associated control, or an external power source.

76 References to individual ASME Code publications are not provided in Appendix D, but they are publicly available from ASME for a fee.

  • Paragraph NB-7411, "Relieving Capacity of Pressure-Relief Devices," specifies that the total rated relieving capacity shall be sufficient to prevent a rise in pressure of more than 10 percent above system design pressure ( at design temperature) within the pressure-retaining boundary of the system, under any pressure transient anticipated to arise as summarized in the Overpressure Protection Report.
  • Paragraph NB-7421 , "Required Number and Capacity of Pressure-Relief Devices for Nuclear Systems," states that the required relieving capacity intended for overpressure protection of a nuclear power system or portions of the system shall be secured by the use of at least two pressure-relief devices.

At the time of the Byron and Braidwood operating license review, Revision 1 of Standard Review Plan (SRP) Sections 15.5.1-15.5.2 and Section 15.6.1 provided general staff guidance for these plant transients.n In March 2007, the NRC staff issued Revision 2 to these SRP sections with significantly more detail, including a statement Jndicatinq !1sM34Jthat PSVs and PORVs are assumed to fail open if they relieve water without being qualified. 78 Actions Following Three Mile Island, Unit 2 Accident The accident at Three Mile Island, Unit 2 (TMl-2) on March 28. 1979, included failure of a PORV on the pressurizer to reclose properly during the event. Based on lessons learned from the TMl-2 accident, the NRG issued recommendations regarding performance testing of safety and relief valves used in nuclear power plants in NUREG-0578.79 In particular, the NRC staff recommended in Section 2.1.2, "Performance Testing for BWR [boiling-water reactor] and PWR Relief and Safety Valves," of NUREG-0578 that nuclear power plant licensees commit to provide performance verification by full-scale prototypical testing for all relief and safety valves.

In October 1980, the NRC issued a letter to all then-operating nuclear power plants and applicants for operating licenses and holders of construction permits forwarding NUREG-0737.80 TMI Action Plan Item 11.D.1 in NUREG-0737 specified the NRC position that PWR and BWR licensees and applicants shall conduct testing to "qualify" the reactor coolant system {RCS) relief and safety valves under expected operating conditions for design-basis transients and accidents. The detailed clarification in NUREG-0737 of this NRC position specified the following:

Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test-pressures shall be the highest predicted by conventional safety analysis procedures. [RCS] relief and safety valve qualification shall include qualification of associated control circuitry, piping, and supports, as well as the valves themselves.

A. Performance Testing of Relief and Safety Valves-The following information must be provided in report form by October 1, 1981 :

77 NRC 1981 band NRC 1981 c 78 NRC 2007b and NRC 2007c 79 NRC 1979a 80 NRC 1980b and NRC 1980c

( 1) Evidence supported by test of safety and relief valve functionability for expected operating and accident (non-[anticipated transient without scram])

conditions must be provided to NRC. The testing should demonstrate that the valves will open and reclose under the expected flow conditions.

(2) Since it is not planned to test all valves on all plants, each licensee must submit to NRC a correlation or other evidence to substantiate that the valves tested in the EPRI {Electric Power Research Institute) or other generic test program demonstrate the functionability of as-installed primary relief and safety valves. This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the final safety analysis report (FSAR). The effect of as-built relief and safety valve discharge piping on valve operability must also be accounted for, if it is different from the generic test loop piping.

(3) Test data including criteria for success and failure of valves tested must be provided for NRC staff review and evaluation. These test data should include data that would permit plant-specific evaluation of discharge piping and supports that are not directly tested.

In describing the type of review to be conducted for this regulatory position, the NRC staff stated the following:

Pre-implementation review will be performed for EPRI and BWR test programs with respect to qualif ication of relief and safety valves. Also, the applicants' proposal for functional testing or qualification of PWR valves will be reviewed.

Post-implementation review will also be performed of the test data and test results as applied to plant-specific situations.

In specifying the documentation required to satisfy this regulatory position, the NRC staff stated the following:

Pre-implementation review will be based on EPRI, BWR, and applicant submittals with regard to the various test programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following valve qualification dates can be met:

Final PWR (EPRI) Test Program--July 1, 1980 Final BWR Test Program--October 1, 1980 Block Valve Qualification Program--Janua ry 1, 1981 Post-implementation review will be based on the applicants' plant-specific submittals for qualification of safety relief valves and block valves. To properly evaluate these plant-specific applications, the test data and results of the various programs will also be required by the following dates:

PWR (EPRl)/BWR Generic Test Program Results--July 1, 1981 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee/applicant preliminary review of generic test program results--July 1, 1981 Plant-specific reports for safety and relief valve qualification--October 1, 1981 Plant-specific submittals for piping and support evaluations--January 1, 1982 Plant-specific submittals for block valve qualification--July 1, 1982 EPRI Testing In October 1982, EPRI issued NP-2670-LD to address testing of PORVs.81 This report has been referenced by certain licensees (e.g., Section 15.2.14 of the North Anna, Units 1 and 2 Updated Final Safety Analysis Report (UFSAR)B2 ).

In December 1982, EPRI issued NP-2628-SR, which described safety and relief valve tests for types of valves in service at nuclear power plants.83 In particular, Section 3.5 documented the testing of Crosby safety valves similar to the PSVs at Byron and Braidwood, including two water tests. The report indicated chattering of the safety valves, with subsequent inspection finding galled surfaces and damage to internal parts. Section 4.6 documented testing of Copes-Vulcan relief valves similar to the pressurizer PORVs at Byron and Braidwood, although the extent of water testing was not fully described. The report indicated no damage found during the inspection of the Copes-Vulcan relief valves. The report did not indicate any failures of the Crosby or Copes-Vulcan valves to reseat after discharging water during the testing.

EPRI also published NP-2770-LD in the early 1980s to describe the testing of PWR primary system safety valves. Volume 1, issued in December 1982, provides a summary of the test program and its results.84 Section 4.5 of Volume 1 Indicates that the following tests were performed on the Crosby 6M6 PSV: 11 steam tests with lfilled loop seals, 3 steam-to-water transition tests, and 2 water tests. The report states that the valve experienced chatter during the tests, and one water test had to be terminated. The individual volumes of EPRI NP-2770-LD discuss the test results for each specific PSV type. Volume 6, issued in March 1983, provides the test details for the Crosby 6M6 PSV.

Westinghouse Evaluation of EPRI Testing In July 1982, the Westinghouse Owners Group (WOG) submitted WCAP-10105.85 In WCAP-10105, the WOG indicated that the design specification for PSVs in Westinghouse-designed nuclear power plants is for steam service only. Based on a review of the EPRI test data, the WOG concluded that the valves performed with chatter, but did not identify any valve damage.

In January 1988, Westinghouse issued WCAP-11677, which compared the EPRI test data with feedwater line break safety analyses.86 Westinghouse determined that all nuclear power plants addressed in the EPRI testing had PSVs that would operate reliably during water discharge.

Westinghouse evaluated the performance of the Crosby 6M6 PSVs during the EPRI tests, and 81 EPRI 1982a 62 VEPCO 2015 63 EPRI 1982b 64 EPRI 1982c as woG 1982 86 Westinghouse 1988 considered that the performance involved less significant flutter (half lift motion) than the chatter (full lift motion) determined in the EPRI report. Westinghouse concluded that the Crosby 6M6 PSV can pass slightly subcooled water at a minimum up to three times without damage.

Byron and Braidwood Licensing and Response to TMI Requirements The NRC safety evaluation reports (SERs) associated with the issuance of the operating licenses for Byron and Braidwood included evaluation of the TMI Action Plan items. 87 In the introduction to the Braidwood SER, the NRC staff stated that the review and evaluation of compliance by the applicant with the licensing requirements established in NUREG-066088 and TMI Action Plan Item 11.D.1 were incorporated into the reviews summarized throughout the SER.

Appendix E, "Requirements Resulting from TMl-2 Accident," to the Byron and Braidwood UFSAR in Section E.23, "Relief and Safety Valve Test Requirements (11.D.1 )," references the 1982 transmittal from Consumers Power of a test report for the EPRI safety and relief valve test program.89 The UFSAR states that the final evaluation of the data indicated that the relief and safety valves will perform their intended functions for all expected fluid inlet conditions. The UFSAR also references the October 1982 licensee evaluation of the adequacy of the relief and safety valves that had been submitted to the NRC.90 In Supplement 1 to the Braidwood SER,91 in Section 3.9.3.3, "Design and Installation of Pressure Relief Devices," the NRC staff stated that EPRI had completed a full-scale valve testing program and referenced the July 1982 submittal of WCAP-10105. The NRC staff stated that the applicant responded to a requirement to demonstrate operability of these valves through submittals dated July 1, 1982, October 26, 1982, and December 30, 1983. On the basis of a preliminary review, the NRC staff concluded that the applicant's general approach to responding to this item was acceptable, and provided adequate assurance that the RCS overpressure protection systems at Braidwood could adequately perform their intended functions. The NRC staff stated that if the detailed review revealed that modifications or adjustments to safety valves, PORVs, PORV block valves, or associated piping, would be needed to ensure that all intended design margins were present, the NRG staff would require that the applicant make appropriate modifications. The NRC staff categorized this issue as a Confirmatory Item. The NRC issued operating licenses for all four Byron and Braidwood Units between February 1985 and May 1988.

Closure of TMI Action Plan Item 11.D.1 for Byron and Braidwood Following the issuance of the operating licenses, the NRC staff documented its review of the response to TMI Action Plan Item 11.D.1 for Byron and Braidwood via two letters that transmitted similar Technical Evaluation Reports (TERs) developed by Idaho National Engineering Laboratory (INEL).92 In its letters, the NRC staff indicated that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The TERs described the INEL review of the EPRI testing of PSVs and PORVs 87 NRC 1983 and NRC 1986b (Braidwood), NRC 1984 and NRC 1987a (Byron) 88 NRC 1980a S9 Consumers 1982 9

° ComEd 1982 91 NRC 1986b. Similar discussion appears in NRC 1984 for Byron, and NRG 1987a (also for Byron) states that TMI Action Plan Item 11.D.1 had been closed in NRC 1984.

92 NRG 1988c (Byron) and NRG 1990a (Braidwood) similar to the Byron and Braidwood pressurizer valves. The TERs concluded that Byron and Braidwood had provided an acceptable response to TMI Action Plan Item 11.D.1.

Section 4.2.3, "Extended High Pressure Injection [HPI] Event," of the TERs stated that the potential for water discharge in extended HPI events can be disregarded for an extended high pressure injection event because at least 20 minutes would be available for operator action.

Water discharge was evaluated, however, in Section 4.2.2, "FSAR Liquid Transients," of the TERs. This section discussed the evaluation of the PSVs and PORVs for feedwater line break accidents that would include water discharge, and determined that the EPRI tests were applicable to the Byron and Braidwood PSVs and PORVs.

In addition, Section 4.3.1 , "Safety Valves," and Section 4.3.2, "Power Operated Relief Valves,"

of the TERs determined that the performance of the PSVs and PORVs was acceptable based on the EPRI tests, including water discharge tests. The TERs indicated that the PSV had two applicable tests: a loop seal steam-water transition test where the valve opened, chattered and stabilized to close; and a saturated water test where the valve opened with water, chattered, and stabilized. The TERs indicated that the PORV opened and closed on demand in the loop seal steam-water transition test, with a bending moment that was evaluated by analysis.

APPENDIX C: CONCERNS REGARDING PERFORMANCE OF PRESSURIZER VALVES UNDER WATER FLOW CONDITIONS Westinghouse Nuclear Safety Advisory Letter In 1993 and 1994, Westinghouse issued Nuclear Safety Advisory Letter (NSAL)93-013 and its Supplement 1 to operating nuclear power plants (including Byron and Braidwood). 93 These advisories resulted from Westinghouse's discovery that potentially nonconservative assumptions were used in the licensing analysis of the Inadvertent Operation of the Emergency Core Cooling System at Power (IOECCS) event.

In NSAL-93-013, Westinghouse recommended that licensees determine whether their pressurizer safety relief valves (PSRVs)~ are capable of closing following discharge of subcooled water. Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse indicated that water discharge through the power-operated relief valves (PORVs) is not a concern, becaus,e the PORV block valves can be used to isolate the PORVs if they fail to close. If the PSRVs are not designed or qualified for subcooled water discharge, Westinghouse recommended that licensees re-evaluate the IOECCS event with three possible options of (1) reducing emergency core cooling system (ECCS) flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the event.

In Supplement 1 to NSAL-93-013, Westinghouse informed licensees of a potential reduced time for operator action if a positive displacement pump is in service. Westinghouse also advised licensees to qualify the PSRVs and the piping downstream of the PSRVs and PORVs if water discharge from the pressurizer were predicted.

Some licensees of operating nuclear power plants informed the NRC of their actions to address the potential concerns regarding water discharge from pressurizer safety valves (PSVs) and PORVs. A sample of actions by nuclear power plant licensees is summarized below In the "Plant-Specific Actions" section.

Additional NRC Generic Communications and Guidance In 2003, the NRC staff issued a review standard for extended power uprate (EPU) reviews.95 Item 8 on page 7 of the review standard states that pressurizer level should not be allowed to reach a pressurizer water-solid condition .

In 2005, the NRC issued Regulatory Issue Summary (RIS) 2005-29 to notify nuclear power plant licensees of a concern identified during reviews of power uprate requests. 96 In RIS 2005-29, the NRC staff stated that typically Condition II scenarJos97 involve discharging water through relief or safety valves that are not qualified for water discharge. The NRC staff stated that these valves are then assumed to fail in the open position and create a small-break loss-of-coolant 93 Westinghouse 1993 and Westinghouse 1994 94 Although W estinghouse used the term PSRVs. the specific valves for Byron and Braidwood should be designated as "safety valves" or "pressurizer safety valves" as they are by the manufacturer, in the ASME BPV Code. and by the licensee.

95 NRC 2003 96 NRC 2005b 97 As defined in American Nuclear Society (ANS) Standard 51 .1/N18.2-1973 (ANS 1973).

accident (LOCA). The NRC staff stated that it was concerned that some licensees may be crediting PORVs without qualification for water discharge and without establishing additional restrictions to ensure the availability of PORVs and block valves. The NRC staff stated that the advice in Westinghouse NSAL-93-013 to use the PORV block valves to isolate the PORVs is inconsistent with the non-escalation position.

In draft Revision 1 to RIS 2005-29, the NRC staff addresses the specific ANS Condition II scenarios of chemical volume and control system (CVCS) malfunction, inadvertent opening of a PORV or PSV (IOPSRV), and the IOECCS event.98 Regarding the eves malfunction, the NRC staff states that performing only a reactivity anomaly analysis or assuming that this malfunction is not as severe as the IOECCS event is not acceptable. Regarding the IOPSRV event, the NRC staff stated that inadvertent opening of PSV or PORV could continue as an ANS Condition Ill small break LOCA and fails to meet the non-escalation position. Regarding the IOECCS event, the NRC staff states that five of the alternative approaches in NSAL-93-013 fail to meet the non-escalation position. The NRC staff indicated that these unacceptable alternative approaches are:

1. closing the block valve,
2. assuming that the PORV is not operable,
3. addressing a stuck-open PORV or PSV as a separate ANS Condition II event,
4. determining that a stuck-open PORV or PSV is not as severe as a small break LOCA, or
5. determining that RCS loss through PORV is made up by ECCS flow.

Additional General PSV/PORV Information In 2004, EPRI issued Technical Report 1011047, which evaluated the potential increase in failure rates following steam and liquid relief through safety valves based on expert judgement.99 The report found that the increase in failure rates is difficult to estimate because of limited data.

However, the experts considered that repeated water discharge through safety valves might cause increased chatter, and therefore, an increased failure rate.

In 2011 , the NRC summarized relief valve performance data in NUREG/CR-7037, based on a study by the Idaho National Laboratory. 100 With respect to pressurizer PORVs, the report found four separate water discharge events at four PWR plants. The report estimated 698 total demands on these PORVs during their water discharge events with no failures to close. The report also summarized test data for three valve types from the Equipment Performance and Information Exchange (EPIX) database maintained by the Institute of Nuclear Power Operations. The report indicates two failures of PORVs to reclose during 2070 demands, but does not specify water or steam service for the EPIX test information. With respect to PSVs, the report indicates two failures out of four total demands following plant scrams, but does not indicate water or steam service. Following a request by the Panel, NRC staff from the Office of Nuclear Regulatory Research provided Licensee Event Report information indicating that the two PSV failures involved incomplete reseating of the valves with leakage of 25 and 200 gallons 98 NRC 2015a 99 EPRI 2004 100 NRC 2011 per minute, respectively. The report summarized EPIX test data for PSVs as no failures to reclose during 1805 demands.

Plant-Specific Actions Diablo Canyon In 1996, the licensee for Diablo Canyon Power Plant (Diablo Canyon) submitted a report of its evaluation under Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, "Changes, tests and experiments," of the potential for an IOECCS event. 101 The submittal included NSAL-93-013 and its Supplement 1 as enclosures. The licensee indicated that the PSVs had not been initially qualified for water discharge, but were subsequently qualified to discharge water for a brief period. The licensee indicated that WCAP-11677 (which evaluated the EPRI testing) was applicable and demonstrated that the PSVs were operable.

In 2004, the NRC issued a license amendment for Diablo Canyon that allowed credit for actuation of the PORVs in response to inadvertent safety injection (SI) actuation, to avoid challenges to the PSVs. 102 To support the NRC staff's review, the licensee submitted additional information related to the capability of the PORVs to function adequately under conditions predicted for design-basis transients and accidents. 103 In response to a question regarding the design adequacy of the PORVs if the pressurizer becomes water solid, the licensee s.tated that the PORV had no requirements for ASME BPV Code certification, but referenced a January 1986 NRC letter that had accepted the adequacy of the PORV and block valve design and confirmatory testing for a range of fluid conditions (full pressure steam, steam to water transition, and subcooled water fluid).104 Salem In 1997, the NRC issued a license amendment revising the technical specification (TS) for Salem Nuclear Generating Station, Units 1 and 2 (Salem) to ensure that the automatic capability of the PORVs to relieve pressure would be malntained. 1015 In response to NSAL-93-013, the licensee determined that an inadvertent SI actuation at power could cause the pressurizer to become water solid. The PSVs would lift and discharge water if the automatic operation of the PORVs were not made available for reactor coolant system (RCS) depressurization early in the transient. In that the Salem PSVs were not designed to relieve water, it was noted that water discharge could cause the PSVs to fail in the open position.

During the review, the NRC staff noted that the PORVs were not designed to "safety related" standards and, thus, could not be credited for automatic mitigation of an inadvertent SI actuation at power. In response, the licensee proposed an upgrade of the PORVs to eliminate the possibility that a single active failure of a PORV component could prevent the mitigation of an inadvertent SI actuation at power. As discussed in the NRC staff's safety evaluation (SE), the licensee implemented modifications to the PORV circuitry to qualify the upgraded circuitry as safety-related.

10 1 PG&E 1996 10 2 NRC 2004a 103 PG&E 2003 104 NRC 1986a 10s NRC 1997 Regarding PORV performance, the licensee evaluated the PORV air accumulators and determined that they had sufficient capacity for the inadvertent SI event. The licensee also reported that endurance tests had been performed with five different trims (with different trim materials) on one PORV at Wyle Laboratories to demonstrate that (1) after 2000 consecutive operations, there were no packing leaks or packing gland adjustments required; (2) there was no diaphragm failure; and (3) the solenoid valve withstood 10,000 operations without any loss of function. Based on this information. the NRC staff concluded that the PORV performance was acceptable to mitigate an inadvertent SI event.

Millstone 3 In 1998, the NRC issued a license amendment for Millstone Nuclear Power Station, Unit 3 (Millstone 3) that revised the TS to ensure that the capability of the PORVs to relieve pressure would be maintained.106 The revised TS Bases stated that the PORVs and their associated piping had been demonstrated to be "qualified" for water discharge. The PORVs would prevent water discharge from the PSVs, for which qualification for water discharge had not been demonstrated. The TS Bases also stated that the prime i'mportance for the capability to close the block valve is to isolate a stuck-open PORV. In the SE, the NRC staff referenced a December 1997 Licensee Event Report that notified the NRC of the issue of potential failure of PSVs following water discharge.10 7 As part of this license amendment, the licensee upgraded the PORV circuitry, added additional PORV surveillance requirements, qualified the PORVs and associated piping for water discharge, and revised emergency procedures to allow plant operators additional time to terminate the event. With respect to the PORV circuitry, the NRC staff concluded that the PORV circuitry modifications qualifiied the PORV control circuitry as safety-related. With respect to PORV performance, the licensee reanalyzed the inadvertent SI event with the LOFTRAN computer code to determine the time available for operator action to make a PORV available and provide the mass and energy releases needed to qualify the PORVs and associated piping for water discharge. The licensee referenced EPRI testing that was said to generically resolve TMI Action Plan Items associated with PORVs and safety valve qualification for water and steam discharge, specifically the results from four tests of a Garrett PORV (such as used at Millstone 3) for water discharge.108 The licensee determined that the PORVs and associated piping are qualified for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of water discharge for an IOECCS event. The licensee also stated that the PORV manufacturer performed numerous cycle tests to verify the performance of the valve design, and also verified that valve seat leakage was acceptable. The licensee stated that the PORV block valves had been evaluated for water discharge in accordance with the program established in response to Generic Letter (GL) 89-10.109 The NRC staff found the licensee information regarding the qualification of the PORVs for water discharge during the inadvertent SI event to be acceptable.

106 NRC 1998 10 7 Northeast 1997 10a EPRI 1982a (Volume 11) 109 NRC 1989 Callaway In 2000, the NRG issued a license amendment for Callaway Plant, Unit 1 (Callaway) that revised the TS to change the PSV lift setting range. 110 The changes also credited automatic actuation of at least one PORV during an IOECCS event to prevent water discharge through the PSVs; to enable this credit, the licensee modified and upgraded the PORV circuitry to full Class 1E. In its license amendment request, 111 the licensee had stated that the design function of the valves was not being changed and the conclusions documented in the NRG staff's previous evaluation of Callaway's response to TMI Action Plan Item II.D.1112 were also unchanged. As a result, the licensee stated that the PORVs and associated discharge piping can accommodate water discharge.

Byron and Braidwood In 1998, the licensee for Byron and Braidwood requested an amendment to its TS to take credit for automatic operation of the PORVs to mitigate an IOECCS event. 113 In the amendment request, the licensee stated that the PSVs had not been qualified to reseat after passing subcooled liquid. The licensee stated that the PORVs at Byron and Braidwood are safety-related components with safety-related actuators and accumulator tanks, with PORV control circuits classified as safety-related. The licensee noted that some portions of the PORV circuitry are nonsafety-related, with improvements implemented in response to GL 90-06. 114 The licensee stated that the PORV block valves are within the scope of the GL 89-10 program.

In 1999, the NRG staff requested additional information related to concerns that the PORV circuitry did not meet the single failure criterion. 115 The licensee reevaluated its approach and withdrew its TS amendment request. 116 No further action regarding this amendment request was identified by the Panel. However, in a public meeting during the review of the NRR Appeal,117 the licensee stated that the PORVs and their block valves at Byron and Braidwood are safety-related with the exception of one circuitry aspect of the PORV. 118 In 2001, the NRC issued a license amendment for Byron and Braidwood to increase the maximum thermal power for each unit from 3411 megawatts thermal (MWt) to 3586.6 MWt (commonly referred !Q_as a stretch power uprate). 119 During its review, the NRG staff requested that the licensee address water solid conditions in the pressurizer, because Uthe NRG staff had generally not accepted a solid pressurizer for an IOECCS event given the potential for all three PSVs to be stuck open due to liquid relief through these safety valves. In response, the licensee stated that Section 15.5.1 , "Inadvertent Operation of Emergency Core Cooling System During Power Operation," of the UFSAR had been revised to credit the PSVs to pass water. 120 The licensee discussed the EPRI testing program in response to TMI Action Plan Item 11.D.1, with 110 NRC 2000 111 Union Electric 2000 112 NRC 1987b 11 3 ComEd 1998 114 NRC 1990b 115 NRC 1999 11e ComEd 1999 m Exelon 2015 118 NRC 2016a 119 NRC 2001 b 120 ComEd 2000b the results summarized in EPRI NP-2628-SR. 121 The licensee referenced previous NRC approvals related to TMI Action Plan Item 11.0.1 .122 The NRC staff made a further request regarding the temperature of water that would be discharged by the PSVs and the length of time that the PSVs would be expected to discharge water. The NRC staff also asked the licensee to discuss which EPRI tests are applicable to the Byron and Braidwood condition. In response, the licensee stated that the PSVs would close after discharging water, although they may not be leaktight. 123 The licensee stated that the leakage from up to three leaking PSVs is bounded by one fully open PSV. The licensee indicated that the EPRI testing of the Crosby safety valves in EPRI NP-2770-LD, Volumes 1 and 6,124 are applicable. The licensee indicated that valve chatter occurred during the tests with damage to the internals, but that the safety valve closed in response to system depressurization. The licensee stated that the Byron and Braidwood pressurizer water temperature of 590 °Fis higher than the EPRI tests (530 °F). The licensee stated that the assumed length of the event is 20 minutes from initial SI signal to when the system pressure is restored below PSV lift setpoint.

In Section 3.2 of the SE accompanying the license amendment, the NRC staff discussed its review of the performance of the PORVs and PSVs to discharge liquid water for approximately 20 minutes. The NRC staff discussed the EPRI testing program, with the conclusion that the PSV would close in response to system depressurization. The NRC staff reviewed the licensee's evaluation of the performance of the PSVs for liquid water conditions. The NRC staff found that the EPRI tests adequately demonstrated the performance of the valves for the expected water temperature conditions, and that there was reasonable assurance that the valves would adequately reseat following the spurious SI event. The NRC staff determined that EPRI test data indicated that the PSVs might chatter for the expected fluid inlet temperature, but that the resulting PSV seat leakage following the water discharge would be less than the discharge from one stuck-open PSV. Therefore, the NRC staff found the licensee's crediting of the PSVs to discharge liquid water during the spurious SI event to be acceptable. This portion of the SE was based on input provided by the Office of Nuclear Reactor Regulation (NRR) Reactor Systems Branch, with techn ical input from the responsible staff member for safety valves in the NRR Division of Engineering.12s As noted by the licensee, Section 15.5.1 of the Byron and Braidwood UFSAR at the time of the stretch power uprate includes PSV water discharge and references the TMI Action Plan Item 11.D.1 approvals. 126 The current UFSAR Revision 15 concludes that the IOECCS event does not progress into a stuck-open PSV LOCA event. 127 The UFSAR states that all three PSVs may lift but will reclose, and that the leakage is bounded by one fully open valve with the consequences bounded by the IOPSRV event. The UFSAR also specifies that if SI results in discharge of coolant through the pressurizer valves, the operators will bring the plant to cold shutdown to inspect the valves.

12 1 EPRI 1982b 122 NRC 1998c and NRC 1990a 12 3 Exelon 2001 124 EPRI 1982c and EPRI 1983 125 NRC 2001 a 126 Exelon 2002 127 Exelon 2014 In 2004, the NRC issued a license amendment for Byron and Braidwood granting an adjustment to the PSV setpoints. 128 As documented in the SE, the NRC staff requested during its review that the licensee perform a quantitative analysis regarding PSV water cycles and discharge water temperature. For the loss of ac power (LOAC) with reactor coolant pump (RCP) seal injection event, the licensee's analysis indicated that continued injection of water into the RCS through the RCP seals would result in a water-solid pressurizer and water discharge through the PSVs. The proposed PSV setpoint tolerance assuming negative tolerance would result in a lower PSV lift setpoint. With the lower setpoint. the PSV would open earlier, and a larger number of PSV water cycles with a lower water discharge temperature could result during the transient. The licensee performed an analysis of the LOAC with RCP seal injection event. and determined the revised PSV setpoint would result in an increase of about one PSV water cycle and a reduction in the water discharge temperature of about 0.5 °F. A comparison of the reanalysis showed that the spurious SI event remained the limiting event since it resulted in a greater increase in the number of PSV water cycles (two cycles vs. one cycle) and a greater decrease in the PSV discharge water temperature (3.0 °F vs. 0.5 °F) than that calculated for the LOAC with RCP seal injection event. The water discharge temperature in the analysis of record for the spurious SI event was 590 °F. The lowest discharge water temperature for the spurious SI event with the revised PSV setpoint was 587 °F. The NRC staff found that the calculated water discharge temperature (587 °F} was significantly higher than the discharge water temperature of 530 °F that was used to support operability of the PSVs as discussed in the analysis of record. As a result, the NRC staff concluded that the analysis was acceptable to assure that the PSVs will remain operable following a spurious SI event.

In 2014, the NRC issued a license amendment for Byron and Braidwood granting a measurement uncertainty recapture (MUR} power uprate.129 The NRC staff determined that the IOECCS event was outside of the scope of the MUR power uprate, because the licensee did not propose to modify the Chapter 15 analyses related to PSV and PORV water discharge.

With respect to inservice testing (1ST) activities, the Byron 1ST program 130 references the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code}, 2004 Edition through 2006 Addenda; and the Braidwood 1ST program 131 references the ASME OM Code, 2001 Edition through 2003 Addenda. The Byron 1ST Program specifies the following testing and intervals for the PORVs, PORV block valves, and PSVs:

  • PORV: fall safe test closed (cold shutdown interval}, stroke-time exercise open and closed (cold shutdown interval), and position indication test (2 year interval}
  • PORV Block Valve: exercise open and closed (2 year interval}; position indication test (Joint Owners Group (JOG) Program interval); an:d open and closed test in accordance with ASME OM Code Case OMN-1, "Alternative Rules for Preservice and lnservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants" (JOG Program interval) 128 NRC 2004b 129 NRG 2014a 130 Exelon 2016b 1 31 Exelon 2009
  • PSV: position indication test (2 year interval) and relief valve test (5 year interval),

referencing ASME OM Code, Appendix I, "lnservice Testing of Pressure Relief Devices in light-Water Reactor Nuclear Power Plants" The Braidwood 1ST Program specifies the following testing and intervals for the PORVs, PORV block valves, and PSVs:

  • PORV: fail safe test closed (refueling outage interval), stroke-time exercise open and closed (refueling outage interval), and position indication test (2 year interval).
  • PORV Block Valve: exercise open and closed (quarterly interval) and position indication test (2 year interval)
  • PSV: position indication test (2 year interval), and relief valve test (5 year interval),

referencing ASME OM Code, Appendix I Shearon Harris In 2001, the NRC issued a license amendment to Shearon Harris Nuclear Power Plant, Unit 1 (Shearon Harris) for steam generator replacement and a power uprate to a maximum power level of 2900 MWt (approximately 4.5 percent). 132 In addressing the licensee's evaluation of Standard Review Plan (SRP ) Section 15.5.1, the NRC staff found that the analysis showed that the calculated inlet pressures and temperatures required for the PORVs and safety relief valves (SRVs)133 to operate in a water environment were within the valve operable ranges, and thus ensured that the PGR\tPORVs and SRVSRVs would be operable during the transient. The valve operable ranges were previously determined by the licensee to support operability of the PGRVPORVs and SRVSRVs during the discharge of subcooled water in accordance with the TMI Action Plan Item 11.D.1 irequirements. Based on the analysis meeting the acceptance criteria of SRP Section 15.5. 1 with respect to the RCS pressure limit and departure-from-nucleate-boiling limit, the NRC staff concluded that the analysis was acceptable.

Beaver Valley In 2006, the NRC issued a license amendment authorizing an EPU for Beaver Valley Power Station, Units 1 and 2 (Beaver Valley), an approximate 8-percent increase in thermal power to 2,900 MWt. 134 In the SE accompanying the amendment, the NRC staff described its review of the capability of the PSVs to discharge liquid and adequately reseat for a spurious SI actuation.

The NRC staff specifically evaluated whether the PSVs could reasonably be expected to reseat to prevent the spurious SI actuation (an ANS Condition II event) from causing a stuck-open PSV (an ANS Condition Ill event). This issue was said to be further discussed in RIS 2005-29. While the PSVs for Beaver Valley were qualified to discharge steam, if the valves discharged water with sufficient subcooling, the NRC staff was concerned that they might not reseat properly.

Based the licensee's analysis, during a spurious SI event, the PSVs would be required to discharge steam followed by high temperature water after the pressurizer filled. The licensee provided plots of the pressuri.zer water temperatures for this event that indicated that the 132 NRC 2001 d 133 This term is used in the Shearon Harris SE. The licensee's RAI response (CP&L 2000) makes clear that the referenced SRVs and PORVs are pressurizer valves.

134 NRC 2006 minimum temperature of the discharged liquid for Beaver Valley was approximately 620 °F. To evaluate the capability of the valves to discharge and reseat, the NRC staff reviewed the available data from the full-flow tests performed during the EPRI test program in 1981 for the specific PSV models representative of those installed at Beaver Valley. The licensee also used the methodology contained in WCAP-11677 and determined that the minimum acceptable liquid temperature for which the PSVs were expected to successfully discharge and reseat was less than the minimum expected temperature for the spurious SI event for Beaver Valley.

The NRC staff agreed that both the minimum expected water discharge temperature and the minimum acceptable water temperature had been conservatively calculated. Therefore, the NRC staff determined that, for purposes of preventing the occurrence of a more serious ANS Condition Ill event, there was reasonable assurance that the PSVs would discharge water and reseat adequately following a spurious SI actuation. A consideration of the NRC staff in making this finding was that, in the unlikely event of a stuck-open PSV, the ECCS was fully capable of mitigating the resulting LOCA.

Turkey Point In 2012, the NRC issued a license amendment authorizing an EPU for Turkey Point Nuclear Generating, Units 3 and 4 (Turkey Point), increasing the thermal power level of each unit approximately 15 percent to 2644 MWt. 135 In the SE accompanying the amendment, the NRC staff indicated that ECCS actuation was not a possible Initiator of inadvertent increase in reactor coolant inventory because the high head SI pumps have a shut-off head below the normal RCS operating pressure. The NRC staff stated that a eves malfunction that increases RCS inventory was evaluated for the effects of adding water inventory to the RCS. If the pressurizer filled and caused water to be relieved through the PORVs or PSVs, then these valves could stick open and create a small break LOCA. The NRC staff stated that this would violate the acceptance criterion that prohibits the escalation of an anticipated operational occurrence (AOO) into a more serious event. Satisfaction of this acceptance criterion was demonstrated by showing that sufficient time would exist for the operator to recognize the situation and end the charging flow before the pressurizer could fill.

The NRC staff concluded that the licensee's analyses of IOECCS and eves events adequately accounted for operation of the plant at the proposed power level.

Regarding an inadvertent opening of a PORV, the licensee initially proposed that the consequences of this event were bounded by the small break LOCA. The NRG staff did not accept this proposed disposition. If action were not taken to secure the open valve by either closing the PORV or its block valve, the NRC staff stated that this event could escalate to a small-break LOCA, which would be contrary to the non-escalation position. When the pressurizer filled, water would begin to flow through the open PORV. If the PORV were not qualified for water discharge, the NRC staff stated that it was likely the PORV would not close upon demand. In this way, the NRC staff stated that the inadvertent opening of a PORV, an AOO, would become a small break-LOCA at the top of the pressurizer, an ANS Condition Ill event. The NRC staff requested that the licensee address the inadvertent opening of the PORV with respect to the third criterion for.an ANS Condition II event.

The licensee provided an analysis performed largely in accordance with NRG-approved, Westinghouse analytic methodology using the RETRAN computer code; however, this analysis 135 NRG 2012a was performed assuming that the PORV opened instead of the PSV. The NRC staff stated that assuming the opening of the PORV is acceptable, because the PSV is differently qualified, and reseats mechanically. An additional independent fault would be required to cause the PSV to fail to close. The analysis indicated that the pressurizer would fill within about 240 seconds. The licensee stated that there were multiple alarms to indicate the opening of a PORV. The licensee stated that a prompt operator action would be needed to close the PORV and, if the PORV does not close, the operator would be directed to close the block valve. Because the necessary actions would be prompt and simple, the NRC staff agreed that there would be sufficient time to secure the inadvertently open PORV without filling the pressurizer.

St. Lucie In 2012, the NRC issued a license amendment authorizing an EPU for St. Lucie Plant, Unit 2 (St. Lucie, Unit 2) that increased the authorized thermal power level about 12 percent to 3020 MWt.

Regarding an IOECCS event, the high pressure SI pumps would be incapable during power operations of delivering flow to the RCS because the pumps' shut-off head would be less than the normal RCS operating pressure of 2250 pounds per square inch absolute. Therefore, the licensee determined that the inadvertent operation of the ECCS at power event was not a credible event and did not analyze it for the proposed EPU. The NRC staff found that the licensee's position for not analyzing the IOECCS event to be acceptable.

Regarding a CVCS malfunction, the licensee evaluated it as an AOO for the effects of adding water inventory to the RCS. The NRC staff reviewed the licensee's analyses of the CVCS malfunction event and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff determined that the licensee*~ analysis analyses I rsM35ldemonstrated that the pressurizer did not become water solid, assuring no water was discharged through the PSVs.

Regarding an IOPORV event, the NRC staff stated that, when viewed from the mass addition perspective, this event could be evaluated in two phases: (1) an inadvertent opening of a pressurizer relief valve, followed by (2) an inadvertent ECCS actuation. In the first phase, the NRC staff stated that this event could be mitigated by closing the open PORV or its block valve.

If the PORV or its block valve was not closed, the NRC staff stated that the IOPORV event would enter the second pha:se with actuation of the ECCS. Based on its review, the NRC staff determined that the pressurizer overfill analysis, available alarming system, and procedures, in combination with simulator exercise results, provided reasonable assurance that the pressurizer would not be expected to fill to a water solid condition that could prevent the PORV or PSV from closing after they were open. The NRC staff therefore concluded that the event would not generate a more serious plant condition, meeting the non-escalation criterion. The NRC staff stated that it reviewed the licensee's analyses of the inadvertent opening of a pressurizer PORV event, and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models.

The NRC staff concluded that the licensee demonstrated that all AOO acceptance criteria were satisfactorily met.

North Anna North Anna Power Station, Units 1 and 2 (North Anna) UFSAR Section 15.2.14, "Spurious Operation of the Safety Injection System at Power," describes plant response to an inadvertent SI event. In particular, UFSAR Section 15.2.14.2.3, "Event Propagation," states the following:

Safety valve (Reference 18) and PORV (Reference 19) testing has revealed no instances of failure of the valves to reseat following water relief. Resulting leakage is within the capacity of the normal makeup system and is therefore not considered to be a small break loss of reactor coolant event. Therefore, the complete filling of the pressurizer and/or water relief via a safety valve as a result of a spurious safety injection does not constitute a failure to meet the event propagation acceptance criterion. Although primary credit for preventing the propagation of the event to a small break loss of reactor coolant event is the reseating of the PORVs and safety valves, It is noted that the PORVs (which open prior to the safety valves and, if open, preclude safety valve actuation for this event) are provided with block valves which the operator will close in the event of excessive PORV leakage.

North Anna UFSAR Section 15.2.14.3, "Conclusions," indicates that the complete filling of the pressurizer and water discharge via a PSV as a result of a spurious SI does not constitute a failure to meet the non-esca lation position. Furthermore, UFSAR Section 15.2, "References,"

lists EPRI NP-2770-LD and EPRI NP-2670-LD.

Conclusion In conclusion, the reliance by the licensee for Byron and Braidwood on the acceptable performance of the PSVs and PORVs following water discharge in response to abnormal events is not inconsistent with similar approaches by some other nuclear power plant licensees. In general, the review of activities by various nuclear power plant licensees related to PSV and PORV performance revealed reliance on EPRI, Wyle, and valve vendor testing to provide support for the performance of these valves under various service conditions. Specific certification for flow capacity of these valves for water discharge in accordance with the ASME BPV Code and National Board was not identified in the review of various justifications prepared by nuclear power plant licensees.

In evaluating the historical documents for Byron and Braidwood, the Panel found it challenging to determine specifically how the licensee resolved the concern raised in NSAL-93-013 in its analyses and plant operations. While the record does not currently support a compliance backfit in this case, if (as recommended by the Panel) the NRC staff undertakes a generic review of licensees' treatment of the potential for pressurizer valve damage following water discharge, it may be appropriate to consider what actions have been taken, how operating experience with water discharge has been considered, and how analysis assumptions are considered in operational practices (including inservice testing) at each plant.

APPENDIX D: REFERENCES

1. AEC 1970: Atomic Energy Commission (AEC), "Backfitting of Production and Utilization Facilities; Construction Permits and Operating Licenses," Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, published March 31 , 1970 (as amended).

Volume 35 of the Federal Register (FR), page 5317.

2. AEC 1971: AEC, "General Design Criteria for Nuclear Power Plants," 10 CFR Part 50, Appendix A, first published February 20, 1971 (as amended). 36 FR 3256.
3. ANS 1973: American Nuclear Society (ANS), ANS-51.1/N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," dated August 6, 1973. [Publicly available from ANS for a fee.]
4. ANS 1981: ANS and American National Standards Institute (ANSI), ANSI/ANS 58.9, "Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems," issued February 17, 1981 (updated in 2002 and reaffirmed in 2015). [Publicly available from ANS for a fee.]
5. ANS 2016: ANS, "American Nuclear Society Standards Committee Glossary of Definitions and Terminology," dated May 19, 2016. Available from ANS at http:/led n .ans.org/standards/resources/toolkit/docs/glossarv-of-defin itions. pdf.
6. ComEd 1982: Commonwealth Edison (ComEd}, letter from T. R. Tramm to Harold R. Denton, U.S. Nuclear Regulatory Commission (NRC), "Byron Station Units 1 and 2; Braidwood Station Units 1 and 2: Pressurizer Safety and Relief Valves," dated October 26, 1982. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8211020633 and microfiche location 15886:171 -15886:231.]
7. ComEd 1998: Commonwealth Edison Company, letter from K.L. Grasser to U.S. NRG, "Change to Credit Automatic PORV Operation for Mitigation of Inadvertent Safety Injection at Power Accident," dated May 29, 1998. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 9806020180 and microfiche location A3765:275-301.]
8. ComEd 1999: Commonwealth Edison Company, letter from RM. Krich to U.S. NRG, "Response to Request for Additional Information Regarding License Amendment Request to Credit Automatic Power-Operated Relief Valve (PORV) Operation for Mitigation of Inadvertent Safety Injection at Power Accident and Withdrawal of License Amendment Request," dated July 16, 1999. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 9907200183 and microfiche location A8671 :349-354.]
9. ComEd 2000a: Commonwealth Edison Company, letter from RM. Krich to U.S. NRC, "Request for a License Amendment to Permit Uprated Power Operations at Byron and Braidwood Stations," dated July 5, 2000. ADAMS Accession No. ML003730608.
10. ComEd 2000b: Commonwealth Edison Company, letter from RM. Krich to U.S. NRG, "Response to Request for Additional Information Regarding the License Amendment Request to Permit Uprated Power Operations at Byron and Braidwood Stations," dated November 27, 2000. ADAMS Accession No. ML003772461.

11 . Consumers 1982: Consumers Energy Company, letter from D. P. Hoffman to Harold R. Denton, U.S. NRC, "Safety and Relief Valve Test Report for the EPRI PWR Safety and Relief Valve Test Program," dated April 1, 1982. (This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8207160337 and microfiche location 13866:001 - 13869: 106.]

12. CP&L 2000: Carolina Power & light Company, letter from James Scarola to U.S. NRC, "License Amendhlent Application Power Uprate," dated December 14, 2000. ADAMS Accession No. ML003780761.
13. EPRI 1982a: EPRI, NP-2670-LD, "EPRI/Wyle Power-Operated Relief Valve Phase Ill Test Report," dated October 1982. [Publicly available from EPRI for a fee.]
14. EPRI 1982b: EPRI, NP-2628-SR, "EPRI PWR Safety and Relief Valve Test Program -

Safety and Relief Valve Test Report," dated December 1982. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8407130197 and microfiche location 25588:082-262.]

15. EPRI 1982c: Electric Power Research Institute (EPRI), NP-2770-LD, Volume 1, "EPRI/C-E PWR Safety Valve Test Report: Summary," dated December 1982. Publicly available from EPRI.
16. EPRI 1983: Electric Power Research Institute (EPRI), NP-2770-LD, Volume 6, "EPRI/C-E PWR Safety Valve Test Report: Test Results for Crosby Safety Valve," dated March 1983. [Publicly available from EPRI for a fee.]
17. EPRI 2004: EPRI, TR-1011047, "Probability of Safety Valve Failure-to-Reseat Following Steam and Liquid Relief - Quantitative Expert Elicitation," dated August 2004. Publicly available from EPRI.
18. Exelon 2001: Exelon Generation Company, LLC, letter from R.M. Krich to U.S. NRC, "Response to Reque*st for Additional Information Regarding the License Amendment Request to Permit Uprated Power Operations at Byron and Braidwood Stations," dated January 31, 2001. ADAMS Accession No. ML010330145.
19. Exelon 2002: Exelon Generation Company, LLC, letter from R. M. Krlch to U.S. NRC, "Byron/Braidwood Stations Updated Final Safety Analysis Report, Revision 9; Byron Station Technical Requirements Manual, Revision 32; Byron Station Technica, Specifications Bases, Revision 31; Braidwood Station Technical Requirements Manual, Revision 26; Braidwood Station Technical Specifications Bases, Revision 36/ dated December 16, 2002. [Non-publicly available; electronic files are in the NRC file center, and the transmittal letter is at ADAMS Accession No. ML023650683.]
20. Exelon 2009: Exelon Generation Company, LLC, letter from Bryan Hanson to U.S.

NRC, "Braidwood Station, Units 1 and 2 - Submittal of lnservice Testing Program Plan for the Third Ten-Year Interval," dated July 27, 2009. ADAMS Accession No. ML092110538.

21 . Exelon 2014: Exelon Generation Company, LLC, "Byron/Braidwood Nuclear Stations Updated Final Safety Analysis Report (UFSAR)," Revision 15, dated December 2014.

ADAMS Accession No. ML14363A393.

22. Exelon 2015: Exelon Generation Company, LLC, letter from J. Bradley Fewell to William M. Dean, U.S . NRC, "Appeal of Imposition of Backfit Regarding Compliance with Title 1O of the Code of Federal Regulations (10 CFR) Section 50.34(b), General Design Criteria (GDC) 15, GDC 21, GDC 29, and Licensing Basis," dated December 8, 2015. ADAMS Accession No. ML15342A112.
23. Exelon 2016a: Exelon Generation Company, LLC, Jetter from J. Bradley Fewell to Victor M. Mccree, U.S. NRC, "Appeal of Imposition of Backfit Regarding Compliance with 10 CFR § 50.34(b), General Design Criteria (GDC) 15, GDC 21, GDC 29, and Licensing Basis," dated June 2, 2016. ADAMS Accession No. ML16154A254.
24. Exelon 2016b: Exelon Generation Company, LLC, letter from Thomas D. Chalmers to U.S. NRC, "Byron Station, Units 1 and 2, Transmittal of lnservice Testing Program for the Fourth Ten-Year Interval," dated July 21 , 2016. ADAMS Accession No. ML16203A108.
25. IAEA 2009: International Atomic Energy Agency (IAEA), IAEA-TECDOC-1624, "Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants," dated November 2009. Publicly available from IAEA.
26. Northeast 1997: Northeast Nuclear Energy Company, Licensee Event Report 97-063-00, "Millstone Nuclear Power Station Unit 3 - Inadequate Operator Response Time for Inadvertent Safety Injection (SI) Event," dated December 31, 1997.

Text available through ADAMS Public Legacy Library using Accession No. 9802110033.

27. NEI 2016: Nuclear Energy Institute (NEI), letter from Anthony R. Pietrangelo to Victor M. Mccree, U .S. NRC, "Nuclear Energy Institute Comments in Support of Exelon Generation Company Second-Level Backfit Appeal," dated June 16, 2016. ADAMS Accession No. ML16208A008.
28. NRC 1976: U.S. NRC, NUREG-0053, "Safety Evaluation Report Related to the Operation of North Anna Power Station," dated April 1976. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8001100821 and microfiche location 04043:051-278.]
29. NRC 1977: U.S. NRC, SECY-77-439, "Single Failure Criterion," dated August 17, 1977.

ADAMS Accession No. ML060260236.

30. NRC 1979a: U.S. NRC, NUREG-0578, "TMl-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," dated July 1979. ADAMS Accession No. ML090060030. {Transmitted to licensees via Generic Letter 79-32, "TMl-2 Lessons Learned Task Force Report - NUREG-0578," dated July 25, 1979. Publicly available on the NRC website.)

31 . NRC 1979b: U.S. NRC, NUREG-0585, "TMl-2 Lessons Learned Task Force Final Report," dated October 1979. ADAMS Accession No. ML061430367.

32. NRC 1980a: U.S. NRC, NUREG-0660, "NRC Action Plan Developed as a Result of the TMl-2 Accident," dated May 1980. ADAMS Accession Nos. ML072470526 and ML072470524.
33. NRC 1980b: U.S. NRC, letter from Darrell G. Eisenhut, "Post-TMI Requirements," dated October 31, 1980. Text available through ADAMS Public Legacy Library using Accession No. 8012160050.
34. NRC 1980c: U.S. NRC, NUREG-0737, "Clarification of TMI Action Plan Requirements,"

dated November 1980. ADAMS Accession No. ML051400209.

35. NRC 1981a: U.S. NRC, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" [SRPJ, Section 15.0, "Introduction," Revision 2, dated July 1981. ADAMS Accession No. ML052350113.
36. NRC 1981b: U.S. NRC, NUREG-0800, SRP Section 15.5.1 -15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," Revision 1, dated July 1981. ADAMS Accession No. ML052350142.
37. NRC 1981c: U.S. NRC, NUREG-0800, SRP Section 15.6.1, "Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve," Revision 1, dated July 1981 . ADAMS Accession No. ML052350144.
38. NRC 1982: U.S. NRC, memorandum from R.L. Tedesco to RF. Fraley, "Issuance of the Draft Safety Evaluation Report for the Watts Bar Nuclear Plant, Units 1 and 2," dated January 5, 1982. ADAMS Accession No. ML073580524 [non-public].
39. NRC 1983: U.S. NRC, NUREG-1002, "Safety Evaluation Report Related to the Operation of Braidwood Station, Units 1 and 2," dated November 1983. ADAMS Accession No. Ml12114A272.
40. NRC 1984: U.S. NRC, NUREG-0876, Supplement No. 5, "Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2," dated October 1984. ADAMS Accession No. ML091340252.
41. NRC 1985: U.S. NRC, "Revision of Backflttlng Process for Power Reactors - Parts 2 and 50," dated September 20, 1985. 50 FR 38097 .
42. NRC 1986a: U.S. NRC, letter from H. Schierling to J.D. Shiffer, Pacific Gas and Electric Company (PG&E), "Safety and Relief Valve Testing, NUREG-0737 Item 11.D.1," dated January 27, 1986. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8602180450 and microfiche location 34630:271 -299.)
43. NRC 1986b: U.S. NRC, NUREG-1002, Supplement No. 1, "Safety Evaluation Report Related to the Operation of Braidwood Station, Units 1 and 2," dated September 1986.

[This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8610220157 and microfiche location 38332:281 -

38333:087.)

44. NRC 1987a: U.S. NRC, NUREG-0876, Supplement No. 8, "Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2," dated March 1987. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8704010167 and microfiche location 40321 : 134-156.]
45. NRC 1987b: U.S. NRC, letter from Thomas W. Alexion to D.F. Schnell, Union Electric Company, dated September 10, 1987. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8709170429 and microfiche location 42712:212-229.]
46. NRC 1988a: U.S. NRC, SECY-88-102, "Final Backfit Rule," dated April 18, 1988. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8805130269 and microfiche location 45795:299-362.]
47. NRC 1988b: U.S. NRC, "Revision of Backfitting Process for Power Reactors- Final Rule," dated June 6, 1988. 53 FR 20603.
48. NRC 1988c: U.S. NRG, letter from L. Olshan to Henry E. Bliss, Commonwealth Edison Company, "NUREG-0737, Item 11.D.1 , Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and 2," dated August 18, 1988. ADAMS Accession No. ML003772409 (pages 161- 188 of file].
49. NRC 1988d: U.S. NRC, NRC Manual Chapter 0514, "NRG Program for Management of Plant-Specific Backfitting of Nuclear Power Plants," dated August 26, 1988. ADAMS Accession No. ML041400111.
50. NRC 1988e: U.S. NRC, Bulletin 88-11 , "Pressurizer Surge Line Thermal Stratification,"

dated December 20, 1988. Publicly available on the NRC website.

51. NRC 1989: U.S. NRC, Generic Letter (GL) 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," dated June 28, 1989. ADAMS Accession No. ML031150300.
52. NRC 1990a: U .S. NRC, letter from S. Sands to Thomas J. Kovach, Commonwealth Edison Company, "NUREG-0737, Item 11.D.1 , "Performance Testing on Relief and Safety Valves for Braidwood Station, Units 1 and 2," dated May 21 , 1990. ADAMS Accession No. ML003772409 [pages 189-217 of file].
53. NRC 1990b: U.S. NRC, GL 90-06, "Resolution of Generic Issue 70, 'Power-Operated Relief Valve and Block Valve Reliability' and Generic Issue 94, 'Additional Low-Temperature Over Pressure Protection for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f)," dated June 25, 1990. ADAMS Accession No. ML031210416.
54. NRC 1990c: U .S. NRC, NUREG-1409, "Back.fitting Guidelines," dated July 1990.

ADAMS Accession No. ML032230247.

55. NRC 1991: U.S. NRC, letter from Anthony H. Hsia to Thomas J. Kovach, Commonwealth Edison Company, "Issuance of Amendments [for Byron and Braidwood]

(TAC Nos. M77332, M77333, M77334, M77335, M77402, M77403, M77404, and M77405)," dated November 18, 1991. ADAMS Accession No. ML020860105.

56. NRC 1994a: U.S. NRC, SECY-94-084, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs," dated March 28, 1994. ADAMS Accession No. ML003708068.
57. NRC 1994b: U.S. NRC, SRM-SECY-94-084, "SECY-94-084 - Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs and COMSECY-94-024 - Implementation of Design Certification and Light-Water Reactor Design Issues," dated June 30, 1994. ADAMS Accession No. ML003708098.
58. NRC 1996: U.S. NRC, NUREG-0800, SRP Section 15.5.1 - 15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," draft Revision 2, dated April 1996. ADAMS Accession No. ML052070725.
59. NRC 1997: U.S. NRC, letter from Leonard N. Olshan, NRC, to Leon R. Eliason, Public Service Electric & Gas Company, "Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. M97827 and M97828}," dated June 4, 1997. ADAMS Accession No. ML011720397.
60. NRC 1998: U.S. NRC, letter from James W. Andersen to Martin L. Bowling, Jr.,

Northeast Nuclear Energy Company, "Issuance of Amendment - Millstone Nuclear Power Station, Unit No. 3 {TAC No. MA1527)," dated June 5, 1998. ADAMS Accession No. ML011800207.

61 . NRC 1999: U.S. NRC, letter from John B. Hickman to Oliver D. Kingsley, Commonwealth Edison Company, "Request for Additional Information - Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2 (TAC Nos. MA2043, MA2044, MA2048, and MA2049," dated May 13, 1999. [This document is not In ADAMS, but is available through the NRC Public Document Room using Accession No. 9905170241 and microfiche location A8035:313-316.J

62. NRC 2000: U.S. NRC, letter from Jack Donohew to Garry L. Randolph, Union Electric Company, "Callaway Plant, Unit 1 - Issuance of Amendment Re: Pressurizer Safety Valves and Power Operated Relief Valves (PORVs) (TAC No. MA9080}," dated September 26, 2000. ADAMS Accession No. ML003753326.
63. NRC 2001a: U.S. NRC, memorandum from Frank Akstulewicz to Anthony Mendiola, "Byron and Braidwood Stations, Units 1 and 2 - Requests for a License Amendment to Permit Uprated Power Operations (TAC Nos. MA9426, MA9427, MA9428 and MA9429}," dated March 15, 2001. ADAMS Accession No. ML010740316 [non-public].
64. NRC 2001b: U.S. NRC, letter from George F. Dick, Jr., to Oliver D. Kingsley, Exelon Generation Company, LLC, "Issuance of Amendments; Increase in Reactor Power, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (TAC Nos. MA9428, MA9429, MA9426, and MA9427)," dated May 4, 2001 . ADAMS Accession No. ML033040016.
65. NRC 2001c: U.S. NRC, NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," dated July 2001. ADAMS Accession No. ML012070413.
66. NRC 2001d: U.S. NRC, letter from N. Kalyanam to James Scarola, Carolina Power &

Light Company, "Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Steam Generator Replacement and Power Uprate (TAC Nos. MB0199 and MB0782)," dated October 12, 2001. ADAMS Accession No. ML012880381.

67. NRC 2003: U.S. NRC, Review Standard (RS) 001, "Review Standard for Extended Power Uprates," dated December 2003. ADAMS Accession No. ML033640024.
68. NRC 2004a: U.S. NRC, letter from Girija S. Shukla to Gregory M. Rueger, PG&E, "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment Re: Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves (TAC Nos. MB6758 and MB6759)," dated July 2, 2004. ADAMS Accession No. ML041950300.
69. NRC 2004b: U.S. NRC, letter from George F. Dick, Jr., to Christopher M. Crane, Exelon Generation Company, LLC, "Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 - Issuance of Amendments, Re: Pressurizer Safety Valve Setpoints, (TAC Nos. MB9762, MB9763, MB9760, and MB9761 )," dated August 26, 2004. ADAMS Accession No. ML042250531 .
70. NRC 2005a: U.S. NRC, SECY-05-0138, "Risk-Informed and Performance-Based Alternatives to the Single-Failure Criterion," dated August 4, 2005. ADAMS Accession No. ML051950610.
71. NRC 2005b: U.S. NRC, Regulatory Issue Summary 2005-29, "Anticipated Transients that Could Develop into More Serious Events," dated December 14, 2005. ADAMS Accession No. ML051890212.
72. NRC 2006: U.S. NRC, letter from Timothy G. Colburn to James H. Lash, FirstEnergy Nuclear Operating Company, "Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) - Issuance of Amendment Regarding the 8-Percent Extended Power Uprate (TAC Nos. MC4645 and MC4646)," dated July 19, 2006. ADAMS Accession No. ML061720274.
73. NRC 2007a: U.S. NR.C, NUREG-0800, SRP Section 15.0, "Introduction - Transient and Accident Analyses," Revision 3, dated March 2007. ADAMS Accession No. ML070710376.

74 . NRC 2007b: U.S. NRC, NUREG-0800, SRP Section 15.5.1 - 15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory," Revision 2, dated March 2007. ADAMS Accession No. ML070820081.

75. NRC 2007c: U.S. NRC, NUREG-0800, SRP Section 15.6.1, "Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve," Revision 2, dated March 2007. ADAMS Accession No. ML070820094.
76. NRC 2011: U.S. NRC, NUREG/CR-7037, "Industry Performance of Relief Valves at U.S. Commercial Nuclear Power Plants through 2007," dated March 2011. ADAMS Accession No. ML110980205.
77. NRC 2012a: U.S. NRC, letter from Jason C. Paige to Mano Nazar, Florida Power and Light Company, "Turkey Point Units 3 and 4- Issuance of Amendments Regarding Extended Power Uprate (TAC Nos. ME4907 and ME4908)," dated June 15, 2012.

ADAMS Accession No. ML11293A359.

78. NRC 2012b: U.S. NRC, letter from Tracy J. Orf to Mano Nazar, Florida Power and Light Company, "St. Lucie Plant, Unit 2 - Issuance of Amendment Regarding Extended Power Uprate (TAC No. ME5843)," dated September 24, 2012. ADAMS Accession No. ML12268A132.
79. NRC 2013: U.S. NRC, Management Directive 8.4, "Management of Facility-Specific Backfitting and Information Collection," dated October 9, 2013. ADAMS Accession No. ML12059A460.
80. NRC 2014a: U.S. NRC, letter from Joel S. Wiebe to Michael J. Pacilio, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. MF2418, MF2419, MF2420, and MF2421)," dated February 7, 2014. ADAMS Accession No. ML 13281AOOO.

81 . NRC 2014b: U.S. NRC, memorandum from Samuel Miranda to Christopher P. Jackson, "Making Non-Concurrence NCP-2013-04 Public," dated February 28, 2014. ADAMS Accession No. ML14063A174.

82. NRC 2015a: U.S. NRC, Draft Revision 1 to RIS 2005-29, "Anticipated Transients that Could Develop into More Serious Events," dated July 13, 2015. ADAMS Accession No. ML15014A469. (Also published at for public comment at 80 FR 42559.)
83. NRC 2015b: U.S. NRC, letter from Anne T. Boland to Bryan Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Backfit Imposition Regarding Compliance with 10 CFR § 50.34(b), GDC 15, GDC 21, GDC 29, and Licensing Basis (TAC Nos. MF3206, MF3207, MF3208, and MF3209)," dated October 9, 2015. ADAMS Accession No. ML14225A871.
84. NRC 2016a: U.S. NRC, "Official Transcript of Proceedings - Public Meeting to Discuss Exelon Generating Company, LLC's Appeal of Compliance Backfit Affecting Byron and Braidwood Generating Stations," dated March 7, 2016. ADAMS Accession No. ML16070A364.
85. NRC 2016b: U.S. NRC, memorandum from Anthony T. Gody, Jr., to Marissa G. Bailey, "Input for Exelon Backfit Review Panel," dated March 21 , 2016. ADAMS Accession No. ML16081A405 [non-public].
86. NRC 2016c: U.S. NRC, memorandum from Marissa G. Bailey to William M. Dean, "Backfit Review Panel Recommendation Regarding Exelon Appeal of Backfit Affecting Byron and Braidwood Stations Regarding Compliance with 10 CFR 50.34(b), GDC 15, GDC 21, GDC 29, and the Licensing Basis," dated March 25, 2016. ADAMS Accession No. ML16082A542 [non-public].
87. NRC 2016d: U.S. NRC, letter from William M. Dean to J. Bradley Fewell, Exelon Generation Company, LLC, "U.S. Nuclear Regulatory Commission Response to Backfit Appeal - Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2," dated May 3, 2016. ADAMS Accession No. ML16095A204.
88. NRC 2016e: U.S. NRC, memorandum from Victor M. Mccree to Gary M. Holahan, K.

Steven West, Thomas G. Scarbrough, and Michael A. Spencer, "Charter for Backfit Appeal Review Panel Associated with Byron and Braidwood Compliance with 1 OCFR 50.34(b), GDC 15, GDC 21, GDC 29, and the Licensing Basis, " dated June 22, 2016.

ADAMS Accession No. ML16173A311 [non-public].

89. NRC 2016f: U.S. NRC, "An Assessment of Core Damage Frequency for Byron/Braidwood Nuclear Power Plants Supporting Backfit Appeal Review Panel," dated August 11 , 2016. ADAMS Accession No. ML16214A199 [non-public].
90. PG&E 1996: PG&E, letter from Gregory M. Rueger to U.S. NRC, "Forwards completed Licensing Basis Impact Evaluation of FSAR Update change which contains SE performed JAW 10CFR50.59 re reanalysis of inadvertent ECCS actuation accident,"

dated August 13, 1996. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 9608200112 and microfiche location 89419:294-322.]

91. PG&E 2003: Pacific Gas and Electric Company, letter from David H. Oatley to U.S. NRC, "Response to NRC Request for Additional Information Regarding License Amendment Request 01-018, 'Credit for Automatic Actuation of Pressurizer Power Operated Relief Valves; Pressurizer Safety Valve Loop Seal Temperature,"' dated November 21, 2003. ADAMS Accession No. ML033360735.
92. TVA 1983: Tennessee Valley Authority (TVA), letter from L.M. Mills to E. Adensam, U.S. NRC, enclosing "Watts Bar Nuclear Plant Units 1 and 2 Safety Valve Sizing," dated April 18, 1983. ADAMS Accession No. ML080360226.
93. Union Electric 2000: Union Electric Company, letter from Alan C. Passwater to U.S. NRC, "Revision to Technical Specifications 3.3.2, 3.4.10, and 3.4.11 - Pressurizer Safety Valves and PORVs," dated May 25, 2000. ADAMS Accession No. ML003719636.
94. VEPCO 2009: Virginia Electric and Power Company (VEPCO), letter from L.N. Hartz to U.S. NRC, "North Anna Power Station Units 1 and 2 Updated Final Safety Analysis Report Revision 45," dated October 1, 2009. ADAMS Accession No. ML092810047.
95. VEPCO 2015: Virginia Electric and Power Company, letter from Gianna C. Clark to U.S. NRC, "North Anna Power Station Units 1 and 2 Updated Final Safety Analysis Report Revision 51 ," dated September 30, 2015. ADAMS Accession No. ML15296A098

[non-public].

96. Westinghouse 1988: Westinghouse, R.J. Dickinson and J.G. Bass, "Pressurizer Safety Relief Valve Operation for Water Discharge during a Feedwater Line Break,"

WCAP-11677, dated January 1988. (Submitted to the NRC as part of a May 8, 1989, response to a request for additional information related to Seabrook.) [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8905120191 and microfiche location 49755:336- 49756:017.]

97. Westinghouse 1993: Westinghouse, Nuclear Safety Advisory Letter (NSAL)93-013, "Inadvertent ECCS Actuation at Power," dated June 30, 1993. [This document is not in public ADAMS, but is available through the NRC Public Document Room using Accession No. 9608200112 and microfiche location 89419:311-315, as well as in non-public ADAMS Accession No. ML052930330, pages 2-6 of file.]
98. Westinghouse 1994: Westinghouse, NSAL-93-013, Supplement 1, "Inadvertent ECCS Actuation at Power," dated October 28, 1994. ADAMS Accession No. ML050320117

[pages 9-15 of file].

99. Westinghouse 2000: Westinghouse, NSAL-00-013, "CVCS Modeling Assumption for Loss of Offsite Power Analyses," dated August 23, 2000. ADAMS Accession No. ML103200150 [pages 131-139 offile].

100. Westinghouse 2007: Westinghouse, NSAL-07-10, "Loss-of-Normal Feedwater Loss-of-Offsite AC Power Analysis PORV Modeling Assumptions," dated November 7, 2007. ADAMS Accession No. ML100140163 [pages 23-26 of file].

101. Westinghouse 2011: Westinghouse, "AP1000 Design Control Document," Revision 19, dated June 13, 2011 . ADAMS Accession No. ML11171A500.

102. WOG 1982: Westinghouse Owners Group (WOG), letter from Oliver D. Kingsley, Alabama Power Company, to Harold R. Denton, U.S. NRC, "NUREG-0737, Item 11.0.1,

'Pressurizer Safety Valve Operability,'" dated July 27, 1982. Forwards Westinghouse WCAP-10105, "Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program," dated June 1982. [This document is not in ADAMS, but is available through the NRC Public Document Room using Accession No. 8208190307 and microfiche location 14387:189-301 .]

~PPENDIX lcr3a1E: LIST OF ABBREVIATIONS From: Spencer, Michael Sent: Monday, July 18, 2016 9:14 AM To: Holahan, Gary; West, Steven; Scarbrough, Thomas; Clark, Theresa

Subject:

Backfit Appeal Panel Report 2016 07 13 - MAS INPUT TO 7-15 2.35PM DRAFT 2.docx Attachments: Backfit Appeal Panel Report 2016 07 13 - MAS INPUT TO 7-15 2.35PM DRAFT 2.docx

All, As we discussed at the last meeting, I have added input on the backfit rule and compliance exception. The input begins on page 2 of the attached. This file is also saved in my working folder.

Michael

Background

On July 22,2016 in accordance with NRC Management Directive (MD) 8.4, "Management of Facility-specific Backfitting and Information Collection," the NRC Executive Director for Operations (EDO) established a Backfit Appeal Review Panel (Panell to review the appeal by Exelon Generation Company, LLC {Exelon or the licensee) of the U.S. Nuclear Regulatory Commission (NRC) staff's determination that a backfit is necessary at Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Units 1 and 2 (Byron), as well as the staffs application of the compliance backfit exception provided in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109.

The specific NRC staff activity at issue relates to an October 9, 2015, letter, in which the NRC issued the results of a staff review of licensing basis documents for Braidwood and Byron (Agencywide Documents Access and Management System (ADAMS) Accession No. ML1422SA871). The staff determined that Braidwood and Byron were not in compliance with 10 CFR Appendix A, General Design Criteria IGDCs) 15, 21, and 29; 10 CFR S0.34(b); and the plant-specific design bases. Specifically, Braidwood and Byron were determined not to comply with provisions for ensuring that Condition II events (analyses of inadvertent operation of the emergency core cooling system (ECCS), malfunction of the chemical and volume control system, and inadvertent opening of a pressurizer safety or relief valve) do not progress into more serious Condition Ill events following water relief through certain valves. The staff acknowledged that the staff position differed from a previous staff position documented in a 2001 power uprate safety evaluation. However, the staff determined that the backfitting was justified under the compliance exception in 10 CFR 50.109(a)(4)(i). The licensee was directed to take action to resolve the non-compliance.

On December 8, 2015, the licensee appealed the staff's decision stating its disagreement with the NRC's conclusion that the compliance exception to the backfit rule applies in this case, and that the NRC has twice approved the underlying analysis {ADAMS Accession No. ML15342Al12). The referenced approvals were an August 26, 2004, license amendment associated with pressurizer safety valve setpoints (ADAMS Accession No. ML042250531) and a May 4, 2001, license amendment associated with a stretch power uprate (ADAMS Accession No. ML033040016). In a letter dated May 3, 2016, the NRC responded to the licensee's appeal and reaffirmed its decision that the backfit per the compliance exception provisions of 10 CFR S0.109(al(4)(i) issued to the licensee is appropriate (ADAMS Accession No. ML16095A204).

On June 2, 2016, the licensee again appealed the staffs decision (ADAMS Accession No. ML16154A254).

The purpose of this report is to provide information and recommendations to support the decision of the EDO.

Conduct of the review In order to establish a sound, well informed, and .... basis for its recommendations, the Backfit Appear Panel undertook a review of the relevant documents in this case. This included the licensee and NRC staff letters mentioned above plus the 2001 power uprate and the 2004 license amendment plus a June 16, 2016 letter form the Nuclear Energy Institute supporting the Exelon backfit appeal ADAMS XXXXXX.XX). The panel also reviewed numerous other documents related to the topic of inadvertent operation of the ECCS. These documents fall into six broad categories:

1

the Back.fit Rule (10CFR50.109), related court actions, and Commission and staff guidance on application of the Backfit Rule; licensing activities on Byron and Braidwood (License Amendment Requests, License Amendments, staff Requests for Additional Information, licensee responses, meeting summaries, staff Safety Evaluation Reports, and Updates of the Final Safety Analysis Report UFSAR)) aver the period of 1997 to the present; the NRC guidance relevant to the analysis of inadvertent operation of the ECCS (Standard Review Plan (SRP) sections 15.0, 15.5.1, and 15.6.1) over the period of 1981 to the present; the Westinghouse Nuclear Safety Advisory Letter (NSAL-93-013, June 30, 1993) and its supplement (NSAL-93-13 Suppl. 1, October 28, 1994)(Adams XXXXXXXXXXX);

actions taken by other licensees in response to Westinghouse NSAL-93-013; the history of NRC and industry activities related to Power Operated Relief Valves (PORV), their Block Valves, and Safety Valves (including Three Mile Island Action Plan (NU REG 0737) items 11.D.1, 11.0.3,11.G.1, 11.K.3, and Generic Letter 89-10 and supplements), and related Electric Power Research Institute (EPRI) valve testing, and operating experience (NUREGXXXX .... ).

ln addition to the document review, the panel had the benefit of meetings with the Office of Nuclear Reactor Regulation, the Office of the General Council, and the NRC Committee to Review Generic Requirements (CRGR). Both Exelon and NEI declined offers for a public meeting, but indicated a willingness and interest in providing information if the panel identified the need.

Assistance from RES, TIC etc. . .................. If any HistePJ of the Backfit Rule and the Compliance ExceptionJMAS11 Backfittinq is defined by 10 CFR 50.109{a) as:

the modification of or addition to systems. structures, components, or design of a facility; or the design approval or manufacturing license for a facility: or the procedures or organization required to design, construct or operate a facility: any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position ....

Unless one of three specified exceptions apply, the NRG may impose a backfrt only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50.109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived 2

from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection.*

Section 50.109(a}(4 l sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2}

and (a}(3l. The first exception. the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee.* 10 CFR 50.109(a)(4)(i}. The second and third exceptions relate to actions necessary to ensure adequate protection or to actions that involve defining or redefining adequate protection.

The Commission explained its intended application of the compliance exception in the 1985 final rule amending 10 CFR 50.109 (50 FR at 38103):

The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.

In the 1985 rule, the Commission acknowledged that staff interpretations of rules are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal factor in initiating a backfit." Id. at 38102. 1

!By its rMAS21terrns 1 the compliance exception applies to actions necessary for compliance with rules, licenses, and orders, or for conformance with written commitments. 2 Also, the Commission explicitly acknowledged the importance of staff interpretations of rules in the regulatory process. Thus, the Panel understands the term "known and established standard" to include standards established in rules, licenses, orders, and written commitments, and NRC interpretations of rules that might be announced in a variety of forms. Some standards mav be broad-based while others may apply only to a limited number of plants. As stated in NUREG-1409. "[ilnformal or formal communications to one licensee are not official positions to all licensees .... Orders. licenses, and written commitments are applicable only to a particular licensee."

The failure to meet a known and established standard is grounds for a compliance back.fit if this failure is due to "omission or mistake of fact." Thus, if a licensee obtains NRC approval of an alternative to a standard set forth in guidance. the guidance could not be used to support a compliance backfit unless the NRC's approval of the alternative was based on an omission or mistake of fact. Omissions and mistakes of fact are to be distinguished from "new or modified interpretations of what constitutes compliance." which do not fall within the compliance 1 The 1985 backfit rule was vacated by a Federal court on grounds unrelated to the compliance backfit exception. See Union of Concerned Scientists v. U.S. NuclearRequlatqyCom'n, 824 F.2d 108. 119*20 (D.C. Cir. 1987). In 1988, the Commission amended the backfrt rule (53 FR 20603) to address the court's concerns. but did not change the 1985 rule's compliance exception provision. Thus. the quoted statements from the 1985 rule are the applicable expression of Commission intent regarding compliance backfits.

2 NUREG-1409, Backfitting Guidelines, defines written commitments broadly to include the "final safety analysis report, licensee event reports. and docketed correspondence, including responses to NRC bulletins. generic letters, inspection reports, or notices of violation and confirmatory action letters."

3

exception. The Panel understands the term "new or modified interpretations" to include situations where the NRC has changed its mind on how to interpret the language of a requirement or on how much assurance is necessary to conclude that the requirement is met.

Levels of assurance might be established in terms such as acceptable probabilities or consequences. conservative assumptions, or sufficient margin.

Summary and reference to enclosure 1 Address ... the compliance exception, and related guidance with focus on ... "failed to meet known and established standards of the Commission because of omission or mistake of fact" ...

or ... "new or modified interpretations of what constitutes compliance" History of the PORV and Safety Valve Issues (requirements, agency actions, standards, guidance)

Summary and reference to enclosure 2 History of Westinghouse Nuclear Safety Advisory Letter and Related Activities Summary and reference to Enclosure 3 Discussion THE QUESTION A simple explanation of the effort:

The Backfit Appeal Panel will answer the question ... does this case fit. .. "failed to meet known and established standards of the Commission because of omission or mistake of fact" ... or ...

"new or modified interpretations of what constitutes compliance" Response to the EDO questions:

1. Were the approvals based on a mistake? If so, what was the mistake and what are the implications for Braidwood and Byron?

In the May 3, 2016 letter to Exelon, NRC claims that "The NRC erred in approving a sequence of events that allowed the inadvertent operation of the emergency core cooling system, chemical and volume control system malfunction, and inadvertent opening of a pressurizer safety or relief valve analyses in the 2001 and 2004 Safety Evaluations" and "the NRC staff understood the PSVs to be qualified for water relief when, in fact, they were not."

Exelon claims, in their December 8, 2015 backfit appeal letter, that "the compliance 4

exception requires mare than simply asserting that the prior staff approvals were wrong - the NRC must demonstrate that the prior approvals were erroneous because of an omission or mistake of fact at the tim 3 of the approval. The NRC has not made that case here."

2. What is the known and established standard for water qualification of pressurizer safety valves?
3. What is the known and established standard for progression of postulated events between categories Jf severity? Include a discussion of Regulatory Issue Summary 2005-29, "Anticipated Transients that Could Develop into More Serious l:vents," dated December 14, 2005, and the draft Revision 1 that was is~ ued for public comment in 2015.
4. Does the current licensing t asis for Braidwood and Byron comply with the applicable regulations? Is i1 adequate to provide protection to public health and safety?
5. Given that Exelon suggests that the NRC pursue a cost-justified substantial safety enhancer,ent backfit, what is the contribution to overall plant risk of the current con1iguratian at Braidwood and Byron?

Recommendation(s) of whether a backfit ie necessary at Braidwood and Byron, as well as whether the requirements of the compli,mce backfit exception of 10 CFR 50.109{a)(4)(i) have been met for :his issue.

Recommendation for any additional action.t ..

5

ENCLOSURE 1 SPENCER INPUT ON HISTORY OF BACKFIT RULE. Rev. O. 2016-07-16

{In preparing this input, I read the original 1970 back/it rule, the 1983 ANPR and interim policy statement, the 1984 proposed revision to the Back/it Rule, the 1985 final revision to the Back/it Rule, the 1987 court decision invalidating the 1985 rulemaking, the 1987 proposed rule responding to the court decision, the 1988 final rule responding to the court decision, and the 1989 court decision upholding the Back/it Rule as revised. I also read the SECY papers and SRMs for the 1985 final rule, 1987 proposed rule, and 1988 final rule. In all of this, I did not find much on the compliance exception that was useful.}

IV. History of the Backfit Rule and the Compliance Exception The Backfit Rule, 10 CFR 50.109, was originally promulgated in 1970 (35 FR 5317). Because of perceived deficiencies in the rule, the NRC substantially revised it in 1985 (50 FR 38097). The 1985 rule was challenged in court, and the U.S. Circuit Court for the District of Columbia (D.C. Circuit) vacated this rule in its entirety. The D.C. Circuit took this action because it concluded that the revised rule could be interpreted to allow the NRC to consider costs in defining or redefining what is required for adequate protection of the public health and safety. Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 824 F.2d 108, 119-20 (1987). In response the NRC revised the Backfit Rule in 1988 (53 FR 20603) to remove any implication that costs could be considered in defining or redefining adequate protection.

The 1988 revisions only differed from the 1985 rule to the extent necessary to address the court's concerns. The 1988 rule was also challenged in court, but this time the D.C. Circuit upheld the rule.

Union of Concerned Scientists v. U.S. Nuclear Regulatory Com'n, 880 F.2d 552 (1989).

In its current form, 10 CFR 50.109(a){l) defines backfitting as "the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position ...." Unless one of three specified exceptions apply, the NRC may impose a backfit only if it performs a backfit analysis in accordance with 10 CFR 50.109(a)(2) and determines in accordance with 10 CFR 50.109(a)(3) "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection."

Section S0.109(a)(4) sets forth the three exceptions to the requirements of 10 CFR 50.109(a)(2l and (a)(3). The first exception, the compliance exception, applies if the "modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee." 10 CFR 50.109(a)(4)(i]. The second and third exceptions 6

relate to actions ensuring adequate protection or to actions that involve defining or redefining adequate protection.

The Commission addressed its intended application of the compliance exception in the 1985 rulemaking (50 FR at 38103):

The compliance exception is intended to address situations in which the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact. It should be noted that new or modified interpretations of what constitutes compliance would not fall within the exception and would require a backfit analysis and application of the standard.

ln the 1985 rule, the Commission acknowledged that staff interpretations of regulations are not legally binding, but the Commission also stated that "staff interpretations of broadly stated rules are often necessary to give a rule effect and in some instances may be a causal fact?r in initiating a backfit." Id. at 38102. The Commission also stated, "Many of the most important changes in plant design, construction, operation, organization, and training have been put in place at a level of detail that is expressed in staff guidance documents which interpret the intent of broad, generally worked [sic]

regulations." Id. at 38103. 3

{In addition to the above, I found the following potentially useful statements from NRC guidance documents on backfitting. I am not sure how much of this will be relevant or where it should be included if relevant, but I am providing it for your information and for potential future use in the document.}

NUREG-1409. Backfitting Guidelines, 1990 To be a backfit, "a new or revised staff position or requirement must be involved, that is, there must be a change in content or applicability of the previously applicable regulatory staff position (in the direction of increased safety requirements} .... "

At 3:

Applicable Regulatory Staff Position A requirement or position already specifically imposed on or committed to by a licensee is called an applicable regulatory staff position. There are several different types of positions, such as

- legal requirements, as in explicit regulations, orders, and plant licenses and in amendments, conditions, and technical specifications

~ The 1988 rulemaking neither revised the compliance exception as stated in the 1985 rule nor provided additional guidance on its interpretation.

7

- written licensee commitments such as those contained in the final safety analysis report, licensee event reports, and docketed correspondence, including responses to NRC bulletins, generic letters, inspection reports, or notices of violation and confirmatory action letters

- NRC staff positions that are documented explicit interpretations of more general regulations and are contained in documents such as the Standard Review Plan, branch technical positions, regulatory guides, generic letters, and bulletins For the purpose of this report, a change in the applicable regulatory staff position will be subsequently referred to as a new or revised position.

(4) How does the bock/it rule apply to new staff positions that reflect an evolving understanding af technical issues?

New or revised staff positions are backfits when they are imposed on licensees and result in a change in structures, systems, design, or procedures (as described in 10 CFR 50.109). A backfit analysis is required whenever new or revised positions are imposed to achieve cost-justified substantial safety enhancements. A backfit analysis is not required if the new or changed position is imposed to bring a facility into compliance or if it is necessary to provide assurance of adequate protection. In those cases, however, a written evaluation is needed to provide the objectives of and reasons for the modification and the basis for invoking the exception.

An evolving understanding of issues does not, by itself, define which category fits a particular backfit.

Judgment must be applied to the facts of each particular case to determine whether the backfit is for compliance, to provide adequate protection, to redefine adequate protection, or to achieve a cost-justified substantial safety enhancement. For example, with regard to compliance, the 1985 statement of considerations for 10 CFR 50.109 indicates that 11 the compliance exception is intended to address situations where the licensee has failed to meet known and established standards of the Commission because of omission or mistake of fact .... new or modified interpretations of what constitutes compliance would not fall within the exception .... "

(7) Is it appropriate for the NRC staff to rely on informal or formal communications to other licensees as official NRC positions? What about NRC tacit approval of documents?

Informal or formal communications to one licensee are not official positions to all licensees. Section 053 of Manual Chapter 0514 identifies what can be applied as official staff positions in a plant-specific context. They are legal requirements such as contained in explicit regulations, orders, and plant licenses; written commitments such as contained in final safety analysis reports, licenses event reports, and docketed correspondence; and documented, approved explicit interpretations such as contained in the SRP, branch technical positions, regulatory guides, generic letters, and bulletins. Orders, licenses, and written commitments are applicable only to a particular licensee.

8

If the staff previously exempted a licensee from a legal requirement or approved position, it is not applicable to that licensee for the purpose of backfit consideration. Explicit exemption would be done formally in writing. The Appendix to NRC Manual Chapter 0514 discusses tacit approval under reanalysis of issues. Two situations are covered. In the first case, staff review of a previously accepted licensee action or program may result in a requested change. This would be classified as a backfit because it represents a change in a previous staff position and would require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed in the backfit rule). In the second case, a licensee submittal committing to a specific course of action that has not received timely NRC staff review is implemented by the licensee. In this case, it is considered that the NRC staff tacitly accepted the licensee's action since timely notice to the contrary was not given. If the NRC staff subsequently adopts a different position and requests a change in the licensee action, this change may be classified as a backfit and thus require a backfit analysis (or a documented evaluation if it meets one of the exceptions listed in the backfit rule).

[Industry claim]: Bulletin 88-11 is completely Jacking any 10 CFR 50.109 justification.

[NRC response]: Although the justification was not printed in the bulletin, NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," was justified as a backfit. It is an example of a backfit that was determined by the responsible NRC official to be required as a matter of compliance with existing requirements and commitments. The CRGR reviewed the bulletin and concurred. The regulations currently require licensees to meet the applicable codes of the American Society of Mechanical Engineers (ASM E), Boiler and Pressure Vessel Code. Because of the staff's concern with the integrity of the surge line, licensees were requested to perform their fatigue analysis in accordance with the latest ASME Section Ill requirements that incorporate high cycle fatigue analysis. The justification provided by the staff was that previously unconsidered thermal stratification phenomenon may invalidate the existing analysis performed to confirm the integrity of the surge line.

Subsequently, it was understood that some licensees believed that the staffs rationale was in error because they were not committed to the latest ASME Section Ill requirements by virtue of their license commitment. However, the issue became moot because these licensees undertook the analysis voluntarily in view of the safety importance of the issue and the fact that previous versions of the ASME Code did not completely address the concern.

At 1S:

(1) If an inspector has previously accepted {ie, provided tacit approval of) a licensee's method, does a specific request for change constitute a back/lit and if so, is a back/it analysis required?

Cases where an inspector provides tacit approval are relatively rare. Simply not challenging a licensee's practice normally would not be considered tacit approval. The only example provided in Manual Chapter 0514 is a case where the NRC has indicated tacit approval by not acting in a reasonable time on a licensee submittal and the licensee has moved ahead to implement the proposal described in the submittal. For the purpose of this question, it would most likely arise in connection with review of a licensee response to an inspection report.

9

Explicit approval could be provided in an inspection report that states that a particular approach is acceptable. However, conclusions of that nature are usually made in safety evaluation reports rather than inspection reports.

NRC Manual Chapter 0514 (1988) {This manual chapter is included as Appendix D to NUREG-14D9. The manual chapter was referenced in the 1988 rulemaking, and a working draft was provided to the Commission in SECY-88-102 for information. The only particularly interesting nugget is a definition of "applicable regulatory staff positions" that is slightly more detailed than the definition in NUREG-14D9. I have shown the additional detail with balded redline text.}

Section 053 053 Applicable Regulatory Staff Positions. Applicable regulatory staff positions are those already specifically imposed upon or committed to by a licensee at the time of the identification of a plant*

specific backfit, and are of several different types and sources:

a. Legal requirements such as in explicit regulations, orders, plant licenses (amendments, conditions, technical specifications). Note that some regulations have update features built in, as for example, 10 CFR 50.SSa, Codes and Standards. Such update requirements are applicable as described in the regulation.
b. Written commitments such as contained in the FSAR, LERs, and docketed correspondence, including responses to Bulletins, responses to Generic Letters, Confirmatory Action Letters, responses to Inspection Reports, or responses to Notices of Violation.
c. NRC staff positions4 that are documented, approved, explicit interpretations of the more general regulations, and are contained in documents such as the SRP, Branch Technical Positions, Regulatory Guides, Generic Letters, and Bulletins; and to which a licensee or an applicant has previously committed to or relied upon. Positions contained in these documents are not considered applicable staff positions to the extent that staff has, in a previous licensing or inspection action, tacitly or ex.plicitly excepted the Ileen see from part or all of the position. 5 FN 4 Requirements may be imposed by rule or order. Staff interpretations such as examples of acceptable ways to meet requirements are not requirements in and of themselves.

FN 5 Imposition of a staff position from which a licensee has previously been excepted is a backfit.

10

ENCLOSURE 2 History of Pressure Relief Valves in Nuclear Power Plants Nuclear power plants in the United States use various types of pressure relief valves to protect personnel and equipment from overpressure events within reactor fluid systems. Pressure relief valves include safety valves, safety relief valves, and relief valves with different designs, operating conditions, and requirements. The ASME BPV Code specifies requirements for the design and testing of pressure relief valves used for various functions in nuclear power plants.

For example, the current ASME BPV Code requirements in Subsection NB-7000 for safety valves specify steam applications and for relief valves specify steam or liquid applications. The ASME OM Code provides requirements for the preservice and inservice testing of pressure relief valves in nuclear power plants.

Braidwood UFSAR (Revision 9, December 2002), Section 5.4.13 describes the pressurizer safety and relief valves. The three pressurizer safety valves (PSVs) are Crosby Model HP-BP-86, size 6M6 (6-inch), spring loaded pop type opened by direct fluid pressure with 2485 psig (another document indicates 2460 psig) set pressure. The 6-inch pipe connecting the PSVs to the pressurizer fom, a loop seal. Each reactor has two power-operated relief valves (PORVs).

The PORVs are Copes-Vulcan Model D-100-160 3-inch pneumatic-actuated globe valves that respond to a signal from the pressure sensing system or to manual control. One PORV is set to open at 2335 psig and the other at 2345 psig. Each PORV can be isolated by a motor-operated block valve. Both the PSVs and PORVs are indicated in Table 5.4-13 to have specific relieving capacities with saturated steam.

The Braidwood UFSAR (Revision 7, December 1998) Table 5.2-1 references the 1971 Edition through the Winter 1972 addenda of the ASME BPV Code, Section Ill, for the PSVs.

What "qualification" requirements are specified in the design and procurement specifications for the pressurizer relief valves used in the pressurizer at Byron and Braidwood?

The 1ST Program at Braidwood Units 1 and 2 is specified to satisfy the ASME OM Code 2001 Edition through the 0Mb 2003 Addenda until July 2018. The 1ST Program at Byron Units 1 and 2 has updated to the ASME OM Code 2004 Edition through the 2006 Addenda.

The Braidwood 1ST Program specifies:

PORV - fail safe test, and exercise open and closed, every refueling outage; and position indication test every 2 years.

PORV Block Valve- exercise open and closed every 3 months; and position indication test every 2 years.

PSV - Relief Valve Test every 5 years; and position indication test every 2 years. The 1ST Program references Appendix I to the OM Code for relief valve testing.

Are relief valve tests performed for the PORVs?

11

The 1979 accident at Unit 2 at the Three Mile Island nuclear power plant included failure of a power-operated relief valve in the pressurizer to perform its intended function. As a result, the NRG issued recommendations regarding performance testing of safety and relief valves used in nuclear power plants in NUREG-0578. In particular, the NRC staff recommended in Section 2.1.2 in NUREG-0578 that nuclear power plant licensees commit to provide performance verification by full-scale prototypical testing for all relief and safety valves.

The NRC subsequently issued a letter to all then-operating nuclear power plants and applicants for operating licenses and holders of construction permits foiwarding NUREG-0737 that included numerous requirements in response to the TMI Unit 2 accident. In particular, Requirement 11.0.1, "Performance Testing of Boiling-Water Reactor and Pressurized-Water Reactor Relief And Safety Valves (NUREG-0578, Section 2.1.2)," specified the NRC position that pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to "qualify" the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents. The detailed clarification in NUREG-0737 of this NRC position specified the following:

Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test-pressures shall be the highest predicted by conventional safety analysis procedures.

Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping, and supports, as well as the valves themselves.

A Performance Testing of Relief and Safety Valves--The following information must be provided in report form by October 1, 1981:

(1) Evidence supported by test of safety and relief valve functionability for expected operating and accident (non-ATWS) conditions must be provided ta NRC. The testing should demonstrate that the valves will open and reclose under the expected flow conditions.

(2) Since it is not planned to test all valves on all plants, each licensee must submit to NRG a correlation or other evidence to substantiate that the valves tested in the EPRI (Electric Power Research Institute) or other generic test program demonstrate the functionability of as-installed primary relief and safety valves. This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the final safety analysis report (FSAR). The effect of as-built relief and safety valve discharge piping on valve operability must also be accounted for, if it is different from the generic test loop piping.

(3) Test data including criteria for success and failure of valves tested must be provided for NRC staff review and evaluation. These test data should include data that would permit plant-specific evaluation of discharge piping and supports that are not directly tested.

In describing the type of review to be conducted far this regulatory position, the NRC staff stated the following:

12

Pre-implementation review will be performed for EPRI and BWR test programs with respect to qualification of relief and safety valves. Also, the applicants' proposal for functional testing or qualification of PWR valves will be reviewed. Post-implementation review will also be performed of the test data and test results as applied to plant-specific situations.

In specifying the documentation required to satisfy this regulatory position, the NRC staff stated the following:

Pre-implementation review will be based on EPRI, BWR, and applicant submittals with regard to the various test programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following valve qualification dates can be met Final PWR (EPRI) Test Program--July 1, 1980 Final BWR Test ProgramHOctober 1, 1980 Block Valve Qualification Program--January 1, 1981 Post-implementation review will be based on the applicants' plant-specific submittals for qualification of safety relief valves and block valves. To properly evaluate these plant-specific applications, the test data and results of the various programs will also be required by the following dates:

PWR (EPRl)/BWR Generic Test Program Results--July 1, 1981 Plant-specific submittals confinming adequacy of safety and relief valves based on licensee/applicant preliminary review of generic test program results--July 1, 1981 Plant-specific reports for safety and-relief valve qualification--October 1, 1981 Plant-specific submittals for piping and support evaluations--January 1, 1982 Plant-specific submittals for block valve qualification--July 1. 1982.

In January 1983, EPRI issued NP-2770-LD, "EPRI/C-E PWR Safety Valve Test Report," that described the testing of PWR primary system safety and relief valves in response to NUREg-0578, Section 2.1.2, and NUREG-0737, Section 11.0.1.A, requirements. Volume 1 is publicly available and provides a summary of the test program and its results. Volumes 5, 6, and 7 provide the detailed results for testing of Crosby pressure relief valves. In addition, ComEd letter dated November 27, 2000, for the Byron/Braidwood power uprate references EPRI Report NP-2628-SR, "EPRI PWR Safety and Relief Valve Test Program - Safety and Relief Valve Test Report,~ December 1982.

Can we obtain copies of the detailed EPRI test reports for the Braidwood and Byron pressurizer pressure relief valve types? Volume 6 of EPRI NP-2770-LP is not available in the NRC Technical Library and the NRC contacts with EPRI do not believe that EPRI would provide the report without cost. Request submitted to the ADAMS support staff to search the ADAMS Legacy library for the EPRI NP-2770-LD (Volume 6).

13

Volume 1 of EPRI NP-2770-LP summarizes the EPRI testing of various PSVs for steam, transition, and water conditions. Section 4.5 indicates that for the Crosby 6M6 the following tests were performed: 11 steam tests with filled loop seals, 3 steam-to-water transition tests, and 2 water tests. The valve experienced chatter during the tests, and one water test had to be terminated. The report does not discuss PORV testing.

The ComEd letter dated November 27, 2000, for the Byron/Braidwood power uprate states that the NRC reviewed and approved EPRI NP-2628-SR ComEd references a letter from L. Olshan (NRC) to H. Bliss, dated August 18, 1988, "NUREG-0737, Item 11.D.1, Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and 2;" and a letter from S. Sands to T.

Kovach, dated May 21, 1990, "NUREG-0737, Item 11.D.1, "Performance Testing on Relief and Safety Valves for Braidwood Station, Units 1 and 2."

In the August 18, 1988, letter from L Olshan to H. Bliss, the cover letter indicates that the attached INEL Technical Evaluation Report (TER) EGG-NTA-8028 (January 1988) provides the review of the Byron response to NUREG-0737, Item 11.D.1. The staff letter indicates that the licensee should develop and adopt plant procedures to inspect the pressurizer valves after each lift involving loop seal or water discharge. The TER reviewed the EPRI testing of a PSV and PORV similar to the Byron valves. The TER indicates that the PSV had two applicable tests: a loop seal/steam water transition test where the valve opened, chattered and stabilized to close; and a saturated water test where the valve opened with water, chattered, and stabilized. The TER indicates that the PORV opened and closed on demand in the loop seal/steam water transition test with a bending moment that was evaluated by analysis. The TER concluded that Byron provided an acceptable response to NUREG-0737, Item 11.D. 1. The May 21, 1990, letter with the Braidwood TER had the same findings.

EPRI NP-2628-SR (dated December 1982) describes the PSV and PORV testing with safety valve tests at CE in Windsor, CT, and relief valve tests at Marshall Steam Station, NC, and Wyle, Norco, CA The report indicates that the Crosby 6M6 is used as PSVs in 38 plants including Byron and Braidwood. The report indicates that two water tests were conducted with galling of guide surfaces and damage to internal parts. The report indicates that the Copes-Vulcan Relief Valve (316 w/Stellite Plug and 17-4PH) used as PO RVs at Byron and Braidwood was tested in 11 Marshall tests and 9 Wyle tests. The report indicates that the PORV opened without damage.

Appendix E, "Requirements Resulting from TMl-2 Accident," in Braidwood UFSAR in Section E.23, "Relief and Safety Valve Test Requirements (II. D.1)," states that a letter dated April 1, 1982, for D. Hoffman (Consumers Power) transmitted the Safety and Relief Valve Test Report for the EPRI PWR Safety and Relief Valve Test Program. The UFSAR also states that ComEd submitted the plant-specific final evaluation confirming the adequacy of the relief and safety valves by letter from T. Tramm, dated October 26, 1982. Braidwood NUREG-1002 SER Supplement 1 {September 1986) states in Section 3.9.3.3, "Design and Installation of Pressure Relief Devices," that EPRI had completed a full-scale valve testing program and that the Owners Group submitted the test results in WCAP-10105 in a letter dated July 27, 1982, from

0. Kinglsey to S. Chilk. The SER states that the applicant responded to a requirement to submit a report which would demonstrate operability of these valves with submittals dated July 1 and October 26, 1982, and December 30, 1983. Byron NUREG-0876 SER Supplement 5 14

(October 1984) references these July 1 and October 26, 1982, and December 30, 1983, letters from the applicant. Request submitted to ADAMS support staff for these Legacy documents.

The July 27, 1982, letter from 0. Kingsley provides the Westinghouse Owners Group review of the EPRI test data in response to NUREG-0737, Item 11.D.1. The letter forwards WCAP-10105, "Review of Pressurizer Safety Valve Performance as observed in the EPRI Safety and Relief Valve Test Program" {June 1982). The report indicates that the design specifications for PSVs in Westinghouse designed nuclear power plants is for steam service only. The conclusions indicate valve chatter (no discussion of valve damage).

In Braidwood NUREG-1002 SER Supplement 1 (September 1986) in Section 3.9.3.3, the NRC staff stated that on the basis of preliminary review, the staff had concluded that the applicant's general approach to responding to this item is acceptable, and provides adequate assurance that the RCS overpressure protection systems at Braidwood can adequately perform their intended functions. The staff stated that if the detailed review reveals modifications or adjustments to safety valves, PORVs, PORV block valves, or associated piping are needed to ensure that all intended design margins are present, the staff will require that the applicant make appropriate modifications. The staff stated that this was a Confirmatory Item.

In Byron NUREG-0876 SER Supplement 5 (October 1984) in Section 3.9.3.3, the NRC staff provided a similar discussion of the status of the NRG review of the capability of the Byron pressurizer valves.

In Byron NUREG-0876 SER Supplement 8 (March 1987), the staff stated TMI Item 11.D.I (3.9.3.3) had been closed in Supplement 5 to the Byron SER. (A similar closure was not found in the Braidwood SER supplements).

The Braidwood and Byron operating license reviews included direct evaluation of the TMI action items as discussed in Braidwood NUREG-1002 SER, Section 1.1, "Introduction." The staff stated that the review and evaluation of compliance by the applicant to the licensing requirements established in NUREG-0660 and NUREG-0737 (including item 11.D.1 in Table 1.1) were incorporated into the reviews summarized throughout the SER.

15

ENCLOSURE 3 History of Westinghouse Nuclear Safety Advisory Letter 93-013 and Related Activities SRP (Revision 1, dated July 1981) Chapter 15.5.1-15.5.2, "Inadvertent Operation of ECCS and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory,"

and Chapter 15.6.1, "Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve," provided general staff guidance for these plant transients.

Revision 2 (dated March 2007) to these SRP chapters provides significantly more detail, including PSVs and PORVs are assumed to fail open if they relieve water without being qualified.

In 1993 and 1994, Westinghouse issued a Nuclear Safety Advisory Letter NSAL-93-013 (June 30, 1993) and NSAL-93-013, Supplement 1 (October 28, 1994) under 10 CFR Part 21 to operating nuclear power plants (including Braidwood and Byron) in response to its discovery that potentially nonconservative assumptions were used in the licensing analysis of the lnadverent Operation of the ECCS at Power accident In this NSAL, Westinghouse recommended that licensees determine if their Pressurizer Safety Relief Valves are capable of closing following discharge of subcooled water. Westinghouse noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. Westinghouse indicated that water relief through the Power Operated Relief Valves (PORVs) is not a concern, because the PORV block valves can be used to isolate the PORVs if they fail to close. If the PS RVs are not designed or qualified for subcooled water relief, Westinghouse recommended that licensees re*

evaluate the Inadvertent ECCS Actuation at Power accident used three possible options of reducing ECCS flow used in the safety analysis, using a less restrictive operator response time, or crediting the use of one or more PORVs to help mitigate the accident.

Did the NRC staff take a position on Westinghouse NSAL-93-013 during the industry response to this Part 21 notice?

What was the response from the Braidwood/Byron licensees to this Westinghouse NSAL?

What was the NRC staff evaluation of the Braidwood/Byron response?

Did Exelon demonstrate the capability of the PORV block valves for Braidwood and Byron as part of its GL 89-1 O program with continued periodic verification as part of its GL 95-05 program?

On May 29, 1998, ComEd proposed an amendment to the Braidwood/Byron TS to take credit for the automatic operation of the PORV to provide mitigation for the IOECCS. In the amendment request, ComEd stated that the PSVs have not been qualified to reseat after passing subcooled liquid. ComEd stated that the PORVs at Braidwood/Byron are safety-related components with safety*related actuators and accumulator tanks. Com Ed stated that the PORV control circuits are classified as safety.related. Com Ed noted that some portions of the circuitry are nonsafety-related with improvements implemented in response to GL 90-06. Com Ed stated that the PORV block valves are within the scope of the GL 89-10 program. In a letter dated May 13, 1999, the NRG staff provided a request for additional information regarding the reliance on the PORVs. In its letter, the staff documented the basis for its concerns that the Byron design 16

related to the PORV circuitry did not meet the single failure criterion. In response to these concerns, ComEd withdrew its TS amendment request in a letter dated July 16, 1999.

Did ComEd (or Exelon) propose a later TS amendment?

In 2001, the NRC approved a stretch power uprate for Braidwood and Byron. In RAls, the staff requested that the licensee address water solid conditions in the pressurizer. The licensee responded that the EPRI testing indicated that the valve could close although with some chatter and damage. These references include NRC letters dated October 19 and November 21, 2000 (Dick to Kingsley), and ComEd letters dated November 27, 2000, and January 31, 2001 (Krich).

Stretch power update license amendment dated May 4, 2001.

In its letter dated November 27, 2000, ComEd responded to RAls on the license amendment request (LAR) to permit uprated power operations at Byron and Braidwood. NRC staff question G.9 indicated that the NRC staff had generally not accepted a solid pressurizer for IOECCS to order to avoid the potential for all three PSVs to be stuck open due to liquid relief through these safety valves. The licensee stated that Section 15.5.1 of the UFSAR had been revised to credit the pressurizer safety valves to pass water. The licensee discussed the EPRI testing program in response to NUREG-0737 with the results summarized in EPRI NP-2628-SR. The licensee references NRC letter from L Olshan to H. Bliss, dated August 18, 1988, and S. Sands to T.

Kovach, dated May 21, 1990, transmitting Technical Evaluation Reports with the results of the NRC's review of the Byron and Braidwood response to NUREG-0737, Item 11.D. 1, respectively.

In its letter dated January 31, 2001, Exelon provided a response to a supplement to Question G. 9 requesting the temperature of water to be passed by the pressurizer safeties and the length of time that the safeties are expected to pass water. The supplement also asked the licensee to discuss what EPRI tests are applicable to the Byron and Braidwood condition. In response, Exelon stated that the PSVs would close after passing water, although they may not be leaktight. Exelon stated that the leakage from up to three leaking PSVs is bounded by one fully open PSV. Exelon indicated that the EPRI testing of the Crosby safety valves in EPRI NP-2770-LD, Volumes 1 and 6, are applicable. Exelon indicated that valve chatter occurred during the tests with damage to the internals, but that the safety valve closed in response to system depressurization. Exelon stated that the Byron/Braidwood pressurizer water temperature of 590F is higher than the EPRI tests (530F). Exelon stated that the assumed length of the event is 20 minutes from initial SI signal to system pressure restored below PSV lift setpoint. Exelon notes that all PSVs set to lift at same pressure.

In the NRC SER granting the Byron/Braidwood power uprate (May 4, 2001) in Section 3.2, "Non-LOCA Transient Analysis," the staff discussed its review of the performance of the PORVs and PSVs to discharge liquid water for approximately 20 minutes. The staff discussed the EPRI testing program with the conclusion that the safety valve closed in response to system depressurization. The staff stated that it had reviewed the licensee's evaluation of the performance of the PSVs for liquid water conditions. The staff stated that it found that the EPRI tests adequately demonstrate the performance of the valves for the expected water temperature conditions and that there is reasonable assurance that the valves will adequately reseat following the spurious SI event. The staff stated that a review of the EPRI test data indicates that the PSVs may chatter for the expected fluid inlet temperature but that the resulting PSV seat leakage following the liquid discharge would be less than the discharge from one stuck-open PSV. Therefore, the staff found the licensee's crediting of the PS Vs to discharge liquid water during the spurious SI event to be acceptable.

17

Where is the capability of the PORVs to perform with liquid water addressed?

Can PSVs be set safely at slightly different setpoints?

Byron/Braidwood UFSAR (Revision 9, dated December 2002) in Chapter 15.5.1 includes PSV water relief, and references INEL 1988 report and L. Olshan August 1988 SER. UFSAR Revision 15 (dated December 2014) includes IOECCS does not progress into a stuck open PSV LOCA event. UFSAR (Revision 15) states that all three PSVs may lift but will reclose, and that the leakage is bounded by one fully open valve with the consequences bounded by the IOPSRV event. UFSAR (Revision 15) references 2001 Stretch Power Uprate SER and 2004 PSV Setpoint SER.

In 2004, the NRC approved a PSV setpoint amendment for Braidwood and Byron. In an RAI, the staff requested that Exelon perform a quantitative analysis regarding PSV water cycles and relief/discharge water temperature. These references include NRC e-mail on October 2, 2003 (Chawla to Bauer), and EGG letter dated January 29, 2004 (Ainger).

On August 26, 2004, NRC issued license amendment to adjust PSV setpoints for Braidwood and Byron. The change represents a change in lift tolerance from +/-1 percent around a lift setting of 2485 psig to +/-:2 percent around a lift setting of 2460 psig. As for the LOAC with RCP seal injection event, the licensee's analysis indicated that continued injection of water into the RCS through the RCP seals would result in a water-solid pressurizer and water discharge through the PSVs. The proposed PSV setpoint tolerance assuming negative tolerance would result in a lower PSV lift setpoint. With the lower setpoint, the PSV would open earlier and a larger number of PSV water cycles and a lower water discharge temperature could result during the transient. The licensee performed an analysis of the LOAC with RCP seal injection event and determined the revised PSV setpoint would result in an increase of about one PSV water cycle and a reduction in the liquid discharge temperature of about 0.5F. A comparison of the reanalysis showed that the spurious SI event remained the limiting event since it resulted in a greater increase in the number of PSV water cycles (two cycles vs. one cycle) and a greater decrease in the PSV discharge water temperature (3.0 F vs. 0.5 F) than that calculated for the LOAC with RCP seal injection event The water discharge temperature in the analysis of record (AOR) for the spurious SI event was 590 F. The lowest discharge water temperature for the spurious SI event with the revised PSV setpoint is 587 F (i.e .. 590 F - 3.0 F). The staff found that the calculated water discharge temperature (587 F) is significantly higher than the discharge water temperature of 530 F that was used to support operability of the PSVs as discussed in the AOR. The staff concluded that the reanalysis is acceptable to assure that the PSVs will remain operable following a spurious SI event. This SER references the May 2001 Stretch Power Update SER.

On February 7, 2014, NRC issued Measurement Uncertainty Recapture (MUR) power uprate with decision that IOECCS was out of scope because the licensee did not modify the Chapter 15 analyses related to PSVand PORV water relief.

In the October 2015 Backflt SE, the staff references Millstone (1998) and Callaway (2000) requests (ML011800207 and ML003719636) for upgrading PORVs for water relief; Beaver Valley (2004) EPU requests (ML?) for qualifying PORVs for water relief; and Turkey Point and St. Lucie Unit 2 EPU requests (ML11293A359 and ML12235A463) in support of its position.

18

Salem TS Amendment - June 4, 1997 On June 4, 1997, NRC granted TS change to ensure that the automatic capability of the power operated relief valves to relieve pressure is maintained. In response to NSAL 93-013, the licensee determined that an inadvertent Safety Injection (ISi) actuation at power could cause the pressurizer to become water solid and PSVs lifting with water relief if the automatic operation of the PORVs is not made available for reactor coolant system depressurization early in the transient. The Salem pressurizer safety valves are not designed to relieve water. Thus, the water relief has the potential to cause the PSVs to fail in the open position. In the course of the review of the licensee's January 31, 1997 application, the NRC staff noted that the pressurizer PORVs were not designed to "safety related" standards and thus could not be credited for mitigation of the inadvertent SI actuation at power incident when the PORV is operating in the automatic mode. In response to this observation, the licensee proposed an upgrade of PORVs as described in the March 14 and April 8, 1997 supplements, to eliminate the possibility that a single active failure of a PORV component could prevent the mitigation of the inadvertent SI actuation at power incident. Endurance tests performed with five different trims (with different trim materials) on one PORV at Wyle Laboratories demonstrated that: 1) after 2000 consecutive operations, there were no packing leaks nor packing gland adjustments required; 2) there was no diaphragm failure; and 3) the solenoid valve withstood 10,000 operations without any loss of function.

Millstone Unit 3 License Amendment - June 5, 1998 On June 5, 1998, the NRG granted a license amendment for Millstone Unit 3 for changes to TS to ensure that the automatic capability of PORVs to relieve pressure is maintained. The revised TS Bases stated that the PORVs and their associated piping have been demonstrated to be "qualified for water relief. The PORVs prevent water relief from the PSVs for which qualification for water relief has not been demonstrated. The licensee proposed to qualify the PORVs and associated piping for water relief. To provide added assurance that the PSVs will not be damaged due to water relief during an inadvertent safety injection (ISi) event, the licensee upgraded the PORV circuitry, added additional PORV surveillance requirements, qualified the PORVs and associated piping for water relief, and made EOP changes to allow plant operators additional time to terminate the event. The Millstone Unit 3 PORVs are Garrett pilot-operated, cage-guided globe valves. The licensee references EPRI NP-2670-LD, Volume 11, that was said to have been performed to generically resolve post TMl-2 issues associated with PORVs and safety valve qualification for water and steam relief, that documents the results from four tests of the Garrett PORV for water relief. The licensee stated that the PORVs and associated piping are qualified for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of water relief for an inadvertent ECCS actuation at power operation. The licensee also stated that the PORV manufacturer performed numerous cycle tests to verify the performance of the valve design. The PORV manufacturer also tested the valve seat leakage and found it to be acceptable. The staff found that the manufacturer's tests and the hot functional tests, taken together with the tests performed by EPRI, provide adequate assurance that the plant PORVs will perform adequately for the ISi event. The licensee stated that the PORV block valves are MOVs that have been evaluated for water relief in accordance with the GL 89-10 program. The NRC staff reviewed the licensee's submittals regarding the qualification of the PORVs for water relief during the ISi event. The staff found them acceptable based on the testing to provide adequate assurance that the plant PORVs will perform adequately during an ISi event.

Callaway TS License Amendment SER - September 25, 2000 19

In a letter dated September 25, 2000, the NRC issued a license amendment to revise the Callaway TS, induding TS 3.4.10, "Pressurizer Safety Valves," to change the PSV lift setting range. To prevent water passing through the PSVs in an IOECCS, the licensee will modify and upgrade the PORV circuitry to full Class 1E to take credit for automatic action of at least one PORV during the event. For the inadvertent ECCS actuation at power event, the safety analysis credits operator actions from the main control room to terminate flow from the normal charging pump (NCP) and to open a PORV block valve (assumed to initially be closed) and assure the availability of at least one PORV for automatic pressure relief. Analysis results indicate that water relief through the PSVs, which could result in the Condition II event degrading into a Condition Ill event if the safety valves did not reseat, is precluded if operator actions are taken within the times assumed in the analysis to terminate NCP flow and to assure at least one PORV is available for automatic pressure relief. Reanalysis of the Inadvertent ECCS Actuation at Power event was performed by Westinghouse in support of this amendment application. The reanalysis credits operator actions from the main control room to open a pressurizer PORV block valve (assumed to initially be closed) and assure the PORV handswitches are in the automatic operation position to allow automatic actuation of at least one PORVon demand. This would prevent water relief through the PSVs. In its request, the licensee stated that the design function of the valves was not being changed and the conclusions documented in the NRC Safety Evaluation of Callaway's response to NUREG-0737 Item 11.D.1 (dated September 10, 1987) are unchanged (see also FSAR Section 18.2.5). The licensee stated that the PORVs and associated discharge piping can accommodate water relief.

St. Lucie Unit 2 EPU SER September 24, 2012 This amendment increases the authorized maximum steady-state reactor core power level from 2700 to 3020 MWt. The licensee confirmed that the peak pressurizer water volume would not be sufficient to fill the pressurizer, assuring that this event would not develop into a more serious even1, by causing a PORV to stick open. after it has relieved water. No discussion of NSAL.

During power operations, the high pressure safety injection pumps are incapable of delivering flow to the RCS because the pumps' shut-off head is less than the normal RCS operating pressure of 2250 psia. Therefore, the inadvertent operation of the ECCS at power event is not a credible event and is not analyzed by the licensee for EPU. The licensee's position of not analyzing the event and the associated bases are consistent with that discussed in FSAR Section 15.5.1 and, therefore, are acceptable.

The eves malfunction that increases RCS inventory is an AOO that is evaluated for the effects of adding water inventory to the RCS. The NRC staff reviewed the licensee's analyses of the eves malfunction event and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The staff determined that the licensee's analysis demonstrated that the pressurizer did not become water solid, assuring no water was discharged through the PSVs.

The IOPORVevent, when viewed from the mass addition perspective, can be evaluated in two phases: (1) an inadvertent opening of a pressurizer relief valve, followed by (2) an inadvertent ECCS actuation. In the first phase, this event could be mitigated by closing the open pressurizer relief valve. If the valve could not be closed, then its block valve could be closed.

The first contingency action requires determining if a PORV is open or leaking and provides direction to place the PORV in override and close the PORV block valve. The NRC staff found that (1) the St. Lucie 2 multiple alarms were available for operator to detect occurrence of the IOPORV event, (2) plant procedures provided clear direction to the operator to close the open 20

PORV, and (3) the licensee's simulator exercise showed that closing the open PORV could be completed in 10 seconds, which was earlier than the SI actuation occurred at 40.9 seconds, and thus, would end the IOPORV transient with little or no SI delivery to the RCS. Based on its findings, the NRG staff determined that the pressurizer overfill analysis, available alarming system, and procedures in combination with simulator exercise result had provided reasonable assurance that the pressurizer would not be expected to fill to a water solid condition that could prevent the PORV or PSV from closing after they were open, and thus, supported that the event would not generate a more serious plant conditions, meeting the third AOO acceptance criterion. The NRC staff reviewed the licensee's analyses of the inadvertent opening of a pressurizer pressure relief valve event and concluded that the licensee's analyses adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The staff concluded that the licensee demonstrated that the all AOO acceptance criteria are satisfactorily met.

Turkey Point EPU SER June 15, 2012 The proposed amendment would increase the authorized maximum core power level of each Turkey Point (PTN) unit from the current licensed thermal power of 2300 to 2644 MWt. The CVCS malfunction that increases RCS inventory is evaluated for the effects of adding water inventory to the RCS. If the pressurizer fills and causes water to be relieved through the PO RVs or safety valves, then these valves could stick open and create an SBLOCA. This would violate the acceptance criterion that prohibits the escalation of an AOO into a more serious event.

Satisfaction of this acceptance criterion is demonstrated by showing that sufficient time exists for the operator to recognize the situation and end the charging flow before the pressurizer can fill. The licensee stated that the consequences of this event are bounded by the PTN small break loss of coolant accident. The staff did not accept this disposition. If action is not taken to secure the open valve, by either closing the PORV or its block valve, the event could escalate to a small break LOCA as the licensee stated, which is contrary to the non-escalation criterion set forth for AOOs. When the pressurizer become water solid, water begins to flow through the open PORV If the PORV is not qualified for water relief, then the staff states that it is likely the PORV will not close upon demand. In this way, the inadvertent opening of a PORV, an AOO, becomes a small break LOCA at the top of the pressurizer, a Condition Ill event. The staff requested that the licensee address the inadvertent opening of the PORV with respect to the third criterion. The licensee provided an analysis, performed largely in accordance with NRC-approved, Westinghouse analytic methodology using the RETRAN computer code; however, this analysis was performed assuming that the PORV opened instead of the safety valve.

Assuming the opening of the PORV is acceptable, because the safety valve is differently qualified, and reseats mechanically. An additional independent fault would be required to cause the safety valve to fail to close. The analysis indicated that the pressurizer would fill within about 240 seconds. The licensee also stated that there are multiple alarms to indicate the opening of a PORV. The licensee also stated that a prompt operator action is required to close the PORV, and if a response is not obtained, i.e., the PORV does not close, the operator is to close the block valve. Because the required actions are prompt, and because they are simple, the NRC staff agreed that the analyzed 240 seconds provides enough time to secure the inadvertently open PORV without filling the pressurizer. No discussion of NSAL.

ln December 2003, the NRC staff issued NRR Review Standard for Extended Power Uprates (RS-001, Rev. 0), which states in Item 8 on page 7 that pressurizer level should not be allowed to reach a pressurizer water-solid condition.

21

In RIS 2005-29, the NRC staff stated that typically Condition II event scenarios involve discharging water through relief or safety valves that are not qualified for water relief. The staff stated that these valves are then assumed to fail in the open position and create an SBLOeA.

The staff stated that it is concerned that some licensees may be crediting PORVs without qualification for water relief and without establishing additional restrictions to ensure the availability of PORVs and block valves. The staff states that NSAL allowing block valves to isolate PORVs is inconsistent with non-escalation criterion.

Is it appropriate to assume that relief or safety valves that are not qualified for water relief will fail in the open position rather than closing and leaking through a damaged seat?

EPRI Report 1011047 (August 2004), "Probability of Safety Valve Failure-to-Reseat Following Steam and Liquid Relief- Quantitative Expert Elicitation/ evaluated the potential increase in failure rates following steam and liquid relief through safety valves based on expert judgement.

The report found that the increase in failure rates is difficult to estimate because of limited data.

However, the experts considered that repeated water relief through safety valves might cause increased chatter, and therefore, an increased failure rate. In Appendix B, the EPRI report includes a summary of the EPRI PSV testing from another EPRI report which states:

The thermal hydraulic conditions expected during an SBO accident are such that most of the challenges ta a PSV would be from subcooled water. Because these valves are not designed for liquid flow, and because EPRI tests with subcooled liquid led to unstable conditions more often than not, the likelihood of PSV failure during an SBO accident would be quite high.

Did the staff evaluate this EPRI study?

In proposed Revision 1 to RIS 2005-29, the NRC staff addresses the specific Condition II scenarios of eves malfunction, IOECCS, and inadvertent opening of PORV or PSV.

Regarding the eves malfunction, the staff states that performing only the reactivity anomaly analysis or assuming that this malfunction is not as severe as the inadvertent ECCS is not acceptable. Regarding the IOECCS, the staff states that five of the alternative approaches in NSAL 93*013 fail to meet the non-escalation criterion. These are (1) closing block valve, (2) assuming that the PORV is not operable, (3) stuck-open PORVor PSV is addressed as a separate Condition II event, (4) a stuck-open PORV or PSV is not as severe as an SBLOCA, and (5) RCS loss through PORV is made up by ECCS flow. Regarding inadvertent opening of PORV or PSV, the staff states that inadvertent opening of PSV or PORV could continue as a Condition Ill SBLOCA and fails to meet the nan-escalation criterion.

Did the staff raise previously safety concerns with these alternative approaches in NSAL 93-0137 22

From: Scarbrough, Thomas Sent: Wednesday, July 27, 2016 1:16 PM To: Holahan, Gary; Clark, Theresa; Spencer, Michael; West, Steven

Subject:

RE: DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx Attachments: DRAFT Preliminary Findings July 27 2016 - tvc MAS - TGS.docx Here is my markup of the latest TVC/MAS version with some additional information.

Thanks.

Tom From: Holahan, Gary Sent: Wednesday, July 27, 201612:55 PM To: Clark, Theresa <Theresa.Clark@nrc.gov>; Spencer, Michael <M ichael.Spencer@nrc.gov>; West, Steven

<Steven.West@nrc.gov>; Scarbrough, Thomas <Thomas.Scarbrough@nrc.gov>

Subject:

RE: DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx Let's use the 2pm meeting to try to finalize the Preliminary Finding From: Clark, Theresa Sent: Wednesday, July 27, 2016 12:41 PM To: Spencer, Michael <Michael.Spencer@nrc.gov>; Holahan, Gary <Gary.Holahan@nrc.gov>; West, Steven

<Steven.West@nrc.gov>; Scarbrough, Thomas <Thomas.Scarbrough@nrc.gov>

Subject:

Re: DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx Something to discuss if we do meet at 2 :)

On: 27 July 2016 12:02, "Spencer, Michael" <Michael.Spencer@nrc.gov> wrote:

Another thought. Should we address this new ASME Code argument that has surfaced in discussions with DE Staff? It seems like we should for completeness, and if we don't, we will probably get that back as a comment.

From: Spencer, Michael Sent: Wednesday, July 27, 2016 11:11 AM To: Ho,lahan, Gary <Gary.Holahan@nrc.gov>; West, Steven <Steven.West@nrc.gov>; Scarbrough, Thomas

<Thomas.Scarbrough@nrc.gov>; Clark, Theresa <Theresa.Clark@nrc.gov>

Subject:

DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx I made comments on top of Theresa's.

1

July 27, 2016 Exelon Backflt Appeal Panel Preliminary Findings The compliance fMASt Jexception to the Backfit Rule is intended to address failures to meet known and established Commission st andards because of omission or mistake of fact. New or modified interpretations of what const i tutes compliance do not fall within the exception. The Panel concludes that in 2001 and 2004 there was no !!known and established standard of the Commission!! relating to the potential of pressurizer safety valves (SVs} to fail following water discharge, fer allllr1&slAB peteAtlal PJ failwres during Inadvertent Operation of Emergency Core Cooling System (ECCS} events.

During the Exelon power uprate review in 2001 and the review of a later valve setpoint amendmentl.MeF In 2004, the staff exercised reasonable and well-informed, engineering Judgement when concluding that the SVs were unlikely to fail.stick open (i.e., fall to reseat). The position on valve qualification in the 2015 backfit is a new or modified interpretation of what constitutes compliance.

In the absence of an SV failure to &lesereseat, all-the concerns articulated in the backfit relatedrelatieA to event classification, event escalation, and compliance with 10 CFR S0.34(b) and General Design Criteria 15, 21, and 29 aRd 10 CFR so a4(s) are lrrelewAtno longer at lssue.

The panel findings support the Exelon appeal.

The panel's finding relative to t reat ment of SV failure potential derives from t he following:

~1971 IMAS2J10 CFR Part SO Appendix A, ffootnote 21 " The conditions under which a single f ailure of a passive component in a fluid system should be considered in designing the system against a single failure are under development." (emphasis added ).

1977 tsECY-:77-439~1" Application of the /single failure{MAS3J[ concept is complicated by the interrelationships between the various fluid and electrical systems and their supporting auxiliaries in a nuclear power plant. Furthermore, there is a need to stipulate the events and associated assumptions which must be considered during application of the Single Failure Criterion". ;(MAS4J (emphasis added).

1977 ~SECY-77-439} {on aAdditional P2assive fFailuresl;. "During subsequent years (1969-1977]

staff assumptions regarding the nature of passive failures which should be considered have not been completely consistent and there has been some disagreement. However, on the basis of the licensing review experience accumulated in the period since 1969, it has been judged in most instances that the probabilit y of most t ypes of passive failures in fluid systems is sufficiently small that they need not be assumed in addition to the initiating failure in application of the Single Failure Criterion to assure safety of a nuclear power plant". (emphasis added).

=The 1977 SECY paper st resses the use of engineering j 1:1elgement judgment relating to the probability of failure and never ancedoes not suggests t hat va lve "certification" or "qualification" in accordance with American Society of Mechanica l Engineers (ASM E) st anda rds lfMAss1should be evoked as basis for such decisions.

1979 TMI Action Plan item 11.-0.-lL. {$hart terAI L.esseAs LearA1H* ;u,3) Performance testiR& Testing of BWR and PWR Relief and Safety Valve[MAS6J Pesitien: PresS[MAS7Jurized-wat er reactor and boiling-wat er

reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents. NUREG-0737 ~pecifiedJSTSJ provisions for then-operating nuclear power plants and applicants for operating licenses and holders of construction permits to address the TMI action items, including item 11.D.1.

NUREG-0737 stated, for the performance testing of relief and safety valves for item 11.D.1. that the testing should demonstrate that the valves will open and reclose under the expected flow conditions with reference to EPRI testing and other generic industry test programs. Although limited in scope. the test results did not identify any generic issues with PSVs or PORVs sticking open following water relief.

The historical record shows [MAS9Jthat the word "qualify" in this TMI item was not intended to refer to ASME valve certification or qualification. Instead, "qualify" was used in a less formal sense to refer to a reasonable assurance that the valve would open to relieve pressure and then reliably reseat!Jill iA 01e

1$flkl. As referenced in NUREG-0737, the EPRI test program was the widely used approach to address TMI Action Item 11.0.1 at PWR nuclear power plants. In a letter dated July 27.1982. the WestinRhouse Owners Group submitted WCAP-10105 that provided the acceptability of the EPRI testins program for PSVs and PORVs in WestinRhouse-desisned PWRs. [MaA\l PWRs? A.II P'NRs?] 1 iAelweliAg &yr=eR aRd 8Faielweee. Felled BR ReR A,M~ testiRg perf8FMe£1 by tl:le Eleetrie Pa~*.1er ReseaFel:I IRstitwte (l!PRI),

1988 Letter from L. N. Olshan fNRC) to H. E. Bliss (ComEd),~!!NUREG 0737, Item 11.D.1, Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and Z,~!! found reE11:1ired the licensee's reliance on EPRI testing of SVs to be acceptable. (emphasis added). (add Braidwood letter reference]

In 1993 and 1994, Westinghouse (WECl sent Nuclear Safety Advisory Letter NSAL-93-b13lsr101 (June 30. 1993)and NSAL-93-013, Supplement 1 (October28, 1994)to operating nuclear power plants in response to its discovery that potentially nonconservative assumptions were used in the licensing analysis of the lnadverent Operation of the ECCS at Power <IOECCS) accident. WEC recommended that licensees determine if their Pressurizer Safety Relief Valves

{PSRVs) are capable of closing following discharge of subcooled water. WEC noted that the PSRVs might have been designed or "qualified" to relieve subcooled water. WEC indicated that water relief through the PORVs is not a concern, because the PORV block valves can be used to isolate the PO RVs if they fail to close. If the PS RVs are not designed or qualified for subcooled water relief. WEC recommended that licensees re-evaluate the IOECCS accident with three possible options of (1) reducing ECCS flow used in the safety analysis, (2) using a less restrictive operator response time, or (3) crediting the use of one or more PORVs to help mitigate the accident. In Supplement 1 to NSAL-93-013. Westinghouse alerted licensees to potential reduced time for operator action if a positive displacement pump is in service. and to the need to qualify the PSRVs and the piping downstream of the PSRVs and PORVs if water relief from the pressurizer is predicted. Some licensees submitted license amendments that involved improvements to the PORVs and their circuitry to avoid water relief through the PSVs (e.g., Oiablo Canyon in 1996. Salem in 1997, Millstone 3 in 1998. and Callaway in 2000). The NRC staff review of those proposed improvements relied on engineering judgement of the various test information and PORV circuitry upgrades described by individual licensees.

Exelon submitted a license amendment for similar PORV improvements that was later withdrawn.

2001 Exelon power uprate~- _NRC staff approved Inadvertent Operation of ECCS 101:SSC {IQECCS) analysis which included QPressurizer filling, SV water discharge, ECCS termination,. and SV closure.e..J.n

[MAs111suoport of the 2015 backfit. the staff suggests that the 2001 license amendment was predicated on the NRC's mistaken belief that the valves were ASME qualified. However. a review of the safety evaluation and associated RAls shows that the staff was well aware of the nature of the EPRI testing

being relied on. The panel's conclusion was confirmed via discussions with !Frank's position back then],

who was !role in the 2001 license amendment]. !Frankl informed the Panel that the 2001 license amendment was based on the exercise of staff engineering judgment and there was no discussion of ASME qualification of valves.- In addition. the NRC approved power uprates for other nuclear power plants that included NRC staff evaluation of water relief through PORVs or PSVs based on test information provided by individual licensees (e.g .* Shearon Harris in 2001 and Beaver Valley in

~OO§!su2J).

During the NRC evaluations of license amendments since the TMI accident. the staff has specified in some SERs that a PORV or PSV would be assumed to stick open if it was not qualified for liquid service.

To address this concern. the staff reviewed and accepted a variety oftest information (including EPRI.

Wyle, and vendor testing) from individual licensee for the capability of PORVs or PSVs to reseat following water relief. Contrary to the October 2015 backfit decision that application of the ASME BPV Code and OM Code is necessary to support water qualification of PSVs. a specific requirement for the PO RVs or PSVs to meet the ASME BPV Code for water certification was not found in the reviewed sample of SE Rs. The current staff position that certification to the ASME BPV Code requirements for water relief is the only acceptable method to justify an assumption that a PORV or PSV will reseat following water relief is not consistent wlth the general practice of the staff found during the review of past license amendments for Byron/Braidwood or other nuclear power plants.

20[MASBJ05 RIS 2005--029; states, "The NRC staff's position is noted in the power uprate review standard [RS-0012003], as follows: 'For the inadvertent operation of emergency core cooling system and chemical and vofume control system malfunctions that increase reactor coolant inventory events: (a) non-safety-grade pressure-operated relief valves should not be credited for event mitigation and (b) pressurizer level should not be allowed to reach a pressurizer water-so/id condition."' (emphasis added).

However, the Power Uprate Review Standard (RS-001 2003} also states, "The staff does not intend to impose the criteria and/or guidance in this review standard on plants whose design bases do not include these criteria and/or guidance. No backfitting is intended or approved in connection with the issuance of this review standard. "[MAS14J This intent ("no backfit") was confirmed in personal discussions with the NRR manager responsible for developing and issuing RS-001. Therefore, contrarv to the RIS statement, neither the RS-001 review standard nor the RIS 2005-29 documented "known and established standards of the Commission:'[MAS15JIMAS16J In summary:

The NRR 2015 compliance backfit finding (October 9, 2015 letter to Exelon) hiAges @RtiFel*tis predicated

&R-on the following position: JMAS17!"0ne assumption that is particularly important to the non*esca/ation criteria is that water relief through a valve that is not qualified for water relief will cause that valve to stick in its ful!v open position". The backfit finding also asserts "the licensee has invoked the PSVs

[Sa~t*t ValvesSVs] as a mitigation system but has not applied the single-failure assumption (required in accident analyses to show compliance with GDC 21) to that system (i.e., failure of a PSVta close) nor have they provided ASME water qualification documentation for the PSVs, causing the staff to be unable to conclude that there is compliance with GDC 21." (emphasis added). AAa goes eA tl:le eall JerThe backfit evaluation further calls for "the ASME. .. original Overpressure Protection Report and "in service test history... including both water and steam tests"

However, none of these positions; Aet 1J:-1e "as&1:1FMptien af faihue eA water relief', ABF the eall fer 1:1se et tl=le siRgle fai11:1Fe re'11:1ireFAeAt iA GQC 21, Aer t~e "Aeee fer ASME Cade eertifieatieA", were "known and established standards of the Commission" in 2001 or 2004 for determining when it was appropriate to assume a &a~t\' valve failure of SVs to reseat. In fact, they do not appear to be "established standards of the Commission" at present, since they have not undergone any appFBf!Fia,e ageAE'{ geRerie preeessreview for issuance as generic guidance (e.g.l A.cule, Regulatory Guide, or Standard Review Plan}.

Assertions made in a Regulatory Information Summary (RIS) cannot be taken as "established standards;'T (will need confirmation from Michael on this] [MAS1s]The panel concludes that the r,JF8J!l8Sed positions ftaken to support the compliance backfit finding} represent new and different staff views on how to address potential safet)' *,alveSV failures following water discharge. Although they represent well-intentioned staff positions that could provide additional safety marginRegarltless ef tl=te &888 iAter:itieRS aAEI tJal1:1e ef atlElitieRal AllaJSiA assetiatee with tl=te e1:1rreAt ~esitieRs, they do not provide a basis for a compliance backfit.

In addition to the specific finding relating to the backfit appeal, the panel believes it is important to acknowledge that water discharge through an SV saJety *,alt.*e not specifically designed for such service, is undesirable and should be minimized or avoided as a matter of conservative engineering and prudent operations. The panel concludes this while fully aware that the event sequence being considered appears to be of little safety significance (the panel has requested RES analysis to confirm this belief).

Operator training and emergency procedures to terminate the event before pressurizer filling, aAS-as well as the use of power-operated relief valves?QRV5 to terFMiAate e>JeAt5 ,rier te e*.<er RlliAg rather than relying solely on SVs, are clearly preferred, whether they form the facilities' UFSAR licensing basis or not.

From: Spencer, Michael Sent: Wednesday, July 27, 2016 11:11 AM To: Holahan, Gary; West, Steven; Scarbrough, Thomas; Clark, Theresa

Subject:

DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx Attachments: DRAFT Preliminary Findings July 27 2016 - tvc MAS.docx I made comments on top of Theresa's.

July 27, 2016 Exelon Backfit Appeal Panel Preliminary Findings The compliance JM11s11exception to the Backfit Rule is intended to addre5s failures to meet known and established Commission standards because of omission or mistake of fact. New or modified interpretations of what constitutes compliance do not fall within the exception. The Panel concludes that in 2001 and 2004 there was no !!known and established standard of the Commission! relating to the potential of pressurizer safety valves (SVs) to fail following water discharge, fer addressiRg:

pe,eMial i\' tailwn:11, during Inadvertent Operation of Emergency Core Cooling System (ECCS1 events.

During the Exelon power uprate review in 2001 and the review of a later valve setpoint amend me ntlatef: in 2004t the staff exercised reasonable and well-informed, engl neering judgement when concluding that the SVs were unlikely tot.stick open (i.e., fail to reseat). The position on valve qualification in the 2015 backfit is a new or modified interpretation of what constitutes compliance.

In the absence of an SV failure to do&ereseat, 3'1-the concerns articulated in the backfit relatedFelatieR to event classification, event escalation, and compliance with 10 CFR S0.34(b) and General Design Criteria 15, Zl, and 29 an£1 lQ CFR SQ 34(hl are iFFeleuaRtno longer at issue.

The panel findings support the Exelon appeal.

The panel's finding relative to treatment of SV failure potential derives from the following:

~1971 !MAS2Jl0 CFR Part 50 Appendix A£ fk>otnote 2.: "The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development." (emphasis added).

1977 {SECY-:77-439},;. "Application of the {single failure[MAS3Jl concept is complicated by the interrelationships between the various fluid and electrical systems and their supporting auxiliaries in a nuclear power plant Furthermore, there is a need to stipulate the events and associated assumptions which must be considered during application of the Single Failure Criterion"; [MAS4J (emphasis added).

1977 ~SECV-77-4.39} (on aAdditional .P.J!asslve fFailurest "During subsequent years [1969-1977]

staff assumptions regarding the nature of passive failures which should be considered have not been completely consistent and there has been some disagreement. However, on the basis of the licensing review experience accumulated in the period since 1.969, it has been judged in most instances that the probability of most types of passive failures in fluid systems is sufficiently small that they need not be assumed in addition ta the initiating failure in application of the Single Failure Criterion to assure safety of a nuclear power plant". (emphasis added).

=The 1977 SEC'i' paper stresses the use of engineering judgement iudgment relating to the probability of failure and Ae*,er ancedoes not suggests that valve "certification" or "qualification" in accordance with American Society of Mechanical Engineers (ASME) standards [MASSJshould be evoked as basis for such decisions.

1979 TMI Action Plan item 11.-0.-lL (S~ert te,m lesseRs learnelill 2.1.at Performance le&tiRg Testing of BWR and PWR Relief and Safety ValvecMAS6J Position: Press[MAS7Jurized-water reactor and boiling-water

reactor licensees and applicants shall conduct testing to qualify the reactor coolant system reUef and safety valves under expected operating conditions for design-basis transients and accidents. The historical record shows (MASs1that the word "qualify" in this TMI item was not intended to refer to ASME valve certification or qualification. Instead, qualify" was used in a less formal sense to refer to (fill in the blank). !Many PWRs? All PWRs?j, including Byron and Braidwood, relied on non-ASME testing performed by the Electric Power Research Institute (EPRll.

1988 Letter from L. N. Olshan (NRC) to H. E. Bliss (ComEd),~!!NUREG 0737, Item 11.0.1, Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and 2,~:. found reE11,1irea the licensee's reliance on EPRI testing of SVs to be acceptable. (emphasis added). [add Braidwood letter reference) 2001 Exelon power uprate~- _NRC staff approved Inadvertent Operation of ECCS IOESSC (IQECCS) analysis which included QP,ressurizer filling, SV water discharge, ECCS termination.,, and SV closure.a.Jn IMAS9!support of the 2015 backfit, the staff suggests that the 2001 license amendment was predicated on the NRC's mistaken belief that the valves were ASME qualified. However, a review of the safety evaluation and associated RAIS shows that the staff was well aware of the nature of the EPRI testing being relied on. The panel's conclusion was confirmed via discussions with (Frank's position back then],

who was [role in the 2001 license amendment). (Frank] informed the Panel that the 2001 license amendment was based on the exercise of staff engineering judgment and there was no discussion of ASME qualification of valves.

20[MAS101os RIS 200S-029;, states, "The NRC staff's position is noted in the power uprate review standard [RS-0012003], as follows: 'For the inadvertent operation of emergency core cooling system and chemical and volume control system malfunctions that increase reactor coolant inventory events: {a) non-safety-grade pressure-operated relief valves should not be credited for event mitigation and {b) pressurizer level should not be allowed to reach a pressurizer water-solid condition.'" (emphasis added).

However, the Power Uprate Review Standard (RS-001 2003) also states, "The staff does not intend to impose the criteria and/or guidance in this review standard on plants whose design bases do not include these criteria and/or guidance. No backfitting is intended or approved in connection with the issuance of this review standard. "(MAS111 This intent ("no backfit") was confirmed in personal discussions with the NRR manager responsible for developing and issuing RS-001. Therefore, contrary to the RIS statement, neither the RS-001 review standard nor the RIS 2005-29 documented "known and established standards of the Commission/[MAS12]1MAS131 In summary:

The NRR 2015 compliance backfit finding (October 9, 2015 letter to Exelon) AiAges eRtiirel*,iis predicated Sff-On the following position: IMASt4!"0ne assumption that is particularly important to the non-escalation criteria is that water relief through a valve that is not qualified for water relief will cause that valve to stick in its fully open position". The backfit finding also asserts "the licensee has invoked the PSVs

[Safety \<alt.*esSVs] as a mitigation system but has not applied the single-failure assumption (required in accident analyses to show compliance with GDC 21} to that system (i.e., failure of a PSV to close) nor have they provided ASME water qualification documentation for the PSVs, causing the staff to be unable to conclude that there is compliance with GDC 21." (emphasis added). And gees eR Uie call ferThe

backfit evaluation further calls for "the ASME. .. original Overpressure Protection Report" and "in service test history... including both water and steam tests" However, none of these positions; Ret tl=le "ass1:1mptieA ef fail.,.Fe &Fl water relief', Aer ttle e.all fer wse et the siRgle failure requireFMeAt iA GQC U, Aer the "Reed fer ASM~ Ceee ceFtifieatieA", were "known and established standards of the Commission" in 2001 or 2004 for determining when it was appropriate to assume a saf@t'/ val>.*e failure of SVs to reseat. In fact, they do not appear to be "established standards of the Commission" at present, since they have not undergone aR-; a~prepFiate ageAe'J' geRerie preeessreview for issuance as generic guidance (e.g.,. ~rule, Regulatory Guide, or Standard Review Plan).

Assertions made in a Reeulatory Information summary (RIS) cannot be taken as "established standards~".,. [will need confirmation from Michael on thisj [MAs1sJThe panel concludes that the pFapesee positions ~taken to support the compliance backfit finding} represent new and different staff views on how to address potential safety valveSV failures following water discharge. Although they represent well-intentioned staff positions that could provide additional safety marginR.egaFdless ef the geee iRteRtieAs aAe ';/al1:1e ef aeElitieAal ffiaFgiA asse£iated with tile et1FFeAt pesitiens, they do not provide a basis for a compliance backfit.

In addition to the specific finding relating to the backfit appeal, the panel believes it is important to acknowledge that water discharge throueh an SV safety *.*al*,e not specifically designed for such service, is undesirable and should be minimized or avoided as a matter of conservative engineering and prudent operations. The panel concludes this while fully aware that the event sequence being considered appears to be of little safety significance (the panel has requested RES analysis to confirm this belief).

Operator training and emergency procedures to terminate the event before pressurizer filling, aA&-as well as the use of power-operated reliefvalvesPQIWs te termiRate e*.ieRts pFier te B'o'er filliA@ rather than relying solely on SVs, are clearly preferred, whether they form the facilities' UFSAR licensing basis or not.

    • 00 - * - *...-. ** * * * * , . * * ** * ** * - *....- ~ . ,.... ~~#ft *. ..,.,....,,,..,~.,...,....... .......___o_...w.,,~,1 ........-..... ,1 .."DL***~.,...,.,..r...,.,..,,. _......~ **, ***--

Subject:

Backfit Appeal Panel Meeting Location: 0 -1784 Start: Wed 08/24/2016 11:00 AM End: Wed 08/24/2016 11:30 AM Recurrence: (none)

Meeting Status: Meeting organizer Organizer: Ho lahan, Gary Required Attendees: West, Steven; Scarbrough, Thomas; Spencer, Michael; Mccree, Victor; Clark, Theresa Optional Attendees: Johnson, Michael; ConferenceRoom0 17B4 Resource Scheduled by Psprogeris 8/11/16 POC: Theresa Clark REQUEST: baclcfit appeal panel ...

1

From: Clark, Theresa Sent: Thursday, August 11, 2016 1:04 PM To: Sprogeris, Patricia Cc: Holahan, Gary; West, Steven; Scarbrough, Thomas; Spencer, Michael

Subject:

REQUEST: backfit appeal panel meeting w/ Vic

Patti, Can you please arrange a meeting for the backfit appea l panel (Gary Holahan, Steve West, Tom Scarbrough, and Michael Spencer) with Vic? Mike may also wish to attend. The week of August 22 would be ideal, perhaps 11am 8/24 for ha lf an hour if Gary and Steve can make that work (others are free) . Otherwise, please work your magic t o find a time. Thanks!

Theresa Valentine Clark Executive Technical Assistant (Reactors)

U.S. Nuclear Regulatory Commission Theresa.Clark@n rc.gov I 301-415-4048 I 0 -16E22

From: Holahan, Gary Sent: Tuesday, November 01, 2016 11:39 AM To: Clark, Theresa

Subject:

FW: Nuclear Energy Institute. Comments on September 13, 2016 Pub lic Meeting with Committee to Review Generic Requirements (CRGR)

Attachments: Safety Spotlight 110316 - Backfit Appeal (TVC) Comments 2016 11 01.docx

Theresa, (b)(S)

Comments attached:

General comment, you are presenting because were on the Panel. You should present Panel and EDO positions ... that's why you are taking comments. All other comments are editorial. .. on a difficult 10 minute topic.

Gary From: Dean, Bill Sent: Monday, October 31, 2016 5:40 PM To: Hackett, Edwin <Edwin.Hackett@nrc.gov>

Cc: Johnson, Michael <Michael.Johnson@nrc.gov>; McDermott, Brian <Brian.McDermott@nrc.gov>; Lubinski, John

<John.Lubinski@nrc.gov>; Holahan, Gary <Gary.Holahan@nrc.gov>

Subject:

Re: Nuclear Energy Institute Comments on September 13, 2016 Public Meeting with Committee to Review Generic Requirements (CRGR) j(b)(S)

Th~nh Pfl-1 (b)(S)

On: 31 October 2016 15:56, "Hackett, Edwin" <Edwin.Hackett@nrc.gov> wrote:

1

FYI - Forwarding the latest Letter from NEI on backfitting, reacting to the CRGR Public Meeting and the Exelon decision.

Ed From: PIETRANGELO, Tony (mailto:arp@nei.org]

Sent: Monday, October 31, 2016 3:11 PM To: Hackett, Edwin <Edwin.Hackett@nrc.gov>

Cc: Mccree, Victor <Victor.McCree@nrc.gov>; Dean, Bill <Bill.Dean@nrc.gov>; West, Steven <Steven.West@nrc.gov>;

McDermott, Brian <Brian.McDermott@nrc.gov>; Moore, Scott <Scott.Moore@nrc.gov>; Ordaz, Vanna

<Vonna.Ordaz@nrc.gov>; Wert, Leonard <Leonard.Wert@nrc.gov>; Williamson, Edward

<Edward.Williamson@nrc.gov>; 'Annette.Vetti-Cook@nrc.gov' <Annette.Vetti-Cook@nrc.gov>

Subject:

[External_Sender] Nuclear Energy Institute Comments on September 13, 2016 Public Meeting with Committee to Review Generic Requirements (CRGR)

THE ATTACHED PDF CONTAIN THE COMPLETE CONTENTS OF THE LETTER October 31, 2016 Dr. Edwin M. Hackett Chair, Committee to Review Generic Requirements U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

Nuclear Energy Institute Comments on September 13, 2016 Public Meeting with Committee to Review Generic Requirements (CRGR)

Project Number: 689

Dear Dr. Hackett:

I am writing to thank the CRGR for the opportunity to present industry's views on the agency's backfitting program during the public meeting held on September 13, 2016. We are encouraged by the CRGR's willingness to consider stakeholder input before developing its response to the tasking memorandum issued by the Executive Director for Operations f'ED0'1 on June 9, 2016.ill We would welcome additional opportunities to interact with the CRGR in the future, and would support any additional public meetings that are necessary for the Committee to adequately respond to the tasking memorandum.

Attachment Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer Nuclear Energy Institute 1201 F Street NW, Suite 1100 2

Washington, DC 20004 www.nei.org TAKE THE NEI FUTURE OF ENERGY QUIZ, www.NEl.org/whynuclear FOLLOW USON Sent through www.intermedi9.com w "Tasking Related to Implementation of Agency Backfitting and Issue Finality Guidance," June 9, 2016 f'Tasking Memorandum").

3

"Safety Spotlight" on Backfit Appeal Theresa Clark Quarterly Strategic Alignment Meeting I November 3, 2016 Th*rt*tftKev Messases for "Safety Spotlight" Based on the Appeal Panel's Review:

  • Consistent with the NRC mission and values, it is the staff's right and responsibility to raise safety concerns.
  • It is the agency's responsibility to employ hs processes to evaluate and document the resolution of such concerns.

_*_Evaluation of issues should consider plant licensing basis, industry*wide precedent, and safety significance.

  • . compliance backfits must be associated with a previous staff error in not properly implementing known and establl hed standards not with han e in staff inte retatlon of wha n itutes om Han What was the backfit about?
  • Staff's concern that pressurizer valve failure following water discharge could cause escalation of events to more serious conditions, counter to~ plant licensing basis {e.g. ANS*51.1/N18. 2-19731 predicated on several positions:

o ASME Caae water 1:11,1ali~eationCode qualification of water relief had not been conducted.

o Water relief through an unqualified valve will cause it :to stick fully open:.

o The single-failure criterion in the Regulationsass1,1m13tioA had also not been applied to the valves.

  • Staff 2015 positions differed from those taken for Byron and Braidwood (B/B) in 2001 and 2004 license amendments.
  • The staff determined the 2001/2004 positions were in error and that backfitting was justified under the compliance exception (10 CFR 50.109(al(4)(il). The licensee was directed to take action to resolve the non-compliance.

What is (some ofl the history?

  • 1968*1972: GDCs define AOOs (normal operation to once in plant life..fil!filJ.ll) and fpostulated ~accidents
  • 1970+: ANS (and Westinghouse) formulate ANS Conditions I (normal), II (frequent}, Ill (infrequent), IV (accident) and non-escalation for transient analysis
  • 1979+: TM I Action Plan item 11.D.1 requires "qualification" by testing of pressurizer valves; EPRI testing showed that valves did not stick open on water discharge; NRC wrote safety evaluations for each plant
  • 1993: Westinghouse (NSAL-93-013) identifies analysis problems (no Part 21 or generic NRC action)
  • 1996+: Licensees update analysis under 10 CFR 50.59 or request license amendments with varying approaches including reanalysis, PORV upgrades, safety valve crediting, etc.
  • 2001 and 2004: NRR issues B/B amendments, including credit for safety valve water discharge
  • 2005: RIS-05-029 observes that PWR analyses include errors (e.g., non-safety PORVs, un-qualified valves)
  • 2013: Proposed RAI on a B/B measurement uncertainty uprate declared out of scope
  • 2015: Backfit calling for ASME qualification of safety valves, application of single failure criterion, assumption of valve failure on water discharge What happened next?
  • Backfit was overturned on second appeal to EDD, who agreed with the Backfit Appeal Panel.

2.......Positions taken by the NRC staff in the 2015 backfit decision represent new and different staff views on how to address pressurizer safety valve performance following water discharge.

o 2001/2004 reviews were not in error, they were different from the current staff approach but still well*informed and technical founder decisions o These staff positions are well-intentioned and conservative approaches that could provide ~additiona l safety margin, but they were not "known and established standards of the Commission" at the time of the 2001 and 2004 approvals and do not provide an appropriate basis for a compliance backfit.

o Very small risk reduction would be expected from the backfit (separate from defense*in-depth considerations).

  • NRR is preparing a plan to reassess issues identified in RIS 2005-29 and its draft Revision 1 (due January 2017).

References:

From: Holahan, Gary Sent: Wednesday, August 17, 2016 10:32 AM To: Scarbrough, Thomas Cc: Clark, Theresa

Subject:

transcript discussion of "qualification" The attached 81-page Public Attachments: 2016 Exelon meeting transcript 3-7-2016 ML1GOlOA364*Pdf available Me~ting Transcript is publicly in ADAMS.

Tom ,

Discussion of Exelon's view on valve qualification for water discharge and how it related to ASME ... starts on line 5 p 36 Gary

From: Holahan, Gary Sent: Wednesday, July 27, 2016 12:53 PM To: Holahan, Gary

Subject:

1pm 7/27 DRAFT July 26, 2016 Exelon Backfit Appeal Panel Preliminary Findings The Panel concludes that In 2001 and 2004 there was no "known and established standard of the Commission" relating to water discharge, for addressing potential SV failures during Inadvertent Operation of ECCS events. During the Exelon power uprate review in 2001 and later in 2004, the staff exercised reasonable and well-informed, engineering judgement when concluding that the SVs were unlikely to fail open.

In the absence of a SV failure to close, all the concerns relation to event classification, event escalation, compliance with General Design Criteria 15, 21, 29 and 10 CFR 50 34(b) are irrelevant.

The panel findings support the Exelon appeal.

The panel's finding relative to treatment of SV failure potential derives from the following:

1969 10 CFR 50 Appendix A footnote 2 "The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development. "

1977 (SECY 77-439) "Application of the concept is complicated by the interrelationships between the various fluid and electrical systems and their supporting auxiliaries in a nuclear power plant. Furthermore, there is a need to stipulate the events and associated assumptions which must be considered during application of the Single Failure Criterion".

1977 (SECY 77-439) Additional Passive Failures uouring subsequent years [1969-1977] staff assumptions regarding the nature of passive failures which should be considered have not been completely consistent and there has been some disagreement. However, on the basis of the licensing review experience accumulated in the period since 1969, it hos been judged in most instances that the probability of most types of passive failures in fluid systems is sufficiently small that thev need not be assumed in addition to the initiating failure in application of the Single Failure Criter;on to assure safety of a nuclear power plant".

The 1977 SECY paper stresses the use of engineering judgement relating t o the probability of failure and never once suggests that valve "certification" or "qualificatio n" should be evoked as basis for such decisions.

1979 TMI Action Plan item II. D. 1 (Short-term Lessons Learned 2.1.2) Performance testing of BWR and PWR Relief and Safety Valve Position: Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct t esting to qualify the react or coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.

1988 Letter from L. N. Olshan (NRC) to H. E. Bliss (ComEd),"NUREG 0737, Item 11.0.1, Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and 2," found required testing of SVs t o be acceptable

2001 Exelon power uprate - NRC staff approved IOESSC analysis which included Pressurizer filling, SV water discharge, ECCS termination and SV closure 2005 RIS 2005-029 - states, "The NRC staff's position is noted in the power uprate review standard [RS-001 2003 L as follows: 'For the inadvertent operation of emergency core cooling system and chemical and volume control system ma/functions that increase reactor coolant inventory events: (a) non-safety-grade pressure-operated relief valves should not be credited for event mitigation and (b} pressurizer level should not be allowed to reach a pressurizer water-solid condition.'"

However, t~e Power Uprate Review Standard (RS-0012003) also states, "The staff does not intend to impose the criteria and/or guidance in this review standard on plants whose design bases do not include these criteria and/or guidance. No backfitting is intended or approved in connection with the issuance of this review standard. "

This intent (~no backfit") was confirmed in personal discussions with the NRR manager responsible for developing and issuing RS-001. Therefore, contrary to the RIS statement, neither the RS-001 review standard nor the RtS 2005-29 documented "known and established standards of the Commission" In summary:

The NRR 2015 compliance backfit finding (October 9, 2015 letter to Exelon) hinges entirely on "One assumption that is parlicularly important to the non-escalation criteria is that water relief through a valve that is not qualified for water relief will cause that valve to stick in its fully open position". The backlit finding also asserts *the licensee has invoked the PSVs [Safety Valves] as a mitigation system but has not applied the single-failure assumption (required in accident anafyses to show compliance with GDC 21) to that system (i.e., failure of a PSV to close) nor have they provided ASME water qualification documentation for the PS Vs, causing the staff to be unable to conclude that there is compliance with GDC 21." And goes on the call for "the ASME ...original Overpressure Protection Repoff' and Kin service test history ... including both water and steam tests" However, none of the positions; not the Kassumption of failure on water relief', nor the call for use of the single failure requirement in GDC 21, nor the "need for ASME Code certification", were "known and established standards of the Commission" in 2001 for determining when it was appropriate to assume a safety valve failure. In fact, they do not appear to be "established standards of the Commission" at present, since they have not undergone any appropriate agency generic process (e.g. Rule, Regulatory Guide, or Standard Review Plan}. Assertions made in a Regulatory Information Summary (RIS) cannot be taken as "established standards". The panel concludes that the proposed positions (taken to support the compliance backfit finding) represent new and different staff views on how to address potential safety valve failures following water discharge. Regardless of the good intentions and value of additional margin associated with the current positions, they do not provide a basis for a compliance backfit.

In addition to the specific finding relating to the backfit appeal, the panel believes it is important to acknowledge that water discharge through a safety valve not specifically designed for such service, is undesirable and should be

  • minimized or avoided as a matter of conservative engineering and prudent operations. The panel concludes this while full aware that event sequence being considered appears to be of little safety significance (the panel has requested RES analysis to confirm this belief). Operator training and emergency procedures, and the use of PORVs to terminate events prior to over-filling, are clearly preferred, whether they form the facilities' UFSAR licensing basis or not.

According to MO 8.18, a RIS may "Communicate and/or clarify staff technical or policy positions on regulatory matters that have not been communicated or are not broadly understood by the nuclear industry."

However, "ARIS may NOT-

{i) Provide guidance for the implementation of rules and regulations, 2

(ii) Provide guidance to NRC staff on regulatory or technical matters, (iii) Require a response, commitments, or action, or (ivl Be used in lieu of other established agency products."

Based on this, a RIS may not establish standards for compliance, but it may communicate/clarify standards already established by other appropriate means. In this regard, an open question for me is whether the 2003 RS could be said to establish standards. Until I began working on this panel, I didn't know what an RS was (and the regulatory status of it is still unclear to me).

Also, you say, "[Tlhey do not appear to be "established standards of the Commission" at present." However, the 2007 SRP states, "If the inadvertent operation of the ECCS causes one or more pressurizer power-operated relief valves (PORVs) to open while the pressurizer is water*solid, then the PORV is generally assumed to fail open (i.e., PO RVs are assumed to fail in the open position after having relieved water, if they are not (1) safety-related equipment and (2) qualified for water relief).u The SRP also cites RIS 2005-29.

3

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