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| author name =  
| author name =  
| author affiliation = NRC/RGN-III/DRP
| author affiliation = NRC/RGN-III/DRP
| addressee name = Schimmel M A
| addressee name = Schimmel M
| addressee affiliation = Northern States Power Co
| addressee affiliation = Northern States Power Co
| docket = 05000282, 05000306, 07200010
| docket = 05000282, 05000306, 07200010
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See also: [[followed by::IR 05000282/2011010]]
See also: [[see also::IR 05000282/2011010]]


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:UNITED STATES
[[Issue date::June 9, 2011]]
                            NUCLEAR REGULATORY COMMISSION
                                              REGION III
                                2443 WARRENVILLE ROAD, SUITE 210
                                        LISLE, IL 60532-4352
                                            June 9, 2011
EA-11-110
Mr. Mark A. Schimmel
Site Vice President
Prairie Island Nuclear Generating Plant
Northern States Power Company, Minnesota
1717 Wakonade Drive East
Welch, MN 55089
SUBJECT:       PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2,
                NRC INSPECTION REPORT 05000282/2011010; 05000306/2011010
                PRELIMINARY WHITE FINDING
Dear Mr. Schimmel:
On May 20, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents
the results of this inspection, which were discussed on May 20, 2011, with you and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents a finding for Unit 1 that has preliminarily been determined to be White or
a finding with low-to-moderate increased safety significance. In addition, this same finding was
preliminarily determined to be Green, a finding of very low safety significance, for Unit 2.
As documented in Section 4OA5 of this report both trains of safety-related battery chargers
were not capable of performing their safety function from initial installation in 1994 due to being
susceptible to failure during certain design basis events. This finding was assessed based on
the best available information, including influential assumptions, using the applicable
Significance Determination Process (SDP).
Upon identification of this issue and after interaction with the NRC, you concluded that a
designated operator position needed to be established to ensure that a specific individual could
perform actions to recover the battery charger(s) prior to the safety-related batteries being


EA-11-110 Mr. Mark Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company, Minnesota 1717 Wakonade Drive East Welch, MN 55089
M. Schimmel                                    -2-
depleted. Lastly, during your past operability review you concluded that there was reasonable
doubt that the battery chargers would have performed their safety function if called upon prior to
October 22, 2010, (the date the designated operator position was established). Because of the
compensatory actions taken, no current safety concern exists.
This finding is also an apparent violation of NRC requirements and is being considered
for escalated enforcement action in accordance with the NRC Enforcement Policy.
The current Enforcement Policy can be found at the NRCs Web site at
http://www.nrc.gov/reading-rm/doc-collections/enforcement.
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our
evaluation using the best available information and issue our final determination of safety
significance within 90 days of the date of this letter. The SDP encourages an open dialogue
between the NRC staff and the licensee; however, the dialogue should not impact the timeliness
of the staffs final determination.
Before the NRC makes its enforcement decision, we are providing you an opportunity to either:
(1) present to the NRC your perspectives on the facts and assumptions used by the NRC to
arrive at the finding and its significance at a Regulatory Conference, or (2) submit your position
on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held
within 30 days of the receipt of this letter and we encourage you to submit supporting
documentation at least one week prior to the conference in an effort to make the conference
more efficient and effective. If a conference is held, it will be open for public observation.
The NRC will also issue a press release to announce the conference. If you decide to submit
only a written response, such submittal should be sent to the NRC within 30 days of the receipt
of this letter. If you decline to request a Regulatory Conference or to submit a written response,
you relinquish your right to appeal the final SDP determination; in that, by not doing either you
fail to meet the appeal requirements stated in the Prerequisite and Limitation Sections of
Attachment 2 of IMC 0609.
Please contact John Giessner at (630) 829-9619 in writing within 10 days of the date of this
letter to notify the NRC of your intended response. If we have not heard from you within
10 days, we will continue with our significance determination and enforcement decision.
The final resolution of this matter will be conveyed in separate correspondence.
Since the NRC has not made a final determination in this matter, no Notice of Violation is
being issued for this inspection finding at this time. Please be advised that the number and
characterization of the apparent violation described in the enclosed inspection report may
change as a result of further NRC review.


SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, NRC INSPECTION REPORT 05000282/2011010; 05000306/2011010 PRELIMINARY WHITE FINDING
M. Schimmel                                  -3-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website
at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
                                            Sincerely,
                                            /RA/
                                            Steven West, Director
                                            Division of Reactor Projects
Docket Nos.: 50-282; 50-306; 72-010
License Nos.: DPR-42; DPR-60; SNM-2506
Enclosure:    Inspection Report 05000282/2011010; 05000306/2011010
                w/Attachment: Supplemental Information
cc w/encl:    Distribution via ListServ


==Dear Mr. Schimmel:==
          U.S. NUCLEAR REGULATORY COMMISSION
On May 20, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents the results of this inspection, which were discussed on May 20, 2011, with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. This report documents a finding for Unit 1 that has preliminarily been determined to be White or a finding with low-to-moderate increased safety significance. In addition, this same finding was preliminarily determined to be Green, a finding of very low safety significance, for Unit 2. As documented in Section 4OA5 of this report both trains of safety-related battery chargers were not capable of performing their safety function from initial installation in 1994 due to being susceptible to failure during certain design basis events. This finding was assessed based on the best available information, including influential assumptions, using the applicable Significance Determination Process (SDP).
                          REGION III
Docket Nos:         50-282; 50-306; 72-010
License Nos:        DPR-42; DPR-60; SNM-2506
Report Nos:        05000282/2011010; 05000306/2011010
Licensee:          Northern States Power Company, Minnesota
Facility:          Prairie Island Nuclear Generating Plant, Units 1 and 2
Location:          Welch, MN
Dates:              May 13 through 20, 2011
Inspectors:        K. Stoedter, Senior Resident Inspector
                    P. Zurawski, Resident Inspector
                    C. Brown, Reactor Engineer
                    L. Kozak, Senior Reactor Analyst
Observer:          S. Lynch, Nuclear Safety Professional Development
                    Program Participant
Approved by:        John B. Giessner, Chief
                    Branch 4
                    Division of Reactor Projects
                                                                      Enclosure


Upon identification of this issue and after interaction with the NRC, you concluded that a designated operator position needed to be established to ensure that a specific individual could perform actions to recover the battery charger(s) prior to the safety-related batteries being depleted. Lastly, during your past operability review you concluded that there was reasonable doubt that the battery chargers would have performed their safety function if called upon prior to October 22, 2010, (the date the designated operator position was established). Because of the compensatory actions taken, no current safety concern exists. This finding is also an apparent violation of NRC requirements and is being considered for escalated enforcement action in accordance with the NRC Enforcement Policy. The current Enforcement Policy can be found at the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/enforcement. In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our evaluation using the best available information and issue our final determination of safety significance within 90 days of the date of this letter. The SDP encourages an open dialogue between the NRC staff and the licensee; however, the dialogue should not impact the timeliness of the staff's final determination. Before the NRC makes its enforcement decision, we are providing you an opportunity to either: (1) present to the NRC your perspectives on the facts and assumptions used by the NRC to arrive at the finding and its significance at a Regulatory Conference, or (2) submit your position on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held within 30 days of the receipt of this letter and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If a conference is held, it will be open for public observation. The NRC will also issue a press release to announce the conference. If you decide to submit only a written response, such submittal should be sent to the NRC within 30 days of the receipt of this letter. If you decline to request a Regulatory Conference or to submit a written response, you relinquish your right to appeal the final SDP determination; in that, by not doing either you fail to meet the appeal requirements stated in the Prerequisite and Limitation Sections of Attachment 2 of IMC 0609. Please contact John Giessner at (630) 829-9619 in writing within 10 days of the date of this letter to notify the NRC of your intended response. If we have not heard from you within 10 days, we will continue with our significance determination and enforcement decision. The final resolution of this matter will be conveyed in separate correspondence. Since the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this time. Please be advised that the number and characterization of the apparent violation described in the enclosed inspection report may change as a result of further NRC review. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
                                        TABLE OF CONTENTS
SUMMARY OF FINDINGS ......................................................................................................... 1
  4.    OTHER ACTIVITIES .................................................................................................... 3
        4OA3    Follow-Up of Events and Notices of Enforcement Discretion (71153) ............. 3
        4OA5    Other Activities ............................................................................................... 3
        4OA6    Management Meetings ................................................................................... 9
SUPPLEMENTAL INFORMATION............................................................................................. 1
Key Points of Contact ............................................................................................................. 1
List of Items Opened, Closed and Discussed ......................................................................... 1
List of Documents Reviewed .................................................................................................. 1
List of Acronyms Used ............................................................................................................ 4
                                                                                                                      Enclosure


Sincerely,/RA/
                                      SUMMARY OF FINDINGS
Steven West, Director Division of Reactor Projects Docket Nos.: 50-282; 50-306; 72-010 License Nos.: DPR-42; DPR-60; SNM-2506
IR 05000282/2011010; 05000306/2011010; 05/13/21011 - 05/20/2011; Prairie Island Nuclear
Generating Plant, Units 1 and 2; Other Activities.
This report covers the review of a potential common cause failure of the safety-related battery
chargers. The inspectors identified one apparent violation (AV) with a preliminary significance
of White for Unit 1 and a preliminary significance of Green for Unit 2. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter
(IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not
apply may be Green or be assigned a severity level after NRC management review. The NRCs
program for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
A.      NRC-Identified and Self-Revealed Findings
        Cornerstone: Mitigating Systems
    *  Preliminary White. An apparent violation of Technical Specification (TS) 3.8.4 was
        identified by the inspectors due to the licensees failure to maintain the train A and
        train B direct current electrical power subsystems operable while operating the reactor in
        Modes 1 through 4. Specifically, the licensee installed safety-related battery chargers
        which were susceptible to failure during certain design basis events. This issue was
        entered into the licensees corrective action program (CAP) as CAP 1250561. Upon
        identifying this issue, the licensee performed an operability evaluation and determined
        that the battery chargers remained operable because procedures were in place to
        recover the battery chargers if a failure occurred. After further interaction with the NRC,
        the licensee concluded that a designated operator position needed to be established to
        ensure that a specific individual would perform the battery charger recovery actions prior
        to the safety-related batteries being depleted. Long term corrective actions included
        replacing all four battery chargers.
        This finding was determined to be more than minor because it was associated with the
        design control and equipment performance attributes of the Mitigating Systems
        Cornerstone. In addition, this performance deficiency impacted the cornerstone
        objective of ensuring the availability, reliability, and capability of systems that respond to
        initiating events to prevent undesirable consequences. The inspectors performed a
        Phase 1 SDP evaluation and determined that a Phase 2 evaluation was required
        because this finding represented an actual loss of safety function of a single train of
        equipment for greater than the TS allowed outage time. The inspectors performed a
        Phase 2 evaluation using the pre-solved SDP worksheets for Prairie Island and
        determined that this finding screened as Red. A Phase 3 SDP evaluation was required
        to assess reasonable credit for recovery by operators. The results of the Phase 3 SDP
        evaluation showed that this finding was preliminarily determined to be White for Unit 1,
        and Green for Unit 2. No cross-cutting aspect was assigned to this finding because
        licensee decisions made in regards to evaluating the performance of the battery
        chargers were made many years ago and therefore, not reflective of current plant
        performance. (Section 4OA5.1)
                                                  1                                          Enclosure


===Enclosure:===
B. Licensee-Identified Violations
Inspection Report 05000282/2011010; 05000306/2011010
  No violations of significance were identified.
                                            2    Enclosure


===w/Attachment:===
                                        REPORT DETAILS
Supplemental Information cc w/encl: Distribution via ListServ Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket Nos: 50-282; 50-306; 72-010 License Nos: DPR-42; DPR-60; SNM-2506 Report Nos: 05000282/2011010; 05000306/2011010 Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2 Location: Welch, MN Dates: May 13 through 20, 2011 Inspectors: K. Stoedter, Senior Resident Inspector P. Zurawski, Resident Inspector C. Brown, Reactor Engineer L. Kozak, Senior Reactor Analyst Observer: S. Lynch, Nuclear Safety Professional Development Program Participant Approved by: John B. Giessner, Chief Branch 4 Division of Reactor Projects Enclosure  
4.    OTHER ACTIVITIES
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153)
.1  (Closed) Licensee Event Report 05000282/2010-004: Battery Charger Inoperability due
      to Potential Undervoltage Conditions
  a. Inspection Scope
      The inspectors reviewed the licensees response to discovering that safety-related
      battery chargers installed in 1994 were susceptible to failure during certain design basis
      accidents. Specifically, these battery chargers had the potential to stop providing an
      output, or lock-up, if their alternating current input voltage dropped below the
      nameplate minimum voltage of 90 percent at the battery charger motor control center
      (MCC). This item was documented as an unresolved item in NRC Inspection Report
      05000282/2010005; 05000306/201005. Documents reviewed during this inspection are
      listed in the Attachment to this report.
      This event follow-up review constituted one sample as defined in Inspection Procedure
      71153-05.
  b. Findings
      See Section 4OA5.1 below for a discussion of this issue.
4OA5 Other Activities
.1  (Closed) Unresolved Item 05000282/2010005-05; 05000306/2010005-05: Potential for
      Common Mode Failure of Safety-Related Battery Chargers
  a. Inspection Scope
      The inspectors reviewed the circumstances surrounding the licensees failure to maintain
      the both direct current (DC) electrical power subsystems operable during reactor
      operation in Modes 1 through 4.
  b. Findings
      Introduction: An apparent violation of Technical Specification (TS) 3.8.4,
      DC Sources - Operating, was identified by the inspectors due to the licensees failure
      to maintain both DC electrical power subsystems operable during reactor operation in
      Modes 1 through 4.
      Description: In NRC Inspection Report 05000282/2010005; 05000306/2010005,
      the NRC documented several issues regarding the safety-related battery chargers,
      specifically with the 12 battery charger locking up during a simulated loss of offsite
      power (LOOP) event concurrent with a simulated loss of coolant accident (LOCA).
      In the same inspection report, the NRC opened an unresolved item to address the
      potential for a common mode failure of all of the safety-related battery chargers.
                                                  3                                      Enclosure


=SUMMARY OF FINDINGS=
The inspectors reviewed the licensees evaluation of the potential for a common mode
IR 05000282/2011010; 05000306/2011010; 05/13/21011 - 05/20/2011; Prairie Island Nuclear Generating Plant, Units 1 and 2; Other Activities. This report covers the review of a potential common cause failure of the safety-related battery chargers. The inspectors identified one apparent violation (AV) with a preliminary significance of White for Unit 1 and a preliminary significance of Green for Unit 2. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006. A.
failure. The evaluation contained information that the safety-related battery chargers
had the potential to stop providing an output, or lock-up, if the input alternating current
(AC) voltage dropped below the nameplate minimum voltage of 90 percent of 480 Volts
at the battery charger motor control center (MCC). Specifically, the NRC learned that
the lock-up of the battery charger was related to the operation of the silicon controlled
rectifiers (SCRs) on the battery charger control circuitry. As a low voltage condition
occurred in response to the simulated LOOP/LOCA, the firing angle of the SCRs
advanced to maintain output voltage. If the firing angle advanced too far on a low
voltage condition, the control circuitry became reverse biased and unable produce any
output. The licensee was unable to determine the exact voltage, the duration of voltage
dip, and the battery charger loading conditions which caused the lock-up to occur.
In reviewing plant data from the periodic LOOP/LOCA tests, the inspectors determined
that at certain points in the loading sequence the input voltage to the battery chargers
decreased to less than 90 percent, the design minimum, for all four chargers (two on
Unit 1 and two on Unit 2). In addition, the licensee further determined that the
LOOP/LOCA tests did not include all possible loads, including the 121 motor-driven
cooling water pump, and other loads such as an instrument air compressor.
These loads could further decrease 480V bus voltage and contribute to the battery
charger locking up.
On October 22, 2010, the licensee completed an operability evaluation and concluded
that the chargers could be considered operable but non-conforming if compensatory
measures were put in place. These compensatory measures included revising abnormal
and emergency operating procedures, placing copies of needed abnormal operating
procedures and tools needed within the battery charger rooms, and establishing a
specific designated operator (with no other duties) to perform the manual actions needed
to recover the battery chargers if needed.
At the end of 2010, the licensee completed a past operability review of the
safety-related battery chargers and concluded that there was reasonable doubt that the
chargers would have performed their safety function if called upon during specific
design basis accidents. This was documented in LER 2010-004-00 submitted at the
end of January 2011. As a result, the inspectors concluded that the DC electrical
power subsystems (specifically the safety-related battery chargers) had been inoperable
since their initial installation in December 1994.
In May 2011, the licensee replaced and tested both Unit 1 battery chargers.
The licensee planned to replace the Unit 2 battery chargers during the next refueling
outage. The licensees compensatory actions remain in place for Unit 2.
Analysis: The inspectors determined that the licensees failure to ensure that the
DC electrical power subsystems remained operable during reactor operation in
Modes 1 through 4 was a performance deficiency that required an evaluation using
the Significance Determination Process (SDP) described in NRC Inspection Manual
Chapter (IMC) 0609. The inspectors also determined that this finding should be
assigned to the Mitigating Systems Cornerstone because it impacted systems used in
short term and long term heat removal. The inspectors performed a Phase 1 SDP
analysis and concluded that the finding represented the actual loss of safety function of
                                          4                                         Enclosure


===Cornerstone: Mitigating Systems
a single train for greater than its TS allowed outage time which required a Phase 2 SDP
evaluation.
The Phase 2 SDP result was potentially greater than very low safety significance.
The exposure time was assumed to be 1 year since the battery chargers have been
susceptible to failure since they were installed. For the SDP, the initiating events that
could result in one or more battery charger failures were determined to be those events
where input AC voltage at the battery charger MCC would be less than 90 percent and
charger output demand would be high. These initiating events were any non-station
blackout (SBO) LOOP event and any event that resulted in a safety injection (SI) signal.
These events included small LOCAs, medium LOCAs, large LOCAs, stuck-open power
operated relief valves (PORVs), a steam generator tube rupture (SGTR), and a main
steam line break.
The pre-solved SDP worksheets modeled the Train A or 11 battery charger.
Consistent with the SDP usage rules defined in IMC 0609A, Determining the
Significance of Reactor Inspection Findings - At Power, the pre-solved worksheet
assumed that a finding involving a battery charger would increase the Loss of DC
initiating event frequency. However, this finding only involved failure of the battery
charger in response to specific initiating events and would not increase the Loss of DC
initiating event frequency. Therefore, the worksheets were individually solved assuming
that one train of mitigating equipment would be failed as a result of a battery charger
failure. A recovery credit of 1 was applied in all sequences because recovery of the
battery charger was possible. The dominant sequence was a LOOP, a failure of all
auxiliary feedwater, and the failure of feed and bleed.
A Region III Senior Reactor Analyst (SRA) conducted a Phase 3 SDP evaluation to
provide a more realistic estimate of the change in core damage frequency (CDF) for the
finding. Similar to the Phase 2 SDP evaluation, the exposure time was 1 year and the
initiating events that could result in one or more battery charger failures was assumed to
be either a non-SBO LOOP event or an event that resulted in an SI signal. Loss of
coolant accidents and SGTR were considered to be the events that would result in an
SI signal. The Standardized Plant Analysis Risk (SPAR) model for Prairie Island,
Revision 8.15, was used in the evaluation. The SPAR model represents Prairie Island
Unit 1 only, so the results of the Unit 1 evaluation were generally assumed to be
applicable to Unit 2. The model was modified to (1) require battery charger operation for
DC system success in non-SBO events because the safety-related batteries cannot
function for the entire 24 hour mission time and (2) to account for the potential for
common cause failure (CCF) of the battery chargers. The model was solved assuming
one charger would fail in response to the applicable initiating events and the opposite
train charger had the potential to fail. The dominant cut-sets were reviewed and the
potential for recovery of the failed battery charger was evaluated and applied at the
sequence level.
The inspectors performed a review of the licensees abnormal operating procedures and
determined that a locked-up charger could be recovered by locally turning the charger
off and then turning it back on. However, the operators would be required to diagnose
that the charger had locked-up and to restart the charger prior to the safety-related
battery depleting. The DC battery depletion times for each of the batteries were variable
due to loading differences. In addition, the depletion times were uncertain due to
assumptions regarding operation of equipment in response to initiating events.
                                          5                                      Enclosure


===NRC-Identified===
In general, the battery life estimates for the Unit 1 batteries were shorter than Unit 2
and Self-Revealed Findings===
batteries in non-SBO LOOP events, due to differences in battery loading. Also, for both
Preliminary WhiteThis finding was determined to be more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems Cornerstone. In addition, this performance deficiency impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed a Phase 1 SDP evaluation and determined that a Phase 2 evaluation was required because this finding represented an actual loss of safety function of a single train of equipment for greater than the TS allowed outage time. The inspectors performed a Phase 2 evaluation using the pre-solved SDP worksheets for Prairie Island and determined that this finding screened as Red. A Phase 3 SDP evaluation was required to assess reasonable credit for recovery by operators. The results of the Phase 3 SDP evaluation showed that this finding was preliminarily determined to be White for Unit 1, and Green for Unit 2. No cross-cutting aspect was assigned to this finding because licensee decisions made in regards to evaluating the performance of the battery chargers were made many years ago and therefore, not reflective of current plant performance. (Section 4OA5.1) . An apparent violation of Technical Specification (TS) 3.8.4 was identified by the inspectors due to the licensee's failure to maintain the train A and train B direct current electrical power subsystems operable while operating the reactor in Modes 1 through 4. Specifically, the licensee installed safety-related battery chargers which were susceptible to failure during certain design basis events. This issue was entered into the licensee's corrective action program (CAP) as CAP 1250561. Upon identifying this issue, the licensee performed an operability evaluation and determined that the battery chargers remained operable because procedures were in place to recover the battery chargers if a failure occurred. After further interaction with the NRC, the licensee concluded that a designated operator position needed to be established to ensure that a specific individual would perform the battery charger recovery actions prior to the safety-related batteries being depleted. Long term corrective actions included replacing all four battery chargers.
units, the battery life for non-LOOP events was generally longer because AC power was
available to carry emergency lighting loads.
NUREG/CR-6883, The SPAR-H Human Reliability Analysis Method, was used to
estimate the human error probability (HEP) for the failure to recover a battery charger
prior to safety-related battery depletion. For this HEP, the SRA considered both
diagnosis and action failures but determined that diagnosis was the dominant failure
mode. In response to LOOP events, the only performance shaping factor (PSF) that
was considered a performance driver was stress. During this event operators would be
receiving many alarms in the control room and would have a high workload in
responding to the event. The initial alarm indicating battery charger failure received in
the control room would be DC System Trouble. The alarm response procedure
instructed operators to check the DC Panel Undervoltage alarm and if it was also lit to
proceed to Abnormal Operating Procedure 1C20.9 AOP3 or 1C20.9 AOP4, Failure of
11(12) Battery Charger. The inspectors and the SRA determined that the DC System
Trouble alarm would be expected to come in during LOOP events even if the battery
charger functioned properly. After some period of time, with all equipment functioning
normally, the alarm would clear. If the battery charger failed to function, the alarm would
remain lit. The SRA considered that during an event, operators would likely prioritize
other alarms associated with the initiating event and other potential complications before
attending to the DC System Trouble alarm. Once operators entered the Abnormal
Operating Procedure (AOP), an operator in the plant would record data associated with
the DC system in the battery charger room in order to determine that one or more
battery chargers had locked up. Once diagnosis had occurred, the operator would reset
the charger by opening the input breaker, waiting 10 seconds, and then closing the input
breaker. Stress was considered to be high given that the dominant sequences
involved a complicated LOOP. All other PSFs were considered nominal. The SRA also
concluded that procedures existed for resetting the battery charger(s) and that adequate
time existed for diagnosis and action. Based upon this information, the SRA estimated
an HEP for failure to recover the battery charger as 2.2E-2 for most LOOP sequences
on Unit 1.
For all other events, (LOCAs and SGTRs, Unit 2 LOOP events, and Unit 1 LOOP events
that do not involve failures other than battery charger failures) the SRA also considered
time available to be a performance driver. For these events, the licensees battery
depletion calculations showed that much more time was available for operator to
respond to a locked up battery charger(s) prior to safety-related battery depletion.
Therefore, extra time was considered for diagnosis of the condition. The HEP for
failure to recover a battery charger in these scenarios was estimated to be 2.2E-3.
After reviewing the cut-sets, the SRA determined that for certain LOOP events, recovery
of offsite power within the battery depletion time would also mitigate the event by
allowing power to be restored to the train that was unaffected by a failed battery charger.
To account for this, the SRA applied a factor to the applicable Unit 1 LOOP sequences
that represented the probability that the LOOP event exceeds 2.5 hours. This value was
obtained from NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear
Power Plants, Table 3-3. The value is the composite probability of exceedance for all
categories of LOOP events. For Unit 2 LOOP sequences, a factor that represented the
probability that the LOOP event exceeded 8 hours was used.
                                          6                                        Enclosure


B. No violations of significance were identified.
        The SRA determined that the risk contribution from seismic events and internal flooding
        events for this finding was negligible. For internal fire events, the SRA used the
        licensees Individual Plant Examination of External Events (IPEEE) to consider
        fire-induced LOOP scenarios and concluded that there was a small contribution to the
        risk of the finding from Unit 2 fire scenarios.
        The potential risk contribution to large early release frequency (LERF) was evaluated
        using IMC 0609 Appendix H, Containment Integrity Significance Determination
        Process. Prairie Island is a 2-loop Westinghouse pressurized water reactor with a large
        dry containment. Sequences important to LERF included SGTR events and
        inter-system LOCA events. These were not the dominant core damage sequences for
        this finding and thus the risk significance due to LERF was evaluated to be of very low
        safety significance.
        The total delta CDF for Unit 1 was estimated as 1.9E-6/yr and for Unit 2, 5.2E-7/yr.
        This represented a preliminary White finding for Unit 1 and a preliminary Green finding
        for Unit 2. The dominant core damage sequence cut-set for Unit 1 was a LOOP event
        followed by common cause failure of the battery chargers and the failure to recover a
        battery charger. Other important cut-sets involved a LOOP event, failure of a single
        battery charger, and failure of the opposite train emergency AC power sources. The
        Unit 2 results were dominated by the fire contribution which was not fully developed
        since the initial estimates for total delta CDF, including the fire contribution, were less
        than 1.0E-6/yr.
        The SRA reviewed a risk evaluation performed by the licensee for this finding.
        The licensee concluded that the delta CDF for both units was less than 1.0E-6/yr
        (Green). The major differences between the NRCs SDP evaluation and the licensees
        risk assessment were HEPs estimated for failing to recover a battery charger and the
        assessment of the potential for common cause failure of both chargers on a single unit.
        Both evaluations considered the potential for recovering a charger; however, the
        licensees HEP estimate was more optimistic than the NRCs. The NRC believes the
        potential for operators to miss or misinterpret the alarms during diagnosis of the failed
        charger was higher than estimated by the licensee.
With regard to the common cause failure potential, the licensee assumed that both battery
chargers on a single unit could not lock-up in response to the same initiating event. The NRC
concluded that since all the battery chargers were of the same design and would be modeled as
part of the same common cause component group in a PRA model, that it was appropriate to
treat the potential for common cause failure of both battery charger trains probabilistically,
consistent with the failure memory approach used in NRC risk assessments. In the risk
assessment of inspection findings, the failure memory approach models observed successful
components as having a probability of failure rather than concluding that the component would
always be successful. The results of the SDP evaluation were sensitive to both assumptions on
recovery and common cause failure and therefore, the NRC performed sensitivity evaluations to
vary the best-estimate assumptions. In particular, the NRC considered higher probabilities of
common cause failure due to concerns that the actual potential for common cause failure was
under-represented by SPAR model. The sensitivity evaluations varied the battery charger
common cause failure probability and the human error probabilities used the analysis.
The results of the sensitivity evaluations were generally higher than the best-estimate SDP
evaluation but overall supported the preliminary conclusion of a White finding for Unit 1 and a
Green finding for Unit 2.
                                                    7                                        Enclosure


=== Licensee-Identified Violations===
Old Design Issue Review
Inspection Manual Chapter 0305, Operating Reactor Assessment Program,
Section 12.01, states that the NRC may refrain from considering safety significant
inspection findings in the assessment program for a design-related finding in the
engineering calculations or analysis, associated operating procedure, or installation of
plant equipment if the following statements were true:
*      The issue was licensee-identified as a result of a voluntary initiative such as a
        design basis reconstitution;
*      The performance issue was or will be corrected within a reasonable period of
        time following identification;
*      The issue was not likely to have been previously identified by routine efforts such
        as normal surveillance or quality assurance activities; and
*      The issue does not reflect a current performance deficiency associated with
        existing licensee programs, policy or procedures.
Based upon the information provided above, the inspectors have determined that this
finding did not meet the criteria to be considered an old design issue for the following
reasons:
*      The finding was not licensee-identified as a result of a voluntary initiative.
        Although the licensee initiated a CAP document in late September 2010
        regarding the possibility of charger lock up during grid voltage fluctuations,
        NRC prompting was needed and specifically requested during the October 2010
        exigent TS change discussions to ensure that the licensee addressed the
        susceptibility of all chargers to a lock-up condition during other design basis
        accidents.
*      The failure of the battery chargers to operate as expected following a design
        basis event was first discovered in 1996 during the performance of testing which
        simulated a LOOP/LOCA event. However, the licensee failed to recognize the
        significance of this issue and dispositioned the item as use as is. As a result,
        the issue was not corrected within a reasonable period of time.
*      The finding was likely to be identified by past activities such as surveillance
        testing. Specifically, the licensee was unable to successfully perform the
        simulated LOOP/LOCA test following the 1994 battery charger installation.
        After performing at least two additional LOOP/LOCA tests which resulted in
        the lock-up of the 12 battery charger, the licensee ultimately changed the
        LOOP/LOCA test procedure to ensure that the 12 battery charger was turned
        off prior to performing the surveillance test.
No cross-cutting aspect was assigned to this finding, because licensee decisions made
in regards to evaluating the performance of the battery chargers were made many years
ago and therefore, not reflective of current plant performance.
Enforcement: Technical Specification 3.8.4, DC Sources - Operating, requires that the
train A and train B DC electrical power subsystems be operable in Modes 1 through 4.
                                          8                                        Enclosure


=REPORT DETAILS=
    With one battery charger inoperable, TS 3.8.4, Condition A, requires that the battery
    charger be restored to an operable status in 8 hours or that actions be taken to shut the
    plant down within the following 42 hours.
    With both battery chargers inoperable, Limiting Condition for Operation (LCO) 3.0.3
    requires that when an LCO is not met and the associated actions are not met, an
    associated action is not provided, or if directed by the associated actions, the unit shall
    be placed in a mode or other specified condition in which the LCO is not applicable.
    Action shall be initiated within 1 hour to place the unit, as applicable, in:
    *      Mode 3 within 7 hours;
    *      Mode 4 within 13 hours; and
    *      Mode 5 within 37 hours.
    Contrary to the above, from December 21, 1994, to approximately October 22, 2010,
    the safety-related battery chargers on both Unit 1 and 2 failed to maintain the DC
    electrical power subsystems operable in Modes 1 through 4. Specifically, under design
    basis accident conditions, all battery chargers were susceptible to a common cause
    failure under design basis accident conditions whereby the battery chargers would stop
    providing an output, or lock-up, when their AC input voltage dropped below their
    nameplate minimum voltage at the battery charger MCC. This is an apparent violation of
    TS 3.8.4 pending the completion of the final significance determination
    (AV 05000282/2011010-01; 05000306/2011010-01, Failure to Ensure that the Train A
    and Train B DC Electrical Power Subsystems Remained Operable in Modes 1
    through 4).
4OA6 Management Meetings
.1  Exit Meeting Summary
    On May 20, 2011, the inspectors presented the inspection results to Mr. M. Schimmel,
    and other members of the licensee staff. The licensee acknowledged the issues
    presented. The inspectors confirmed that none of the potential report input discussed
    was considered proprietary.
ATTACHMENT: SUPPLEMENTAL INFORMATION
                                              9                                      Enclosure


==OTHER ACTIVITIES==
                              SUPPLEMENTAL INFORMATION
{{a|4OA3}}
                                  KEY POINTS OF CONTACT
==4OA3 Follow-Up of Events and Notices of Enforcement Discretion.1==
Licensee
{{IP sample|IP=IP 71153}}
M. Schimmel, Site Vice President
a. (Closed) Licensee Event Report 05000282/2010-004:  Battery Charger Inoperability due to Potential Undervoltage Conditions The inspectors reviewed the licensee's response to discovering that safety-related battery chargers installed in 1994 were susceptible to failure during certain design basis accidents. Specifically, these battery chargers had the potential to stop providing an output, or "lock-up," if their alternating current input voltage dropped below the nameplate minimum voltage of 90 percent at the battery charger motor control center (MCC). This item was documented as an unresolved item in NRC Inspection Report 05000282/2010005; 05000306/201005. Documents reviewed during this inspection are listed in the Attachment to this report.
K. Davison, Plant Manager
 
T. Allen, Site Engineering Director - Acting
Inspection Scope This event follow-up review constituted one sample as defined in Inspection Procedure 71153-05. b. See Section 4OA5.1 below for a discussion of this issue. Findings
J. Anderson, Regulatory Affairs Manager
{{a|4OA5}}
C. Bough, Chemistry and Environmental Manager
==4OA5 ==
B. Boyer, Radiation Protection Manager
===.1 Other Activities===
K. DeFusco, Emergency Preparedness Manager
a. (Closed) Unresolved Item 05000282/2010005-05; 05000306/2010005-05:  Potential for Common Mode Failure of Safety-Related Battery Chargers The inspectors reviewed the circumstances surrounding the licensee's failure to maintain the both direct current (DC) electrical power subsystems operable during reactor operation in Modes 1 through 4. Inspection Scope
D. Goble, Safety and Human Performance Manager
 
J. Hamilton, Security Manager
====b. Findings====
J. Lash, Nuclear Oversight Manager
IntroductionDescription:  In NRC Inspection Report 05000282/2010005; 05000306/2010005, the NRC documented several issues regarding the safety-related battery chargers, specifically with the 12 battery charger "locking up" during a simulated loss of offsite power (LOOP) event concurrent with a simulated loss of coolant accident (LOCA). In the same inspection report, the NRC opened an unresolved item to address the potential for a common mode failure of all of the safety-related battery chargers. :  An apparent violation of Technical Specification (TS) 3.8.4, "DC Sources - Operating," was identified by the inspectors due to the licensee's failure to maintain both DC electrical power subsystems operable during reactor operation in Modes 1 through 4.
M. Milly, Maintenance Manager
 
J. Muth, Operations Manager
4 Enclosure The inspectors reviewed the licensee's evaluation of the potential for a common mode failure. The evaluation contained information that the safety-related battery chargers had the potential to stop providing an output, or "lock-up", if the input alternating current (AC) voltage dropped below the nameplate minimum voltage of 90 percent of 480 Volts at the battery charger motor control center (MCC). Specifically, the NRC learned that the lock-up of the battery charger was related to the operation of the silicon controlled rectifiers (SCRs) on the battery charger control circuitry. As a low voltage condition occurred in response to the simulated LOOP/LOCA, the firing angle of the SCRs advanced to maintain output voltage. If the firing angle advanced too far on a low voltage condition, the control circuitry became reverse biased and unable produce any output. The licensee was unable to determine the exact voltage, the duration of voltage dip, and the battery charger loading conditions which caused the lock-up to occur.
S. Northard, Recovery Manager
A. Notbohm, Performance Assessment Supervisor
K. Peterson, Business Support Manager
A. Pullam, Training Manager
R. Womack, Outage Manager
J. Ritter, Risk Analyst
Nuclear Regulatory Commission
J. Giessner, Chief, Reactor Projects Branch 4
T. Wengert, Project Manager, NRR
                    LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000282/2011010-01;       AV      Failure to Ensure that the Train A and Train B DC Electrical
05000306/2011010-01                Power Subsystems Remained Operable in Modes 1 through
                                  4 (Section 4OA5.1)
Closed
05000282/2010-004          LER    Battery Charger Inoperability due to Potential Undervoltage
                                  Conditions
05000282/2010005-05;      URI    Potential for Common Mode Failure of Safety-Related
05000306/2010005-05                Battery Chargers
Discussed
None.
                                                1                                    Attachment


In reviewing plant data from the periodic LOOP/LOCA tests, the inspectors determined that at certain points in the loading sequence the input voltage to the battery chargers decreased to less than 90 percent, the design minimum, for all four chargers (two on Unit 1 and two on Unit 2).
                                  LIST OF DOCUMENTS REVIEWED
The following is a partial list of documents reviewed during the inspection. Inclusion on this list
does not imply that the NRC inspector reviewed the documents in their entirety, but rather that
selected sections or portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
Sections 4OA3 and 4OA5
- Risk Assessment of Operational Events RASP Handbook; Volume 1 (Internal Events) and
  Volume 2 (External Events).
- The Prairie Island Standardized Plant Analysis Risk Model
- NUREG/CR-6890; Reevaluation of Station Blackout Risk at Nuclear Power Plants
- NUREG/CR-6883; The SPAR-H Human Reliability Analysis Method
- INL-EXT-10-18533; SPAR-H Step-by-Step Guidance; Revision 1
- V.SPA.10.013; Battery Depletion Calculation; November 4, 2010
- V.SPA.11.001; Evaluation of Battery Charger Operation for a Loss of Offsite Power (LOOP)
  Event; Revision 0; January 17, 2011
- V.SPA.11.002; Evaluation of Battery Charger Operation for a Safety Injection Event While on
  Offsite Power; February 25, 2011
- V.SPA.11.003; Prairie Island Battery Depletion Study PRA LOOP with Emergency Lighting
  and ISI Steady State Test Loads; Revision 0; February 16, 2011
- V.SPA.11.004; Prairie Island PRA SI Only Battery Depletion Study; Revision 0;
  February 3, 2011
- V.SPA.11.008; Evaluation of Battery Charger Operation During Bus Crosstie Operation;
  Revision 0; March 7, 2011
- V.SPA.11.012; Battery Charger Significance Determination Process Fault Tree Analysis;
  Revision 0; March 23, 2011
- V.SPA.11.013; Battery Charger Significance Determination Process Accident Sequence
  Analysis; Revision 0; March 22, 2011
- V.SPA.11.014; Battery Charger Significance Determination Process Human Reliability
  Analysis; Revision 0; March 22, 2011
- V.SPA.11.015; Battery Charger Significance Determination Process Quantification Analysis;
  Revision 0; March 24, 2011
- V.SPA.11.018; Battery Charger Significance Determination Process Accident Sequence
  Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 29, 2011
- V.SPA.11.019; Battery Charger Significance Determination Process Human Reliability
  Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011
- V.SPA.11.020; Battery Charger Significance Determination Process Quantification Analysis
  (121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011
- Work Order 9712763; 12 Battery Charger Test during SP 1083
- CAP 19971622; Intermittent Operation during SP 1083; December 5, 1997
- CAP 19960452; 12 Battery Charger Intermittent Operation During SP 1083; February 22, 1996
- CAP 1250561; Battery Chargers may stop Operating if Undervoltage Setpoint is Reached;
  September 21, 2010
- CAP 1252265; Questions Related to Operability Review and Reportability for CAP 1238842;
  September 30, 2010
- CAP 1253478; Concerns with the Operability Review from CAP 1238842 on 12 Battery
  Charger; October 9, 2010
- CAP 1254359; Compensatory Measures not Evaluated Properly; October 16, 2010
                                                  2                                   Attachment


In addition, the licensee further determined that the LOOP/LOCA tests did not include all possible loads, including the 121 motor-driven cooling water pump, and other loads such as an instrument air compressor. These loads could further decrease 480V bus voltage and contribute to the battery charger locking up.
- CAP 1238842; CDBI 2010 Prep SP 1083 Revised without Proper 50.59 Screening;
  June 24, 2010
- CAP 1270104; Non-conservative Assumption in Unit 1 Battery Calculations; February 9, 2011
- Operability Review 1238842-01; Continued Operability of D2 Emergency Diesel Generator
  due to Testing Question; October 22, 2010
- Operability Review 1250561-02; Continued Operability of Safety-Related Battery Chargers;
  October 22, 2010
- Alarm Response Procedure C47024; 12 DC System Trouble; Revision 35
- 1C20.9 AOP4; Failure of 12 Battery Charger; Revision 010-A
- 1C20.9 AOP3; Failure of 11 Battery Charger; Revision 9
- 1C20.5 AOP 1; Re-energizing 4.16 KV Bus 15; Revision 12
- 1C20.5 AOP2; Re-energizing 4.16 KV Bus 16; Revision 14
- 1C20.5 AOP4; Reenergizing 4.16 KV Bus 15 Via Bus-Tie Breakers; Revision 3W
- 1C20.5 AOP5; Reenergizing 4.16 KV Bus 16 Via Bus-Tie Breakers; Revision 3W
                                            3                                    Attachment


On October 22, 2010, the licensee completed an operability evaluation and concluded that the chargers could be considered operable but non-conforming if compensatory measures were put in place. These compensatory measures included revising abnormal and emergency operating procedures, placing copies of needed abnormal operating procedures and tools needed within the battery charger rooms, and establishing a specific designated operator (with no other duties) to perform the manual actions needed to recover the battery chargers if needed.
                        LIST OF ACRONYMS USED
AC    Alternating Current
ADAMS Agencywide Document Access Management System
AOP  Abnormal Operating Procedure
AV    Apparent Violation
CAP  Corrective Action Program
CCF  Common Cause Failure
CDF  Core Damage Frequency
CFR  Code of Federal Regulations
DC    Direct Current
DRP  Division of Reactor Projects
HEP  Human Error Probability
IMC  Inspection Manual Chapter
IPEEE Individual Plant Examination of External Events
LCO  Limiting Condition for Operation
LER  Licensee Event Report
LERF  Large Early Release Frequency
LOCA  Loss of Coolant Accident
LOOP  Loss of Off-Site Power
MCC  Motor Control Center
NRC  U.S. Nuclear Regulatory Commission
NRR  Office of Nuclear Reactor Regulation
PARS  Publically Available Records System
PORV  Power Operated Relief Valve
PSF  Performance Shaping Factor
SBO  Station Blackout
SCR  Silicon Controlled Rectifier
SDP  Significance Determination Process
SGTR  Steam Generator Tube Rupture
SI    Safety Injection
SPAR  Standardized Plant Analysis Risk
SRA  Senior Reactor Analyst
TS    Technical Specification
URI  Unresolved Item
                                      4              Attachment


At the end of 2010, the licensee completed a past operability review of the safety-related battery chargers and concluded that there was reasonable doubt that the chargers would have performed their safety function if called upon during specific design basis accidents. This was documented in LER 2010-004-00 submitted at the end of January 2011. As a result, the inspectors concluded that the DC electrical power subsystems (specifically the safety-related battery chargers) had been inoperable since their initial installation in December 1994. In May 2011, the licensee replaced and tested both Unit 1 battery chargers. The licensee planned to replace the Unit 2 battery chargers during the next refueling outage. The licensee's compensatory actions remain in place for Unit 2.
M. Schimmel                                                                -3-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website
at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
                                                                          Sincerely,
                                                                          /RA/
                                                                          Steven West, Director
                                                                          Division of Reactor Projects
Docket Nos.: 50-282; 50-306; 72-010
License Nos.: DPR-42; DPR-60; SNM-2506
Enclosure:                Inspection Report 05000282/2011010; 05000306/2011010
                            w/Attachment: Supplemental Information
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OFFICE              RIII                                RIII                            RIII                              RIII
NAME                JGiessner:dtp                        PLougheeed for                 LKozak                            SWest
                                                          SOrth
DATE                06/06/11                            06/06/11                        06/06/11                          06/09/11
                                                          OFFICIAL RECORD COPY


=====Analysis:=====
Letter to M. Schimmel from S. West dated June 9, 2011
The inspectors determined that the licensee's failure to ensure that the DC electrical power subsystems remained operable during reactor operation in Modes 1 through 4 was a performance deficiency that required an evaluation using the Significance Determination Process (SDP) described in NRC Inspection Manual Chapter (IMC) 0609. The inspectors also determined that this finding should be assigned to the Mitigating Systems Cornerstone because it impacted systems used in short term and long term heat removal. The inspectors performed a Phase 1 SDP analysis and concluded that the finding represented the actual loss of safety function of 5 Enclosure a single train for greater than its TS allowed outage time which required a Phase 2 SDP evaluation.
SUBJECT:       PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2,
 
              NRC INSPECTION REPORT 05000282/2011010; 05000306/2011010
The Phase 2 SDP result was potentially greater than very low safety significance. The exposure time was assumed to be 1 year since the battery chargers have been susceptible to failure since they were installed. For the SDP, the initiating events that could result in one or more battery charger failures were determined to be those events where input AC voltage at the battery charger MCC would be less than 90 percent and charger output demand would be high. These initiating events were any non-station blackout (SBO) LOOP event and any event that resulted in a safety injection (SI) signal. These events included small LOCAs, medium LOCAs, large LOCAs, stuck-open power operated relief valves (PORVs), a steam generator tube rupture (SGTR), and a main steam line break. The pre-solved SDP worksheets modeled the "Train A" or 11 battery charger. Consistent with the SDP usage rules defined in IMC 0609A, "Determining the Significance of Reactor Inspection Findings - At Power," the pre-solved worksheet assumed that a finding involving a battery charger would increase the Loss of DC initiating event frequency. However, this finding only involved failure of the battery charger in response to specific initiating events and would not increase the Loss of DC initiating event frequency. Therefore, the worksheets were individually solved assuming that one train of mitigating equipment would be failed as a result of a battery charger failure. A recovery credit of "1" was applied in all sequences because recovery of the battery charger was possible. The dominant sequence was a LOOP, a failure of all auxiliary feedwater, and the failure of feed and bleed.
              PRELIMINARY WHITE FINDING
 
DISTRIBUTION:
A Region III Senior Reactor Analyst (SRA) conducted a Phase 3 SDP evaluation to provide a more realistic estimate of the change in core damage frequency (CDF) for the finding. Similar to the Phase 2 SDP evaluation, the exposure time was 1 year and the initiating events that could result in one or more battery charger failures was assumed to be either a non-SBO LOOP event or an event that resulted in an SI signal. Loss of coolant accidents and SGTR were considered to be the events that would result in an SI signal. The Standardized Plant Analysis Risk (SPAR) model for Prairie Island, Revision 8.15, was used in the evaluation. The SPAR model represents Prairie Island Unit 1 only, so the results of the Unit 1 evaluation were generally assumed to be applicable to Unit 2. The model was modified to (1) require battery charger operation for DC system success in non-SBO events because the safety-related batteries cannot function for the entire 24 hour mission time and (2) to account for the potential for common cause failure (CCF) of the battery chargers. The model was solved assuming one charger would fail in response to the applicable initiating events and the opposite train charger had the potential to fail. The dominant cut-sets were reviewed and the potential for recovery of the failed battery charger was evaluated and applied at the sequence level.
Daniel Merzke
 
RidsNrrPMPrairieIsland Resource
The inspectors performed a review of the licensee's abnormal operating procedures and determined that a locked-up charger could be recovered by locally turning the charger off and then turning it back on. However, the operators would be required to diagnose that the charger had locked-up and to restart the charger prior to the safety-related battery depleting. The DC battery depletion times for each of the batteries were variable due to loading differences. In addition, the depletion times were uncertain due to assumptions regarding operation of equipment in response to initiating events.
RidsNrrDorlLpl3-1 Resource
 
RidsNrrDirsIrib Resource
6 Enclosure In general, the battery life estimates for the Unit 1 batteries were shorter than Unit 2 batteries in non-SBO LOOP events, due to differences in battery loading. Also, for both units, the battery life for non-LOOP events was generally longer because AC power was available to carry emergency lighting loads. NUREG/CR-6883, "The SPAR-H Human Reliability Analysis Method," was used to estimate the human error probability (HEP) for the failure to recover a battery charger prior to safety-related battery depletion. For this HEP, the SRA considered both diagnosis and action failures but determined that diagnosis was the dominant failure mode. In response to LOOP events, the only performance shaping factor (PSF) that was considered a "performance driver" was stress. During this event operators would be receiving many alarms in the control room and would have a high workload in responding to the event. The initial alarm indicating battery charger failure received in the control room would be "DC System Trouble."  The alarm response procedure instructed operators to check the "DC Panel Undervoltage" alarm and if it was also lit to proceed to Abnormal Operating Procedure 1C20.9 AOP3 or 1C20.9 AOP4, "Failure of 11(12) Battery Charger."  The inspectors and the SRA determined that the "DC System Trouble" alarm would be expected to come in during LOOP events even if the battery charger functioned properly. After some period of time, with all equipment functioning normally, the alarm would clear. If the battery charger failed to function, the alarm would remain lit. The SRA considered that during an event, operators would likely prioritize other alarms associated with the initiating event and other potential complications before attending to the "DC System Trouble" alarm. Once operators entered the Abnormal Operating Procedure (AOP), an operator in the plant would record data associated with the DC system in the battery charger room in order to determine that one or more battery chargers had locked up. Once diagnosis had occurred, the operator would reset the charger by opening the input breaker, waiting 10 seconds, and then closing the input breaker. Stress was considered to be "high" given that the dominant sequences involved a complicated LOOP. All other PSFs were considered nominal. The SRA also concluded that procedures existed for resetting the battery charger(s) and that adequate time existed for diagnosis and action. Based upon this information, the SRA estimated an HEP for failure to recover the battery charger as 2.2E-2 for most LOOP sequences on Unit 1. For all other events, (LOCAs and SGTRs, Unit 2 LOOP events, and Unit 1 LOOP events that do not involve failures other than battery charger failures) the SRA also considered "time available" to be a performance driver. For these events, the licensee's battery depletion calculations showed that much more time was available for operator to respond to a locked up battery charger(s) prior to safety-related battery depletion.
Cynthia Pederson
 
Steven Orth
Therefore, "extra time" was considered for diagnosis of the condition. The HEP for failure to recover a battery charger in these scenarios was estimated to be 2.2E-3. After reviewing the cut-sets, the SRA determined that for certain LOOP events, recovery of offsite power within the battery depletion time would also mitigate the event by allowing power to be restored to the train that was unaffected by a failed battery charger. To account for this, the SRA applied a factor to the applicable Unit 1 LOOP sequences that represented the probability that the LOOP event exceeds 2.5 hours. This value was obtained from NUREG/CR-6890, "Reevaluation of Station Blackout Risk at Nuclear Power Plants," Table 3-3. The value is the composite probability of exceedance for all categories of LOOP events. For Unit 2 LOOP sequences, a factor that represented the probability that the LOOP event exceeded 8 hours was used.
Jared Heck
 
Allan Barker
7 Enclosure The SRA determined that the risk contribution from seismic events and internal flooding events for this finding was negligible. For internal fire events, the SRA used the licensee's Individual Plant Examination of External Events (IPEEE) to consider fire-induced LOOP scenarios and concluded that there was a small contribution to the risk of the finding from Unit 2 fire scenarios. The potential risk contribution to large early release frequency (LERF) was evaluated using IMC 0609 Appendix H, "Containment Integrity Significance Determination Process."  Prairie Island is a 2-loop Westinghouse pressurized water reactor with a large dry containment. Sequences important to LERF included SGTR events and inter-system LOCA events. These were not the dominant core damage sequences for this finding and thus the risk significance due to LERF was evaluated to be of very low safety significance.
Carole Ariano
 
Linda Linn
The total delta CDF for Unit 1 was estimated as 1.9E-6/yr and for Unit 2, 5.2E-7/yr. This represented a preliminary White finding for Unit 1 and a preliminary Green finding for Unit 2. The dominant core damage sequence cut-set for Unit 1 was a LOOP event followed by common cause failure of the battery chargers and the failure to recover a battery charger. Other important cut-sets involved a LOOP event, failure of a single battery charger, and failure of the opposite train emergency AC power sources. The Unit 2 results were dominated by the fire contribution which was not fully developed since the initial estimates for total delta CDF, including the fire contribution, were less than 1.0E-6/yr. The SRA reviewed a risk evaluation performed by the licensee for this finding. The licensee concluded that the delta CDF for both units was less than 1.0E-6/yr (Green). The major differences between the NRC's SDP evaluation and the licensee's risk assessment were HEPs estimated for failing to recover a battery charger and the assessment of the potential for common cause failure of both chargers on a single unit. Both evaluations considered the potential for recovering a charger; however, the licensee's HEP estimate was more optimistic than the NRC's. The NRC believes the potential for operators to miss or misinterpret the alarms during diagnosis of the failed charger was higher than estimated by the licensee. With regard to the common cause failure potential, the licensee assumed that both battery chargers on a single unit could not lock-up in response to the same initiating event. The NRC concluded that since all the battery chargers were of the same design and would be modeled as part of the same common cause component group in a PRA model, that it was appropriate to treat the potential for common cause failure of both battery charger trains probabilistically, consistent with the "failure memory approach" used in NRC risk assessments. In the risk assessment of inspection findings, the "failure memory approach" models observed successful components as having a probability of failure rather than concluding that the component would always be successful. The results of the SDP evaluation were sensitive to both assumptions on recovery and common cause failure and therefore, the NRC performed sensitivity evaluations to vary the "best-estimate" assumptions. In particular, the NRC considered higher probabilities of common cause failure due to concerns that the actual potential for common cause failure was under-represented by SPAR model. The sensitivity evaluations varied the battery charger common cause failure probability and the human error probabilities used the analysis. The results of the sensitivity evaluations were generally higher than the "best-estimate" SDP evaluation but overall supported the preliminary conclusion of a White finding for Unit 1 and a Green finding for Unit 2.
DRPIII
 
DRSIII
8 Enclosure Inspection Manual Chapter 0305, "Operating Reactor Assessment Program," Section 12.01, states that the NRC may refrain from considering safety significant inspection findings in the assessment program for a design-related finding in the engineering calculations or analysis, associated operating procedure, or installation of plant equipment if the following statements were true: Old Design Issue Review  The issue was licensee-identified as a result of a voluntary initiative such as a design basis reconstitution;  The performance issue was or will be corrected within a reasonable period of time following identification;  The issue was not likely to have been previously identified by routine efforts such as normal surveillance or quality assurance activities; and  The issue does not reflect a current performance deficiency associated with existing licensee programs, policy or procedures. Based upon the information provided above, the inspectors have determined that this finding did not meet the criteria to be considered an old design issue for the following reasons:  The finding was not licensee-identified as a result of a voluntary initiative. Although the licensee initiated a CAP document in late September 2010 regarding the possibility of charger lock up during grid voltage fluctuations, NRC prompting was needed and specifically requested during the October 2010 exigent TS change discussions to ensure that the licensee addressed the susceptibility of all chargers to a lock-up condition during other design basis accidents.
Patricia Buckley
 
Tammy Tomczak
The failure of the battery chargers to operate as expected following a design basis event was first discovered in 1996 during the performance of testing which simulated a LOOP/LOCA event. However, the licensee failed to recognize the significance of this issue and dispositioned the item as "use as is."  As a result, the issue was not corrected within a reasonable period of time.
ROPreports Resource
 
The finding was likely to be identified by past activities such as surveillance testing. Specifically, the licensee was unable to successfully perform the simulated LOOP/LOCA test following the 1994 battery charger installation. After performing at least two additional LOOP/LOCA tests which resulted in the lock-up of the 12 battery charger, the licensee ultimately changed the LOOP/LOCA test procedure to ensure that the 12 battery charger was turned off prior to performing the surveillance test.
 
No cross-cutting aspect was assigned to this finding, because licensee decisions made in regards to evaluating the performance of the battery chargers were made many years ago and therefore, not reflective of current plant performance.
 
=====Enforcement:=====
Technical Specification 3.8.4, "DC Sources - Operating," requires that the train A and train B DC electrical power subsystems be operable in Modes 1 through 4.
 
9 Enclosure With one battery charger inoperable, TS 3.8.4, Condition A, requires that the battery charger be restored to an operable status in 8 hours or that actions be taken to shut the plant down within the following 42 hours. With both battery chargers inoperable, Limiting Condition for Operation (LCO) 3.0.3 requires that when an LCO is not met and the associated actions are not met, an associated action is not provided, or if directed by the associated actions, the unit shall be placed in a mode or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:  Mode 3 within 7 hours;  Mode 4 within 13 hours; and  Mode 5 within 37 hours. Contrary to the above, from December 21, 1994, to approximately October 22, 2010, the safety-related battery chargers on both Unit 1 and 2 failed to maintain the DC electrical power subsystems operable in Modes 1 through 4. Specifically, under design basis accident conditions, all battery chargers were susceptible to a common cause failure under design basis accident conditions whereby the battery chargers would stop providing an output, or "lock-up", when their AC input voltage dropped below their nameplate minimum voltage at the battery charger MCC. This is an apparent violation of TS 3.8.4 pending the completion of the final significance determination (AV 05000282/2011010-01; 05000306/2011010-01, Failure to Ensure that the Train A and Train B DC Electrical Power Subsystems Remained Operable in Modes 1 through 4).
{{a|4OA6}}
==4OA6 ==
===.1 Management Meetings On May 20, 2011, the inspectors presented the inspection results to Mr. M. Schimmel, and other members of the licensee staff.===
The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary. Exit Meeting Summary ATTACHMENT: 
 
=SUPPLEMENTAL INFORMATION=
 
==KEY POINTS OF CONTACT==
: [[contact::M. Schimmel]], Site Vice President Licensee
: [[contact::K. Davison]], Plant Manager
: [[contact::T. Allen]], Site Engineering Director - Acting
: [[contact::J. Anderson]], Regulatory Affairs Manager
: [[contact::C. Bough]], Chemistry and Environmental Manager
: [[contact::B. Boyer]], Radiation Protection Manager
: [[contact::K. DeFusco]], Emergency Preparedness Manager
: [[contact::D. Goble]], Safety and Human Performance Manager
: [[contact::J. Hamilton]], Security Manager
: [[contact::J. Lash]], Nuclear Oversight Manager
: [[contact::M. Milly]], Maintenance Manager
: [[contact::J. Muth]], Operations Manager
: [[contact::S. Northard]], Recovery Manager
: [[contact::A. Notbohm]], Performance Assessment Supervisor
: [[contact::K. Peterson]], Business Support Manager
: [[contact::A. Pullam]], Training Manager
: [[contact::R. Womack]], Outage Manager
: [[contact::J. Ritter]], Risk Analyst 
: [[contact::J. Giessner]], Chief, Reactor Projects Branch 4 Nuclear Regulatory Commission
: [[contact::T. Wengert]], Project Manager, NRR 
==LIST OF ITEMS==
 
===OPENED, CLOSED AND DISCUSSED===
: 05000282/2011010-01;
===Opened===
: 05000306/2011010-01 AV Failure to Ensure that the Train A and Train B DC Electrical Power Subsystems Remained Operable in Modes 1 through 4 (Section 4OA5.1) 
: 05000282/2010-004 Closed LER Battery Charger Inoperability due to Potential Undervoltage Conditions
: 05000282/2010005-05;
: 05000306/2010005-05 URI Potential for Common Mode Failure of Safety-Related Battery Chargers 
===Discussed===
None.
Attachment
==LIST OF DOCUMENTS REVIEWED==
The following is a partial list of documents reviewed during the inspection.
: Inclusion on this list does not imply that the NRC inspector reviewed the documents in their entirety, but rather that selected sections or portions of the documents were evaluated as part of the overall inspection effort.
: Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report.  - Risk Assessment of Operational Events RASP Handbook; Volume 1 (Internal Events) and Volume 2 (External Events). Sections 4OA3 and 4OA5 - The Prairie Island Standardized Plant Analysis Risk Model  - NUREG/CR-6890; Reevaluation of Station Blackout Risk at Nuclear Power Plants - NUREG/CR-6883; The
: SPAR-H Human Reliability Analysis Method -
: INL-EXT-10-18533;
: SPAR-H Step-by-Step Guidance; Revision 1 - V.SPA.10.013; Battery Depletion Calculation; November 4, 2010 - V.SPA.11.001; Evaluation of Battery Charger Operation for a Loss of Offsite Power (LOOP) Event; Revision 0; January 17, 2011 - V.SPA.11.002; Evaluation of Battery Charger Operation for a Safety Injection Event While on Offsite Power; February 25, 2011 - V.SPA.11.003; Prairie Island Battery Depletion Study PRA LOOP with Emergency Lighting and ISI Steady State Test Loads; Revision 0; February 16, 2011 - V.SPA.11.004; Prairie Island PRA SI Only Battery Depletion Study; Revision 0;
: February 3, 2011 - V.SPA.11.008; Evaluation of Battery Charger Operation During Bus Crosstie Operation; Revision 0; March 7, 2011 - V.SPA.11.012; Battery Charger Significance Determination Process Fault Tree Analysis";
: Revision 0; March 23, 2011 - V.SPA.11.013; Battery Charger Significance Determination Process Accident Sequence Analysis; Revision 0; March 22, 2011 - V.SPA.11.014; Battery Charger Significance Determination Process Human Reliability Analysis; Revision 0; March 22, 2011 - V.SPA.11.015; Battery Charger Significance Determination Process Quantification Analysis;
: Revision 0; March 24, 2011 - V.SPA.11.018; Battery Charger Significance Determination Process Accident Sequence Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 29, 2011 - V.SPA.11.019; Battery Charger Significance Determination Process Human Reliability Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011 - V.SPA.11.020; Battery Charger Significance Determination Process Quantification Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011 - Work Order
: 9712763; "12 Battery Charger Test during
: SP 1083" -
: CAP 19971622; Intermittent Operation during
: SP 1083; December 5, 1997 -
: CAP 19960452; 12 Battery Charger Intermittent Operation During
: SP 1083; February 22, 1996 -
: CAP 1250561; Battery Chargers may stop Operating if Undervoltage Setpoint is Reached; September 21, 2010 -
: CAP 1252265; Questions Related to Operability Review and Reportability for
: CAP 1238842; September 30, 2010 -
: CAP 1253478; Concerns with the Operability Review from
: CAP 1238842 on 12 Battery Charger; October 9, 2010 -
: CAP 1254359; Compensatory Measures not Evaluated Properly; October 16, 2010 
: Attachment -
: CAP 1238842; CDBI 2010 Prep
: SP 1083 Revised without Proper 50.59 Screening; June 24, 2010 -
: CAP 1270104; Non-conservative Assumption in Unit 1 Battery Calculations; February 9, 2011 - Operability Review
: 1238842-01; Continued Operability of D2 Emergency Diesel Generator due to Testing Question; October 22, 2010 - Operability Review
: 1250561-02; Continued Operability of Safety-Related Battery Chargers; October 22, 2010 - Alarm Response Procedure C47024; 12 DC System Trouble; Revision 35 - 1C20.9 AOP4; Failure of 12 Battery Charger; Revision 010-A - 1C20.9 AOP3; Failure of 11 Battery Charger; Revision 9 - 1C20.5 AOP 1; Re-energizing 4.16 KV Bus 15; Revision 12 - 1C20.5 AOP2; Re-energizing 4.16 KV Bus 16; Revision 14 - 1C20.5 AOP4; Reenergizing 4.16 KV Bus 15 Via Bus-Tie Breakers; Revision 3W - 1C20.5 AOP5; Reenergizing 4.16 KV Bus 16 Via Bus-Tie Breakers; Revision 3W 
: Attachment
==LIST OF ACRONYMS==
: [[USED]] [[]]
: [[AC]] [[Alternating Current]]
: [[ADAMS]] [[Agencywide Document Access Management System]]
: [[AOP]] [[Abnormal Operating Procedure]]
: [[AV]] [[Apparent Violation]]
: [[CAP]] [[Corrective Action Program]]
: [[CCF]] [[Common Cause Failure]]
: [[CDF]] [[Core Damage Frequency]]
: [[CFR]] [[Code of Federal Regulations]]
: [[DC]] [[Direct Current]]
: [[DRP]] [[Division of Reactor Projects]]
: [[HEP]] [[Human Error Probability]]
: [[IMC]] [[Inspection Manual Chapter]]
: [[IPEEE]] [[Individual Plant Examination of External Events]]
: [[LCO]] [[Limiting Condition for Operation]]
: [[LER]] [[Licensee Event Report]]
: [[LERF]] [[Large Early Release Frequency]]
: [[LOCA]] [[Loss of Coolant Accident]]
: [[LOOP]] [[Loss of Off-Site Power]]
: [[MCC]] [[Motor Control Center]]
: [[NRC]] [[U.S. Nuclear Regulatory Commission]]
: [[NRR]] [[Office of Nuclear Reactor Regulation]]
: [[PARS]] [[Publically Available Records System]]
: [[PORV]] [[Power Operated Relief Valve]]
: [[PSF]] [[Performance Shaping Factor]]
: [[SBO]] [[Station Blackout]]
: [[SCR]] [[Silicon Controlled Rectifier]]
: [[SDP]] [[Significance Determination Process]]
: [[SGTR]] [[Steam Generator Tube Rupture]]
: [[SI]] [[Safety Injection]]
: [[SPAR]] [[Standardized Plant Analysis Risk]]
: [[SRA]] [[Senior Reactor Analyst]]
: [[TS]] [[Technical Specification]]
: [[URI]] [[Unresolved Item]]
: [[M.]] [[Schimmel    -3-  In accordance with 10]]
: [[CFR]] [[2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the]]
: [[NRC]] [[Public Document Room or from the Publicly Available Records System (]]
: [[PARS]] [[) component of NRC's document system (ADAMS).]]
: [[ADAMS]] [[is accessible from the]]
NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,  /RA/
Steven West, Director Division of Reactor Projects  Docket Nos.:  50-282; 50-306; 72-010 License Nos.:
: [[DPR]] [[-42; DPR-60; SNM-2506  Enclosure: Inspection Report 05000282/2011010; 05000306/2011010   w/Attachment:  Supplemental Information cc w/encl: Distribution via ListServ              DOCUMENT NAME:  G:\DRPIII\PRAI\Prairie Island 2011 010 Greater than Green Rpt.docx  Publicly Available  Non-Publicly Available  Sensitive  Non-Sensitive To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy]]
: [[OFFICE]] [[]]
: [[RIII]] [[]]
: [[RIII]] [[]]
: [[RIII]] [[]]
: [[RIII]] [[]]
: [[NAME]] [[JGiessner:dtp  PLougheeed for SOrth  LKozak  SWest]]
: [[DATE]] [[06/06/11  06/06/11  06/06/11  06/09/11]]
: [[OFFICI]] [[AL]]
: [[RECORD]] [[]]
: [[COPY]] [[Letter to]]
: [[M.]] [[Schimmel from S. West dated June 9, 2011]]
: [[SUBJEC]] [[T: PRAIRIE]]
: [[ISLAND]] [[]]
: [[NUCLEA]] [[R GENERATING PLANT,]]
: [[UNITS]] [[1]]
: [[AND]] [[2,]]
: [[NRC]] [[]]
: [[INSPEC]] [[TION]]
: [[REPORT]] [[05000282/2011010; 05000306/2011010]]
: [[PRELIM]] [[INARY]]
: [[WHITE]] [[]]
: [[FINDIN]] [[G]]
}}
}}

Latest revision as of 19:45, 12 November 2019

IR 05000282-11-010 & 05000306-11-010; on 05/13/21011 - 05/20/2011; Prairie Island Nuclear Generating Plant, Units 1 and 2, Other Activities
ML111610249
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 06/09/2011
From:
Division Reactor Projects III
To: Schimmel M
Northern States Power Co
References
EA-11-110 IR-11-010
Download: ML111610249 (20)
v • d • e


See also: IR 05000282/2011010

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE ROAD, SUITE 210

LISLE, IL 60532-4352

June 9, 2011

EA-11-110

Mr. Mark A. Schimmel

Site Vice President

Prairie Island Nuclear Generating Plant

Northern States Power Company, Minnesota

1717 Wakonade Drive East

Welch, MN 55089

SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2,

NRC INSPECTION REPORT 05000282/2011010; 05000306/2011010

PRELIMINARY WHITE FINDING

Dear Mr. Schimmel:

On May 20, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at

your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents

the results of this inspection, which were discussed on May 20, 2011, with you and other

members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

This report documents a finding for Unit 1 that has preliminarily been determined to be White or

a finding with low-to-moderate increased safety significance. In addition, this same finding was

preliminarily determined to be Green, a finding of very low safety significance, for Unit 2.

As documented in Section 4OA5 of this report both trains of safety-related battery chargers

were not capable of performing their safety function from initial installation in 1994 due to being

susceptible to failure during certain design basis events. This finding was assessed based on

the best available information, including influential assumptions, using the applicable

Significance Determination Process (SDP).

Upon identification of this issue and after interaction with the NRC, you concluded that a

designated operator position needed to be established to ensure that a specific individual could

perform actions to recover the battery charger(s) prior to the safety-related batteries being

M. Schimmel -2-

depleted. Lastly, during your past operability review you concluded that there was reasonable

doubt that the battery chargers would have performed their safety function if called upon prior to

October 22, 2010, (the date the designated operator position was established). Because of the

compensatory actions taken, no current safety concern exists.

This finding is also an apparent violation of NRC requirements and is being considered

for escalated enforcement action in accordance with the NRC Enforcement Policy.

The current Enforcement Policy can be found at the NRCs Web site at

http://www.nrc.gov/reading-rm/doc-collections/enforcement.

In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our

evaluation using the best available information and issue our final determination of safety

significance within 90 days of the date of this letter. The SDP encourages an open dialogue

between the NRC staff and the licensee; however, the dialogue should not impact the timeliness

of the staffs final determination.

Before the NRC makes its enforcement decision, we are providing you an opportunity to either:

(1) present to the NRC your perspectives on the facts and assumptions used by the NRC to

arrive at the finding and its significance at a Regulatory Conference, or (2) submit your position

on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held

within 30 days of the receipt of this letter and we encourage you to submit supporting

documentation at least one week prior to the conference in an effort to make the conference

more efficient and effective. If a conference is held, it will be open for public observation.

The NRC will also issue a press release to announce the conference. If you decide to submit

only a written response, such submittal should be sent to the NRC within 30 days of the receipt

of this letter. If you decline to request a Regulatory Conference or to submit a written response,

you relinquish your right to appeal the final SDP determination; in that, by not doing either you

fail to meet the appeal requirements stated in the Prerequisite and Limitation Sections of

Attachment 2 of IMC 0609.

Please contact John Giessner at (630) 829-9619 in writing within 10 days of the date of this

letter to notify the NRC of your intended response. If we have not heard from you within

10 days, we will continue with our significance determination and enforcement decision.

The final resolution of this matter will be conveyed in separate correspondence.

Since the NRC has not made a final determination in this matter, no Notice of Violation is

being issued for this inspection finding at this time. Please be advised that the number and

characterization of the apparent violation described in the enclosed inspection report may

change as a result of further NRC review.

M. Schimmel -3-

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its

enclosure, and your response (if any) will be available electronically for public inspection in

the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website

at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Steven West, Director

Division of Reactor Projects

Docket Nos.: 50-282; 50-306;72-010

License Nos.: DPR-42; DPR-60; SNM-2506

Enclosure: Inspection Report 05000282/2011010; 05000306/2011010

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket Nos: 50-282; 50-306;72-010

License Nos: DPR-42; DPR-60; SNM-2506

Report Nos: 05000282/2011010; 05000306/2011010

Licensee: Northern States Power Company, Minnesota

Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2

Location: Welch, MN

Dates: May 13 through 20, 2011

Inspectors: K. Stoedter, Senior Resident Inspector

P. Zurawski, Resident Inspector

C. Brown, Reactor Engineer

L. Kozak, Senior Reactor Analyst

Observer: S. Lynch, Nuclear Safety Professional Development

Program Participant

Approved by: John B. Giessner, Chief

Branch 4

Division of Reactor Projects

Enclosure

TABLE OF CONTENTS

SUMMARY OF FINDINGS ......................................................................................................... 1

4. OTHER ACTIVITIES .................................................................................................... 3

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153) ............. 3

4OA5 Other Activities ............................................................................................... 3

4OA6 Management Meetings ................................................................................... 9

SUPPLEMENTAL INFORMATION............................................................................................. 1

Key Points of Contact ............................................................................................................. 1

List of Items Opened, Closed and Discussed ......................................................................... 1

List of Documents Reviewed .................................................................................................. 1

List of Acronyms Used ............................................................................................................ 4

Enclosure

SUMMARY OF FINDINGS

IR 05000282/2011010; 05000306/2011010; 05/13/21011 - 05/20/2011; Prairie Island Nuclear

Generating Plant, Units 1 and 2; Other Activities.

This report covers the review of a potential common cause failure of the safety-related battery

chargers. The inspectors identified one apparent violation (AV) with a preliminary significance

of White for Unit 1 and a preliminary significance of Green for Unit 2. The significance of most

findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not

apply may be Green or be assigned a severity level after NRC management review. The NRCs

program for overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

identified by the inspectors due to the licensees failure to maintain the train A and

train B direct current electrical power subsystems operable while operating the reactor in

Modes 1 through 4. Specifically, the licensee installed safety-related battery chargers

which were susceptible to failure during certain design basis events. This issue was

entered into the licensees corrective action program (CAP) as CAP 1250561. Upon

identifying this issue, the licensee performed an operability evaluation and determined

that the battery chargers remained operable because procedures were in place to

recover the battery chargers if a failure occurred. After further interaction with the NRC,

the licensee concluded that a designated operator position needed to be established to

ensure that a specific individual would perform the battery charger recovery actions prior

to the safety-related batteries being depleted. Long term corrective actions included

replacing all four battery chargers.

This finding was determined to be more than minor because it was associated with the

design control and equipment performance attributes of the Mitigating Systems

Cornerstone. In addition, this performance deficiency impacted the cornerstone

objective of ensuring the availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences. The inspectors performed a

Phase 1 SDP evaluation and determined that a Phase 2 evaluation was required

because this finding represented an actual loss of safety function of a single train of

equipment for greater than the TS allowed outage time. The inspectors performed a

Phase 2 evaluation using the pre-solved SDP worksheets for Prairie Island and

determined that this finding screened as Red. A Phase 3 SDP evaluation was required

to assess reasonable credit for recovery by operators. The results of the Phase 3 SDP

evaluation showed that this finding was preliminarily determined to be White for Unit 1,

and Green for Unit 2. No cross-cutting aspect was assigned to this finding because

licensee decisions made in regards to evaluating the performance of the battery

chargers were made many years ago and therefore, not reflective of current plant

performance. (Section 4OA5.1)

1 Enclosure

B. Licensee-Identified Violations

No violations of significance were identified.

2 Enclosure

REPORT DETAILS

4. OTHER ACTIVITIES

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153)

.1 (Closed) Licensee Event Report 05000282/2010-004: Battery Charger Inoperability due

to Potential Undervoltage Conditions

a. Inspection Scope

The inspectors reviewed the licensees response to discovering that safety-related

battery chargers installed in 1994 were susceptible to failure during certain design basis

accidents. Specifically, these battery chargers had the potential to stop providing an

output, or lock-up, if their alternating current input voltage dropped below the

nameplate minimum voltage of 90 percent at the battery charger motor control center

(MCC). This item was documented as an unresolved item in NRC Inspection Report 05000282/2010005; 05000306/201005. Documents reviewed during this inspection are

listed in the Attachment to this report.

This event follow-up review constituted one sample as defined in Inspection Procedure

71153-05.

b. Findings

See Section 4OA5.1 below for a discussion of this issue.

4OA5 Other Activities

.1 (Closed) Unresolved Item 05000282/2010005-05; 05000306/2010005-05: Potential for

Common Mode Failure of Safety-Related Battery Chargers

a. Inspection Scope

The inspectors reviewed the circumstances surrounding the licensees failure to maintain

the both direct current (DC) electrical power subsystems operable during reactor

operation in Modes 1 through 4.

b. Findings

Introduction: An apparent violation of Technical Specification (TS) 3.8.4,

DC Sources - Operating, was identified by the inspectors due to the licensees failure

to maintain both DC electrical power subsystems operable during reactor operation in

Modes 1 through 4.

Description: In NRC Inspection Report 05000282/2010005; 05000306/2010005,

the NRC documented several issues regarding the safety-related battery chargers,

specifically with the 12 battery charger locking up during a simulated loss of offsite

power (LOOP) event concurrent with a simulated loss of coolant accident (LOCA).

In the same inspection report, the NRC opened an unresolved item to address the

potential for a common mode failure of all of the safety-related battery chargers.

3 Enclosure

The inspectors reviewed the licensees evaluation of the potential for a common mode

failure. The evaluation contained information that the safety-related battery chargers

had the potential to stop providing an output, or lock-up, if the input alternating current

(AC) voltage dropped below the nameplate minimum voltage of 90 percent of 480 Volts

at the battery charger motor control center (MCC). Specifically, the NRC learned that

the lock-up of the battery charger was related to the operation of the silicon controlled

rectifiers (SCRs) on the battery charger control circuitry. As a low voltage condition

occurred in response to the simulated LOOP/LOCA, the firing angle of the SCRs

advanced to maintain output voltage. If the firing angle advanced too far on a low

voltage condition, the control circuitry became reverse biased and unable produce any

output. The licensee was unable to determine the exact voltage, the duration of voltage

dip, and the battery charger loading conditions which caused the lock-up to occur.

In reviewing plant data from the periodic LOOP/LOCA tests, the inspectors determined

that at certain points in the loading sequence the input voltage to the battery chargers

decreased to less than 90 percent, the design minimum, for all four chargers (two on

Unit 1 and two on Unit 2). In addition, the licensee further determined that the

LOOP/LOCA tests did not include all possible loads, including the 121 motor-driven

cooling water pump, and other loads such as an instrument air compressor.

These loads could further decrease 480V bus voltage and contribute to the battery

charger locking up.

On October 22, 2010, the licensee completed an operability evaluation and concluded

that the chargers could be considered operable but non-conforming if compensatory

measures were put in place. These compensatory measures included revising abnormal

and emergency operating procedures, placing copies of needed abnormal operating

procedures and tools needed within the battery charger rooms, and establishing a

specific designated operator (with no other duties) to perform the manual actions needed

to recover the battery chargers if needed.

At the end of 2010, the licensee completed a past operability review of the

safety-related battery chargers and concluded that there was reasonable doubt that the

chargers would have performed their safety function if called upon during specific

design basis accidents. This was documented in LER 2010-004-00 submitted at the

end of January 2011. As a result, the inspectors concluded that the DC electrical

power subsystems (specifically the safety-related battery chargers) had been inoperable

since their initial installation in December 1994.

In May 2011, the licensee replaced and tested both Unit 1 battery chargers.

The licensee planned to replace the Unit 2 battery chargers during the next refueling

outage. The licensees compensatory actions remain in place for Unit 2.

Analysis: The inspectors determined that the licensees failure to ensure that the

DC electrical power subsystems remained operable during reactor operation in

Modes 1 through 4 was a performance deficiency that required an evaluation using

the Significance Determination Process (SDP) described in NRC Inspection Manual

Chapter (IMC) 0609. The inspectors also determined that this finding should be

assigned to the Mitigating Systems Cornerstone because it impacted systems used in

short term and long term heat removal. The inspectors performed a Phase 1 SDP

analysis and concluded that the finding represented the actual loss of safety function of

4 Enclosure

a single train for greater than its TS allowed outage time which required a Phase 2 SDP

evaluation.

The Phase 2 SDP result was potentially greater than very low safety significance.

The exposure time was assumed to be 1 year since the battery chargers have been

susceptible to failure since they were installed. For the SDP, the initiating events that

could result in one or more battery charger failures were determined to be those events

where input AC voltage at the battery charger MCC would be less than 90 percent and

charger output demand would be high. These initiating events were any non-station

blackout (SBO) LOOP event and any event that resulted in a safety injection (SI) signal.

These events included small LOCAs, medium LOCAs, large LOCAs, stuck-open power

operated relief valves (PORVs), a steam generator tube rupture (SGTR), and a main

steam line break.

The pre-solved SDP worksheets modeled the Train A or 11 battery charger.

Consistent with the SDP usage rules defined in IMC 0609A, Determining the

Significance of Reactor Inspection Findings - At Power, the pre-solved worksheet

assumed that a finding involving a battery charger would increase the Loss of DC

initiating event frequency. However, this finding only involved failure of the battery

charger in response to specific initiating events and would not increase the Loss of DC

initiating event frequency. Therefore, the worksheets were individually solved assuming

that one train of mitigating equipment would be failed as a result of a battery charger

failure. A recovery credit of 1 was applied in all sequences because recovery of the

battery charger was possible. The dominant sequence was a LOOP, a failure of all

auxiliary feedwater, and the failure of feed and bleed.

A Region III Senior Reactor Analyst (SRA) conducted a Phase 3 SDP evaluation to

provide a more realistic estimate of the change in core damage frequency (CDF) for the

finding. Similar to the Phase 2 SDP evaluation, the exposure time was 1 year and the

initiating events that could result in one or more battery charger failures was assumed to

be either a non-SBO LOOP event or an event that resulted in an SI signal. Loss of

coolant accidents and SGTR were considered to be the events that would result in an

SI signal. The Standardized Plant Analysis Risk (SPAR) model for Prairie Island,

Revision 8.15, was used in the evaluation. The SPAR model represents Prairie Island

Unit 1 only, so the results of the Unit 1 evaluation were generally assumed to be

applicable to Unit 2. The model was modified to (1) require battery charger operation for

DC system success in non-SBO events because the safety-related batteries cannot

function for the entire 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time and (2) to account for the potential for

common cause failure (CCF) of the battery chargers. The model was solved assuming

one charger would fail in response to the applicable initiating events and the opposite

train charger had the potential to fail. The dominant cut-sets were reviewed and the

potential for recovery of the failed battery charger was evaluated and applied at the

sequence level.

The inspectors performed a review of the licensees abnormal operating procedures and

determined that a locked-up charger could be recovered by locally turning the charger

off and then turning it back on. However, the operators would be required to diagnose

that the charger had locked-up and to restart the charger prior to the safety-related

battery depleting. The DC battery depletion times for each of the batteries were variable

due to loading differences. In addition, the depletion times were uncertain due to

assumptions regarding operation of equipment in response to initiating events.

5 Enclosure

In general, the battery life estimates for the Unit 1 batteries were shorter than Unit 2

batteries in non-SBO LOOP events, due to differences in battery loading. Also, for both

units, the battery life for non-LOOP events was generally longer because AC power was

available to carry emergency lighting loads.

NUREG/CR-6883, The SPAR-H Human Reliability Analysis Method, was used to

estimate the human error probability (HEP) for the failure to recover a battery charger

prior to safety-related battery depletion. For this HEP, the SRA considered both

diagnosis and action failures but determined that diagnosis was the dominant failure

mode. In response to LOOP events, the only performance shaping factor (PSF) that

was considered a performance driver was stress. During this event operators would be

receiving many alarms in the control room and would have a high workload in

responding to the event. The initial alarm indicating battery charger failure received in

the control room would be DC System Trouble. The alarm response procedure

instructed operators to check the DC Panel Undervoltage alarm and if it was also lit to

proceed to Abnormal Operating Procedure 1C20.9 AOP3 or 1C20.9 AOP4, Failure of

11(12) Battery Charger. The inspectors and the SRA determined that the DC System

Trouble alarm would be expected to come in during LOOP events even if the battery

charger functioned properly. After some period of time, with all equipment functioning

normally, the alarm would clear. If the battery charger failed to function, the alarm would

remain lit. The SRA considered that during an event, operators would likely prioritize

other alarms associated with the initiating event and other potential complications before

attending to the DC System Trouble alarm. Once operators entered the Abnormal

Operating Procedure (AOP), an operator in the plant would record data associated with

the DC system in the battery charger room in order to determine that one or more

battery chargers had locked up. Once diagnosis had occurred, the operator would reset

the charger by opening the input breaker, waiting 10 seconds, and then closing the input

breaker. Stress was considered to be high given that the dominant sequences

involved a complicated LOOP. All other PSFs were considered nominal. The SRA also

concluded that procedures existed for resetting the battery charger(s) and that adequate

time existed for diagnosis and action. Based upon this information, the SRA estimated

an HEP for failure to recover the battery charger as 2.2E-2 for most LOOP sequences

on Unit 1.

For all other events, (LOCAs and SGTRs, Unit 2 LOOP events, and Unit 1 LOOP events

that do not involve failures other than battery charger failures) the SRA also considered

time available to be a performance driver. For these events, the licensees battery

depletion calculations showed that much more time was available for operator to

respond to a locked up battery charger(s) prior to safety-related battery depletion.

Therefore, extra time was considered for diagnosis of the condition. The HEP for

failure to recover a battery charger in these scenarios was estimated to be 2.2E-3.

After reviewing the cut-sets, the SRA determined that for certain LOOP events, recovery

of offsite power within the battery depletion time would also mitigate the event by

allowing power to be restored to the train that was unaffected by a failed battery charger.

To account for this, the SRA applied a factor to the applicable Unit 1 LOOP sequences

that represented the probability that the LOOP event exceeds 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This value was

obtained from NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear

Power Plants, Table 3-3. The value is the composite probability of exceedance for all

categories of LOOP events. For Unit 2 LOOP sequences, a factor that represented the

probability that the LOOP event exceeded 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was used.

6 Enclosure

The SRA determined that the risk contribution from seismic events and internal flooding

events for this finding was negligible. For internal fire events, the SRA used the

licensees Individual Plant Examination of External Events (IPEEE) to consider

fire-induced LOOP scenarios and concluded that there was a small contribution to the

risk of the finding from Unit 2 fire scenarios.

The potential risk contribution to large early release frequency (LERF) was evaluated

using IMC 0609 Appendix H, Containment Integrity Significance Determination

Process. Prairie Island is a 2-loop Westinghouse pressurized water reactor with a large

dry containment. Sequences important to LERF included SGTR events and

inter-system LOCA events. These were not the dominant core damage sequences for

this finding and thus the risk significance due to LERF was evaluated to be of very low

safety significance.

The total delta CDF for Unit 1 was estimated as 1.9E-6/yr and for Unit 2, 5.2E-7/yr.

This represented a preliminary White finding for Unit 1 and a preliminary Green finding

for Unit 2. The dominant core damage sequence cut-set for Unit 1 was a LOOP event

followed by common cause failure of the battery chargers and the failure to recover a

battery charger. Other important cut-sets involved a LOOP event, failure of a single

battery charger, and failure of the opposite train emergency AC power sources. The

Unit 2 results were dominated by the fire contribution which was not fully developed

since the initial estimates for total delta CDF, including the fire contribution, were less

than 1.0E-6/yr.

The SRA reviewed a risk evaluation performed by the licensee for this finding.

The licensee concluded that the delta CDF for both units was less than 1.0E-6/yr

(Green). The major differences between the NRCs SDP evaluation and the licensees

risk assessment were HEPs estimated for failing to recover a battery charger and the

assessment of the potential for common cause failure of both chargers on a single unit.

Both evaluations considered the potential for recovering a charger; however, the

licensees HEP estimate was more optimistic than the NRCs. The NRC believes the

potential for operators to miss or misinterpret the alarms during diagnosis of the failed

charger was higher than estimated by the licensee.

With regard to the common cause failure potential, the licensee assumed that both battery

chargers on a single unit could not lock-up in response to the same initiating event. The NRC

concluded that since all the battery chargers were of the same design and would be modeled as

part of the same common cause component group in a PRA model, that it was appropriate to

treat the potential for common cause failure of both battery charger trains probabilistically,

consistent with the failure memory approach used in NRC risk assessments. In the risk

assessment of inspection findings, the failure memory approach models observed successful

components as having a probability of failure rather than concluding that the component would

always be successful. The results of the SDP evaluation were sensitive to both assumptions on

recovery and common cause failure and therefore, the NRC performed sensitivity evaluations to

vary the best-estimate assumptions. In particular, the NRC considered higher probabilities of

common cause failure due to concerns that the actual potential for common cause failure was

under-represented by SPAR model. The sensitivity evaluations varied the battery charger

common cause failure probability and the human error probabilities used the analysis.

The results of the sensitivity evaluations were generally higher than the best-estimate SDP

evaluation but overall supported the preliminary conclusion of a White finding for Unit 1 and a

Green finding for Unit 2.

7 Enclosure

Old Design Issue Review

Inspection Manual Chapter 0305, Operating Reactor Assessment Program,

Section 12.01, states that the NRC may refrain from considering safety significant

inspection findings in the assessment program for a design-related finding in the

engineering calculations or analysis, associated operating procedure, or installation of

plant equipment if the following statements were true:

  • The issue was licensee-identified as a result of a voluntary initiative such as a

design basis reconstitution;

  • The performance issue was or will be corrected within a reasonable period of

time following identification;

  • The issue was not likely to have been previously identified by routine efforts such

as normal surveillance or quality assurance activities; and

  • The issue does not reflect a current performance deficiency associated with

existing licensee programs, policy or procedures.

Based upon the information provided above, the inspectors have determined that this

finding did not meet the criteria to be considered an old design issue for the following

reasons:

  • The finding was not licensee-identified as a result of a voluntary initiative.

Although the licensee initiated a CAP document in late September 2010

regarding the possibility of charger lock up during grid voltage fluctuations,

NRC prompting was needed and specifically requested during the October 2010

exigent TS change discussions to ensure that the licensee addressed the

susceptibility of all chargers to a lock-up condition during other design basis

accidents.

  • The failure of the battery chargers to operate as expected following a design

basis event was first discovered in 1996 during the performance of testing which

simulated a LOOP/LOCA event. However, the licensee failed to recognize the

significance of this issue and dispositioned the item as use as is. As a result,

the issue was not corrected within a reasonable period of time.

  • The finding was likely to be identified by past activities such as surveillance

testing. Specifically, the licensee was unable to successfully perform the

simulated LOOP/LOCA test following the 1994 battery charger installation.

After performing at least two additional LOOP/LOCA tests which resulted in

the lock-up of the 12 battery charger, the licensee ultimately changed the

LOOP/LOCA test procedure to ensure that the 12 battery charger was turned

off prior to performing the surveillance test.

No cross-cutting aspect was assigned to this finding, because licensee decisions made

in regards to evaluating the performance of the battery chargers were made many years

ago and therefore, not reflective of current plant performance.

Enforcement: Technical Specification 3.8.4, DC Sources - Operating, requires that the

train A and train B DC electrical power subsystems be operable in Modes 1 through 4.

8 Enclosure

With one battery charger inoperable, TS 3.8.4, Condition A, requires that the battery

charger be restored to an operable status in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or that actions be taken to shut the

plant down within the following 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />.

With both battery chargers inoperable, Limiting Condition for Operation (LCO) 3.0.3

requires that when an LCO is not met and the associated actions are not met, an

associated action is not provided, or if directed by the associated actions, the unit shall

be placed in a mode or other specified condition in which the LCO is not applicable.

Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

  • Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
  • Mode 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
  • Mode 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Contrary to the above, from December 21, 1994, to approximately October 22, 2010,

the safety-related battery chargers on both Unit 1 and 2 failed to maintain the DC

electrical power subsystems operable in Modes 1 through 4. Specifically, under design

basis accident conditions, all battery chargers were susceptible to a common cause

failure under design basis accident conditions whereby the battery chargers would stop

providing an output, or lock-up, when their AC input voltage dropped below their

nameplate minimum voltage at the battery charger MCC. This is an apparent violation of

TS 3.8.4 pending the completion of the final significance determination

(AV 05000282/2011010-01; 05000306/2011010-01, Failure to Ensure that the Train A

and Train B DC Electrical Power Subsystems Remained Operable in Modes 1

through 4).

4OA6 Management Meetings

.1 Exit Meeting Summary

On May 20, 2011, the inspectors presented the inspection results to Mr. M. Schimmel,

and other members of the licensee staff. The licensee acknowledged the issues

presented. The inspectors confirmed that none of the potential report input discussed

was considered proprietary.

ATTACHMENT: SUPPLEMENTAL INFORMATION

9 Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Schimmel, Site Vice President

K. Davison, Plant Manager

T. Allen, Site Engineering Director - Acting

J. Anderson, Regulatory Affairs Manager

C. Bough, Chemistry and Environmental Manager

B. Boyer, Radiation Protection Manager

K. DeFusco, Emergency Preparedness Manager

D. Goble, Safety and Human Performance Manager

J. Hamilton, Security Manager

J. Lash, Nuclear Oversight Manager

M. Milly, Maintenance Manager

J. Muth, Operations Manager

S. Northard, Recovery Manager

A. Notbohm, Performance Assessment Supervisor

K. Peterson, Business Support Manager

A. Pullam, Training Manager

R. Womack, Outage Manager

J. Ritter, Risk Analyst

Nuclear Regulatory Commission

J. Giessner, Chief, Reactor Projects Branch 4

T. Wengert, Project Manager, NRR

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened

05000282/2011010-01; AV Failure to Ensure that the Train A and Train B DC Electrical

05000306/2011010-01 Power Subsystems Remained Operable in Modes 1 through

4 (Section 4OA5.1)

Closed

05000282/2010-004 LER Battery Charger Inoperability due to Potential Undervoltage

Conditions05000282/2010005-05; URI Potential for Common Mode Failure of Safety-Related

05000306/2010005-05 Battery Chargers

Discussed

None.

1 Attachment

LIST OF DOCUMENTS REVIEWED

The following is a partial list of documents reviewed during the inspection. Inclusion on this list

does not imply that the NRC inspector reviewed the documents in their entirety, but rather that

selected sections or portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

Sections 4OA3 and 4OA5

- Risk Assessment of Operational Events RASP Handbook; Volume 1 (Internal Events) and

Volume 2 (External Events).

- The Prairie Island Standardized Plant Analysis Risk Model

- NUREG/CR-6890; Reevaluation of Station Blackout Risk at Nuclear Power Plants

- NUREG/CR-6883; The SPAR-H Human Reliability Analysis Method

- INL-EXT-10-18533; SPAR-H Step-by-Step Guidance; Revision 1

- V.SPA.10.013; Battery Depletion Calculation; November 4, 2010

- V.SPA.11.001; Evaluation of Battery Charger Operation for a Loss of Offsite Power (LOOP)

Event; Revision 0; January 17, 2011

- V.SPA.11.002; Evaluation of Battery Charger Operation for a Safety Injection Event While on

Offsite Power; February 25, 2011

- V.SPA.11.003; Prairie Island Battery Depletion Study PRA LOOP with Emergency Lighting

and ISI Steady State Test Loads; Revision 0; February 16, 2011

- V.SPA.11.004; Prairie Island PRA SI Only Battery Depletion Study; Revision 0;

February 3, 2011

- V.SPA.11.008; Evaluation of Battery Charger Operation During Bus Crosstie Operation;

Revision 0; March 7, 2011

- V.SPA.11.012; Battery Charger Significance Determination Process Fault Tree Analysis;

Revision 0; March 23, 2011

- V.SPA.11.013; Battery Charger Significance Determination Process Accident Sequence

Analysis; Revision 0; March 22, 2011

- V.SPA.11.014; Battery Charger Significance Determination Process Human Reliability

Analysis; Revision 0; March 22, 2011

- V.SPA.11.015; Battery Charger Significance Determination Process Quantification Analysis;

Revision 0; March 24, 2011

- V.SPA.11.018; Battery Charger Significance Determination Process Accident Sequence

Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 29, 2011

- V.SPA.11.019; Battery Charger Significance Determination Process Human Reliability

Analysis (121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011

- V.SPA.11.020; Battery Charger Significance Determination Process Quantification Analysis

(121 Cooling Water Pump Sensitivity); Revision 0; March 31, 2011

- Work Order 9712763; 12 Battery Charger Test during SP 1083

- CAP 19971622; Intermittent Operation during SP 1083; December 5, 1997

- CAP 19960452; 12 Battery Charger Intermittent Operation During SP 1083; February 22, 1996

- CAP 1250561; Battery Chargers may stop Operating if Undervoltage Setpoint is Reached;

September 21, 2010

- CAP 1252265; Questions Related to Operability Review and Reportability for CAP 1238842;

September 30, 2010

- CAP 1253478; Concerns with the Operability Review from CAP 1238842 on 12 Battery

Charger; October 9, 2010

- CAP 1254359; Compensatory Measures not Evaluated Properly; October 16, 2010

2 Attachment

- CAP 1238842; CDBI 2010 Prep SP 1083 Revised without Proper 50.59 Screening;

June 24, 2010

- CAP 1270104; Non-conservative Assumption in Unit 1 Battery Calculations; February 9, 2011

- Operability Review 1238842-01; Continued Operability of D2 Emergency Diesel Generator

due to Testing Question; October 22, 2010

- Operability Review 1250561-02; Continued Operability of Safety-Related Battery Chargers;

October 22, 2010

- Alarm Response Procedure C47024; 12 DC System Trouble; Revision 35

- 1C20.9 AOP4; Failure of 12 Battery Charger; Revision 010-A

- 1C20.9 AOP3; Failure of 11 Battery Charger; Revision 9

- 1C20.5 AOP 1; Re-energizing 4.16 KV Bus 15; Revision 12

- 1C20.5 AOP2; Re-energizing 4.16 KV Bus 16; Revision 14

- 1C20.5 AOP4; Reenergizing 4.16 KV Bus 15 Via Bus-Tie Breakers; Revision 3W

- 1C20.5 AOP5; Reenergizing 4.16 KV Bus 16 Via Bus-Tie Breakers; Revision 3W

3 Attachment

LIST OF ACRONYMS USED

AC Alternating Current

ADAMS Agencywide Document Access Management System

AOP Abnormal Operating Procedure

AV Apparent Violation

CAP Corrective Action Program

CCF Common Cause Failure

CDF Core Damage Frequency

CFR Code of Federal Regulations

DC Direct Current

DRP Division of Reactor Projects

HEP Human Error Probability

IMC Inspection Manual Chapter

IPEEE Individual Plant Examination of External Events

LCO Limiting Condition for Operation

LER Licensee Event Report

LERF Large Early Release Frequency

LOCA Loss of Coolant Accident

LOOP Loss of Off-Site Power

MCC Motor Control Center

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

PARS Publically Available Records System

PORV Power Operated Relief Valve

PSF Performance Shaping Factor

SBO Station Blackout

SCR Silicon Controlled Rectifier

SDP Significance Determination Process

SGTR Steam Generator Tube Rupture

SI Safety Injection

SPAR Standardized Plant Analysis Risk

SRA Senior Reactor Analyst

TS Technical Specification

URI Unresolved Item

4 Attachment

M. Schimmel -3-

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its

enclosure, and your response (if any) will be available electronically for public inspection in

the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website

at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Steven West, Director

Division of Reactor Projects

Docket Nos.: 50-282; 50-306;72-010

License Nos.: DPR-42; DPR-60; SNM-2506

Enclosure: Inspection Report 05000282/2011010; 05000306/2011010

w/Attachment: Supplemental Information

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OFFICE RIII RIII RIII RIII

NAME JGiessner:dtp PLougheeed for LKozak SWest

SOrth

DATE 06/06/11 06/06/11 06/06/11 06/09/11

OFFICIAL RECORD COPY

Letter to M. Schimmel from S. West dated June 9, 2011

SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2,

NRC INSPECTION REPORT 05000282/2011010; 05000306/2011010

PRELIMINARY WHITE FINDING

DISTRIBUTION:

Daniel Merzke

RidsNrrPMPrairieIsland Resource

RidsNrrDorlLpl3-1 Resource

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Steven Orth

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