ML111610249
| ML111610249 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| 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) | |
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
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
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
- 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
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
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 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
DISTRIBUTION:
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