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
E A-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 REPOR
T 05000 282/2011010; 050003 06/2 011010 PRELIMINARY WHITE FINDING Dear Mr. Schimmel:
On May 20, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a
n 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 b
e 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 design ated 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 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 positi
on 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.
Inspection Report 05000282
/2 011010; 050003 06/2 011010 w/Attachment: 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 No s: 05000282/20 11010; 050 003 06/20 11010 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
1 Enclosure SUMMARY OF FINDINGS
IR 05000 282/2011010; 050003 06/2011010; 05/13/21011
- 05/20/2011; Prairie Island Nuclear Generating Plan t, 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. Cornerstone: Mitigating Systems
N RC-Identified and Self-Revealed Findings 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 th
e 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 man
y 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 design
ated 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.
2 Enclosure B. No violations of significance were identified.
Licensee-Identified Violations
3 Enclosure REPORT DETAILS
4. OTHER ACTIVITIES
4OA3 Follow-U p of Events and Notices of Enforcement Discretion
.1 (71153) 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. 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 4OA5 .1 Other Activities
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
b. Findings 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.
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.
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 licensee's compensatory actions remain in place for Unit 2.
Analysis: 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
th e 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
. 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 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 ha
d 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.
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 instruct ed 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 ha
d 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 involve d a complicated LOOP. All other PSFs were considered nominal. The SRA also concluded that procedures exist
ed for resetting the battery charger(s)
and that adequate time exist
ed 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 SGTR
s, 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. 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 represent
ed 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 exceed
ed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was used.
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 include
d 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 represent ed 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.
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.
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.
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 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 .1 Management Meetings
O n May 20, 2011, the inspector
s 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
1 Attachment
SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC
T M. Schimmel, Site Vice President
Licensee 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
J. Giessner, Chief, Reactor Projects Branch 4
Nuclear Regulatory Commission
T. Wengert, Project Manager, NRR
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
05000282/201
1 010-01; Opened 05000306/2011
010-01 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)
-004 Closed LER Battery Charger Inoperability due to Potential Undervoltage Conditions
05000282/20
1 0005-05; 05000306/20
1 0005-05 URI Potential for Common Mode Failure of Safety
-Related Battery Chargers
Discuss ed None.
2 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).
Section s 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
3 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
4 Attachment
LIST OF ACRONYMS USE
D 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
SC R 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 (ADAM S). 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 Project
s Docket Nos.:
50-282; 50-306;72-010 License Nos.: DPR-42; DPR-60; SNM-2506 Enclosure:
Inspection Report 05000282
/2 011010; 050003 06/2 011010 w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
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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 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
05000 282/2011010; 050003 06/2 011010 PRELIMINARY WHITE FINDING DISTRIBUTION
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