ML062510453
| ML062510453 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/08/2006 |
| From: | Christopher Cahill, Cheung L, Marlone Davis, Jack Giessner Division Reactor Projects I, Division of Reactor Safety I |
| To: | |
| References | |
| IR-06-012 | |
| Download: ML062510453 (19) | |
See also: IR 05000317/2006012
Text
September 8, 2006
Mr. James A. Spina, Vice President
Calvert Cliffs Nuclear Power Plant, Inc.
Constellation Generation Group, LLC
1650 Calvert Cliffs Parkway
Lusby, Maryland 20657-4702
SUBJECT:
CALVERT CLIFFS NUCLEAR POWER PLANT - NRC INSPECTION REPORT
NOS. 05000317/2006012 AND 05000318/2006012; PRELIMINARY WHITE
FINDING
Dear Mr. Spina:
On August 16, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Calvert Cliffs Nuclear Power Plant. The results of this inspection were discussed on
August 16, 2006, 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. Specifically, this inspection focused on the 1A Emergency Diesel Generator (EDG)
at Calvert Cliffs Unit 1 and the activities performed by your staff in response to the unexpected
trip of an associated circuit breaker during a spring 2006 refueling outage surveillance test.
This report documents one finding related to the 1A EDG that appears to have low to
moderate safety significance. As described in Section 2 of this report, a circuit breaker which
normally supplies the EDG's support systems was found to have an incorrect trip setpoint. The
low over-current trip setpoint would have impacted the capability of the 1A EDG to perform its
intended safety function during certain design basis events. The self-revealing finding involved
inadequate design control during the establishment of the breaker's over-current trip setpoint.
While this issue did present a potential safety concern, the plant was shutdown at the time of
discovery and the requirements for onsite power systems were met. Actions were taken to
establish the proper circuit breaker settings prior to the end of the refueling outage and no
current safety concern exists.
This finding was assessed using the reactor safety Significance Determination Process (SDP)
and was preliminarily determined to be White for Unit 1 (i.e., a finding with some increased
importance to safety, which may require additional NRC inspection). The finding appears to
have low to moderate safety significance because the 1A EDG would not have been capable
of performing its intended safety function under all conditions for which it was designed.
This finding is an apparent violation of NRC requirements specified in 10 CFR 50, Appendix B,
Criterion III, "Design Control," and is being considered for escalated enforcement action in
accordance with the NRC Enforcement Policy. The current policy is included on the NRC's
website at http://www.nrc.gov; select What We Do, Enforcement, then Enforcement Policy.
We believe that we have sufficient information to make our final risk determination for the
performance issue involving the incorrect circuit breaker trip setpoints. However, before the
NRC makes a final decision on this matter, we are providing you an opportunity to: (1) present
to the NRC your perspective 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 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 Regulatory Conference is held, it will be open for public observation and a press
release will be issued to announce it. If you decide to provide a written response in lieu of the
Regulatory Conference, the submission should be sent to the NRC within 30 days of the receipt
of this letter.
Please contact Mr. Brian McDermott at (610) 337-5233 within 10 business days of the date of
this letter to notify the NRC of your intentions. If we have not heard from you within 10 days, we
will continue with our significance determination and enforcement decision, and you will be
advised by separate correspondence of the results of our deliberations on this matter. Since
the NRC has not made a final determination in this matter, no Notice of Violation is being issued
for the inspection finding at this time. In addition, please be advised that the characterization of
the apparent violation described in this letter 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
and its enclosure will be available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARS) component of NRC's document
system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Brian E. Holian, Director
Division of Reactor Projects
Docket No. 50-317, 50-318
License No. DPR-53, DRP-69
Enclosure:
Inspection Report 05000317/2006012 and 05000318/2006012
w/Attachments: A) Supplemental Information, B) Motor Control Center Loads
cc w/encl:
M. J. Wallace, President, Constellation Generation
J. M. Heffley, Senior Vice President and Chief Nuclear Officer
President, Calvert County Board of Commissioners
C. W. Fleming, Senior Counsel, Constellation Generation Group, LLC
Director, Nuclear Regulatory Matters
R. McLean, Manager, Nuclear Programs
K. Burger, Esquire, Maryland People's Counsel
State of Maryland (2)
Criterion III, "Design Control," and is being considered for escalated enforcement action in accordance with the NRC Enforcement
Policy. The current policy is included on the NRC's website at http://www.nrc.gov; select What We Do, Enforcement, then
We believe that we have sufficient information to make our final risk determination for the performance issue involving the incorrect
circuit breaker trip setpoints. However, before the NRC makes a final decision on this matter, we are providing you an opportunity
to: (1) present to the NRC your perspective 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 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 Regulatory Conference
is held, it will be open for public observation and a press release will be issued to announce it. If you decide to provide a written
response in lieu of the Regulatory Conference, the submission should be sent to the NRC within 30 days of the receipt of this letter.
Please contact Mr. Brian McDermott at (610) 337-5233 within 10 business days of the date of this letter to notify the NRC of your
intentions. If we have not heard from you within 10 days, we will continue with our significance determination and enforcement
decision, and you will be advised by separate correspondence of the results of our deliberations on this matter. Since the NRC has
not made a final determination in this matter, no Notice of Violation is being issued for the inspection finding at this time. In addition,
please be advised that the characterization of the apparent violation described in this letter 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 and its enclosure will be available
electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Brian E. Holian, Director
Division of Reactor Projects
Docket No. 50-317, 50-318
License No. DPR-53, DRP-69
Enclosure:
Inspection Report 05000317/2006012 and 05000318/2006012
w/Attachments: A) Supplemental Information, B) Motor Control Center Loads
Distribution w/encl: (via E-mail)
S. Collins, RA
B. Holian, DRP
M. Dapas, DRA
B. Sosa, RI OEDO
R. Laufer, NRR
B. McDermott, DRP
A. Burritt, DRP
J. Giessner, DRP, Senior Resident Inspector (acting)
M. Davis, DRS
C. Cahill, DRS
D. Holody, ORA
L. Cheng, DRS
C. Newgent - Resident OA
Region I Docket Room (with concurrences)
ROPReports@nrc.gov
SUNSI Review Complete: ALB (Reviewers Initials)
DOCUMENT NAME:C:\\MyFiles\\Copies\\CC IR2006-012.wpd
After declaring this document An Official Agency Record it will/will not be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with
attachment/enclosure "N" = No copy
OFFICE RI/DRP
RI/DRP
RI/DRS
RI/ORA
RI/DRP
NAME
JGiessner/ALB
FOR
BMcDermott
CCahill
DHolody/RJS
FOR
BHolian
DATE
09/6/06
09/6/06
09/7/06
09/8/06
09/8/06
OFFICIAL RECORD COPY
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos.
50-317, 50-318
License Nos.
Report Nos.
05000317/2006012 and 05000318/2006012
Licensee:
Constellation Generation Group, LLC
Facility:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Location:
Lusby, MD
Dates:
July 10, 2006 through August 16, 2006
Inspectors:
John B. Giessner, Senior Resident Inspector (acting, DRP)
Christopher G. Cahill, Senior Reactor Analyst (DRS)
Leonard S. Chueng, Senior Inspector (DRS)
Marlone Davis, Resident Inspector (DRP)
Approved by:
Brian J. McDermott, Chief
Projects Branch 1
Division of Reactor Projects
Enclosure
ii
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.0
Description of Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2.0
Equipment Failures and Causes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
4OA6 Meetings, Including Exits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
ATTACHMENTS: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3
Enclosure
iii
SUMMARY OF FINDINGS
IR 05000317/2006012 and 05000318/2006012; 07/10/2006 - 08/16/2006; Calvert Cliffs Nuclear
Power Plant, Unit 1; Event Followup.
The report documents an event follow-up inspection focused on the 1A Emergency Diesel
Generator (EDG) at Calvert Cliffs Unit 1 and the followup activities performed by your staff in
response to the trip of an associated circuit breaker during a spring 2006 refueling outage
surveillance test. The inspection was conducted by regional and resident inspectors. One
apparent violation (AV) with potential low to moderate safety significance was identified
(Preliminary White). 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 3, dated July 2000.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Preliminary White: An apparent violation of 10 CFR 50, Appendix B, Criterion III
(Design Control) was identified involving the failure to ensure an adequate trip setpoint
for the electrical circuit breaker that supplies the 1A EDG support systems. An SDP
Phase 3 risk analysis determined that the failure to account for possible combinations of
1A EDG support equipment operation in the short-time over-current trip setpoint for the
supply breaker to 1MCC123 was preliminarily of low to moderate safety significance.
Specifically, the short-time over-current trip setpoint was set too low and it did not
account for the in-rush current associated with the possible combinations of equipment
that could start and operate to support the 1A EDG following a loss of offsite power
(LOOP). This low setpoint, combined with normal setpoint drift, resulted in substantial
periods where the 1A EDG would not have been able to perform its safety function,
because the support system supply circuit breaker would have tripped open
inappropriately. Calvert Cliffs took immediate action to correct the breaker setpoint and
evaluate other potential deficiencies of a similar nature. This issue was entered into the
corrective action program at Calvert Cliffs for resolution.
The finding was more than minor because it affected the Mitigating Systems
Cornerstone objective to ensure the availability and reliability of systems
(i.e., emergency AC power) that respond to initiating events to prevent undesirable
consequences, and its related attribute for design control. The "0C" Station Blackout
Diesel Generator (a non-safety related, but risk-important power source) and the breaker
for its support systems were similarly affected by the performance deficiency. SDP
Phase 1, Phase 2, and Phase 3 assessments were used to evaluate the risk
significance of this finding. The Phase 1 screening required performance of a Phase 2
evaluation because the finding represented a loss of safety function of a single train, for
greater than its allowed outage time. The Technical Specification (TS) allowed outage
time is 14 days for a single EDG. To assess the full significance both the Phase 2 and
Phase 3 analyzes assumed a 5407 hour exposure for the 1A EDG being unable to
Enclosure
iv
perform its safety function and an additional 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> where both the 1A EDG and the
0C DG would not have been able to perform their required functions (the 0C EDG
had less instrument drift). The Region I senior reactor analyst (SRA) conducted a Phase
3 Risk Assessment, to refine the Phase 2 analysis and to incorporate external events
and recovery credit. The Phase 3 analysis for internal and external initiating events,
using the above assumptions and licensee risk information, determined a )CDF of
approximately 1 in 150,000 years of operation (mid E-6 per year range) for both internal
and external events, with no associated increase in large early release frequency
(LERF). The risk of the 1A EDG exposure time dominated the analysis by several
orders of magnitude over the risk of the concurrent 1A EDG and 0C DG exposure
time. A large fire in the turbine building, which causes a loss of offsite power, was the
dominating initiating event.
Enclosure
REPORT DETAILS
1.0
Description of Events
On March 24, 2006, Operations personnel secured the 1A Emergency Diesel
Generator (1A EDG) at Calvert Cliffs Unit 1 during a refueling outage surveillance test,
based on the unexpected trip of an associated circuit breaker. Surveillance test
procedure (STP) O-004A-1, Unit 1 A-Train ESF Test, simulates a loss of the 4 kV
safety bus and initiates a fast start of the EDG. During the performance of the STP,
plant operators initially observed that the 1A EDG started and picked up the 4 kV bus
loads as expected. However, Operations subsequently noted that the 480 V feeder
breaker for Motor Control Center (MCC) 123 was tripped. This MCC supplies essential
EDG support systems such as radiator cooling fans, room ventilation cooling fans, and
other loads. Approximately twenty minutes after the EDG start, Operations aborted the
STP and secured the 1A EDG to prevent the engine from overheating due to the loss
of the cooling fans and room ventilation.
During the evaluation of this event, Calvert Cliffs determined that the 480 V feeder
breaker for MCC 123 tripped when the MCC was initially energized during loading of the
EDG. The feeder breaker tripped because the MCC 123 loads that immediately
energized exceeded the breaker's short-time over-current trip setpoint. The licensee
subsequently determined that the over-current trip setpoint was set too low due to a
design error that occurred when the EDG was installed in 1996.
Background
The safety related 1A EDG and the non-safety related 0C Station Blackout Diesel
Generator at Calvert Cliffs are French designed five megawatt dual diesel generators
(i.e., one diesel at each end with the generator in the middle). Each EDG is housed in
its own building, with a 125 V battery room and battery chargers. The 1A EDG is
cooled by six radiator fans and the building is cooled by a series of four ventilation fans.
The room ventilation fans start based on the diesel generator room temperature. One
building ventilation fan starts when the EDG starts, the other three fans start sequentially
based on the diesel generator room temperature. The ventilation system also includes
various heaters for temperature control during the winter months.
The 0C EDG (station blackout diesel) is a non-safety related, manual-start EDG that is
capable of being aligned to any one of the safety-related 4kV busses for either Unit 1 or
Unit 2 at Calvert Cliffs. The 0C EDG has a heating, ventilating and conditioning
(HVAC) design similar to the 1A EDG. The 0C EDG is important to plant risk and its
support system supply breaker setting was impacted, similar to the 1A EDG. As part of
the corrective actions for the issue discussed in this report, the licensee determined that
the Amptector setting for 0C EDG MCC supply breaker was also set too low. The
NRC's significance determination for the finding discussed in this report takes into
consideration the impact of the incorrect breaker settings for the 1A EDG and 0C
EDG.
2
Enclosure
2.0
Equipment Failures and Causes
Incorrect Setting of 1MCC123 Supply Breaker Over-current Trip Device
a.
Inspection Scope
The inspectors, as part of event follow-up, reviewed the events surrounding the breaker
trip and the existing design of the 1A EDG breakers and trip scheme. This review
included the related Updated Final Safety Analysis Report (UFSAR) sections and TSs,
diagrams, various design electrical analyses, discussion with site personnel, summaries
of testing results and various condition reports (CRs) related to the issue. The
inspectors reviewed the design modification which installed the 1A EDG and the testing
performed. In addition, the team reviewed the calculations and evaluations done to
assess the impact (probabilistic risk assessment) and past operability related to the
breaker tripping. This included a review of calculations, cause analysis, operator actions
and the overall thoroughness of the extent of condition. The extent of condition review
included the 0C EDG, the diesel generator with similar design. The inspectors also
reviewed Constellations immediate corrective action to address the issue, which
included the changes to the over-current setting to verify they were adequately set to
prevent inadvertent tripping, and to verify adequate breaker and bus protection. The
inspection included a review of CRs to ensure the scope of the licencees review was
adequate.
b.
Findings
Introduction: An apparent violation of 10 CFR 50, Appendix B, Criterion III (Design
Control) involving the licensee's failure to ensure an adequate trip setpoint for the
electrical circuit breaker that supplies the 1A EDG support systems was identified. A
SDP Phase 3 risk analysis determined that the failure to account for possible
combinations of 1A EDG support equipment operation in the short-time over-current
trip setpoint for the supply breaker to 1MCC123 was preliminarily of low to moderate
safety significance.
The inspectors verified that the condition of the short-time over-current trip setpoint
would not have been identifiable during normal monthly EDG surveillance testing.
When the 1A EDG is started for surveillance testing one room cooling fan and other
loads are already normally running and the other fans would start based on room
temperature alone (they would not simultaneously start as they do following a
loss-of-offsite power).
Description: The 1A EDG was manually secured during testing when it was
determined that the circuit breaker that feeds 1MCC123 had tripped. Since the EDG
would not be able to perform its design functions with its support loads de-energized, the
licensee promptly declared the EDG inoperable and started an investigation. The cause
for the loss of 1MCC123 was due to the feeder breaker short-time over-current
exceeding its trip setpoint, resulting in a trip of the feeder breaker. Attachment B of this
report has a detailed description of the MCC loads and their design ratings.
The original short-time over-current of the Amptector (trip device) for the supply breaker
was set at 2400 amps. The setpoint was based on providing coordination with the
3
Enclosure
upstream breaker; and the starting of the single largest motor, and the MCC supplying
all other loads, as documented in Calculation D-E-94-001, dated August 26,1994. The
combined current from the assumed loading was increased by a factor of two to account
for the direct-current (DC) off-set current and the large uncertainty of the breaker tripping
setpoint (i.e., including calibration tolerance, setpoint drift and the inherent inaccuracy of
the Amptector). However, the setpoint did not account for all the loads that can
simultaneously start after an undervoltage event. The setpoint was inadequate because
once the 1A EDG started and 1MCC123 energized, more than one motor can start
instantly. At the design room temperature of 105°F, all four room ventilation fans would
start simultaneously. Fifteen motors and other breaker loads could start simultaneously,
resulting in a combined in-rush current of more than 2600 amps. In addition, with the
tripping setpoint uncertainty, setpoint drift of up to 10%, and DC off-set current, the in-
rush current could be much higher than the original setpoint. During an extent of
condition review, Constellation determined that the 0C EDG also had an inadequate
short-time over-current setpoint for its equivalent breaker and that the breaker would trip
open prematurely under certain conditions.
The inspectors determined that Calvert Cliffs did not verify and check the adequacy of
the short-time over-current setting of 1MCC123 Amptector for the simultaneous starting
of all loads that would be expected under design basis conditions. Instead,
Constellation used a design standard which only included the largest single load starting
with other loads already running. This standard was not appropriate for the 1MCC 123
supply breaker setting. This issue was most likely not identified in previous surveillance
tests because the setpoint had not drifted as low and because the testing was
conducted during the months when the normal ambient temperature was too low to
require additional ventilation fans to start.
The licensee, prior to plant startup on Unit 1, reset the 1MCC123 breaker trip to 3600
amps to address the issue. The Amptector setpoints are calibrated once every two
years. The calibration procedure allowed the technicians a +/- 2% as-left setpoint
tolerance and the allowable as-found setpoint deviation can be +/- 10%. The inspectors
have reviewed the new short-time over-current setting and have found it acceptable.
The licensee also corrected the setpoint of the equivalent breaker which supplies the
support systems for the 0C EDG.
Analysis: The performance deficiency involved the failure to provide adequate design
control, as required by 10 CFR 50 Appendix B, Criteria III, for the 1A EDG support
system supply circuit breaker short-time over-current trip setpoint. The setpoint was too
low, it did not account for the in-rush current associated with the possible combinations
of equipment that could start and operate to support the 1A EDG following a LOOP.
This low setpoint and normal setpoint drift resulted in substantial periods where the 1A
EDG would not have been able to perform its safety function, because the support
system supply circuit breaker would have tripped open inappropriately.
The inspectors determined, by reviewing licensee data concerning setpoint drift and
outside air temperatures that the 1A EDG would not have been able to perform its
safety function for an exposure period of approximately 5407 hours0.0626 days <br />1.502 hours <br />0.00894 weeks <br />0.00206 months <br /> over the last year of
reactor operation. The setpoint drift contributed approximately 5046 hours0.0584 days <br />1.402 hours <br />0.00834 weeks <br />0.00192 months <br /> when outside
air temperature was below 85 °F (one room cooling fan start) and the low setpoint
contributed approximately 361 hours0.00418 days <br />0.1 hours <br />5.968915e-4 weeks <br />1.373605e-4 months <br /> when outside air temperature was at or above 85
4
Enclosure
°F(more than one room cooling fan start). The finding was more than minor because it is
associated with the Mitigating Systems Cornerstone attribute of design control and
affected the cornerstone objective to ensure the availability and reliability of systems
(i.e., emergency AC power) that respond to initiating events to prevent undesirable
consequences.
The 0C EDG (station blackout diesel) support system supply circuit breaker short-time
over-current trip was similarly affected by this performance deficiency, however, the
as-found setting was not as low as that found on the 1A EDG. The inspectors verified
licensee data that indicated that the 0C EDG would not have been able to perform its
required function when outside air temperature was at or above 95 °F (more than two
room cooling fans start). It was estimated that this exposure period was 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
To assess the full significance of the performance deficiency, both Phase 2 and Phase 3
analyses were based on the 5400 hour0.0625 days <br />1.5 hours <br />0.00893 weeks <br />0.00205 months <br /> exposure for the 1A EDG being unable to
perform its safety function and additional 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> where both the 1A EDG and the
0C EDG would not have been able to perform their required functions.
The finding was evaluated in accordance with IMC 0609, Appendix A, "Significance
Determination of Reactor Inspection Findings for At-Power Situations," using Phase 1,
Phase 2, and Phase 3 SDP analyses. The Phase 1 screening required performance of
a Phase 2 evaluation because the finding represented a loss of safety function of a
single train, for greater than its allowed outage time. The TS allowed outage time is 14
days for a single EDG. A Licensee Event Report (LER) was written on this issue.
The internal events Phase 2 analysis, conducted using the Risk-informed Inspection
Notebook for Calvert Cliffs Nuclear Power Plant Units 1 and 2, revision 2, dated
September 30, 2005 and with IMC 0609, Appendix H, Containment Integrity SDP,
determined that the finding could have substantial safety significance based on )CDF
with no associated change in LERF.
The Region I SRA conducted a Phase 3 Risk Assessment, to refine the Phase 2
analysis and to incorporate external events and recovery credit. The analysis used an
updated Calvert Cliffs SPAR model, Rev 3 plus, dated October 28, 2005. Since the
finding only affected the 1A EDG and the 0C DG, the emergency power system
common cause factors were adjusted to more accurately reflect the physical differences
between the site EDGs. Additionally, the shutdown risk was evaluated and considered
to be small compared to the operational risk.
The final assessment results indicate that the finding had low to moderate safety
significance and was determined to be WHITE for Unit 1 based on )CDF of
approximately 1 in 150,000 years of operation (mid E-6 per year range) for both internal
and external events, with no associated increase in LERF. The Phase 3 internal events
analysis resulted in an increase in )CDF of in the low E-6 range for the 5407-hour
exposure period. The dominant internal event core damage sequence was a station
blackout with a successful reactor shutdown along with a failure to recover the EDGs
and restore offsite power in four hours. A LOOP was the second dominant internal
event sequence with a successful reactor shutdown and the EDGs powering a safety
bus with auxiliary feedwater (AFW) and failure of once through cooling. The SRA
determined, based on information from the licensee, that the dominate external event
5
Enclosure
was a large turbine building fire which causes a LOOP and challenges AFW and failure
of once through cooling. The contribution in )CDF from external events was in the mid
E-6 range. The SRA reviewed the licensees risk assessment which included at-power
internal and external events, shutdown and LERF. Using similar assumptions to those
used in the Phase 3 analysis, the licensee estimated the )CDF to be similarly in the mid-
E-6 range. The risk of the 1A EDG exposure time dominated the analysis by several
orders of magnitude over the risk of the concurrent 1A EDG and 0C EDG exposure
period.
Old Design Issue Considerations: As defined in NRC Inspection Manual Chapter (IMC) 0305, issued date 06/22/06, an old design issue is an inspection finding involving a past
design-related problem in the engineering calculations or analysis, associated operating
procedure or installation of plant equipment that does not reflect a performance
deficiency associated with existing licensee programs, policy, or procedures. As
discussed in IMC 0305 Section 06.06.a, some old design issues may not be considered
in the assessment program.
This issue was self-revealed during the performance of a two-year surveillance test
which is associated with an existing licensee program. Section 06.06.a.1 states, in part,
that the NRC may refrain from considering safety significant inspection findings in the
assessment program for a design-related finding in engineering calculations or analysis,
associated operating procedure or installation of plant equipment that meets all four
criteria listed in IMC 0305, Section 06.06.a.1, Treatment of Old Design Issues in the
Assessment Process. The first criterion states that the finding was licensee-identified as
a result of a voluntary initiative such as a design basis reconstitution. This section
further states that for the purpose of this manual chapter, self-revealing issues are not
considered to be licensee-identified. In this case, the design issue (incorrect short-time
over-current setting of 1MCC123 Amptector) was identified as the result of a
self-revealing event and the licensee had a previous opportunity to identify the issue
during the EDG installation. Therefore, this design-related finding will not be treated as
an old design issue.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that design control measures shall provide for verifying or checking the adequacy
of design, such as by the performance of design reviews, by the use of alternate or
simplified calculational methods or by the performance of a suitable testing program.
Contrary to the above, in 1996, Calvert Cliffs did not verify or check the adequacy of the
short-time over-current setting of the 1MCC123 Amptector when the 1A EDG was
installed. This issue was entered into the corrective action program at Calvert Cliffs
(IRE-013-237) for resolution. Corrective actions included changes to the over-current
trip setting (AV 50-317/20060012-01, Failure to Adequately Control the Design of the
1A EDG Feeder Breaker for Essential EDG Support Systems).
6
Enclosure
4OA6 Meetings, Including Exits
Exit Meeting
On August 16, 2006, the inspectors presented the inspection results to yourself and
other members of your staff, who acknowledged the findings. The inspectors asked
Constellation whether any materials examined during the inspection should be
considered proprietary. No proprietary information was identified.
ATTACHMENTS:
SUPPLEMENTAL INFORMATION
MOTOR CONTROL CENTER LOADS
A-1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
J. Spina
Site Vice President
J. Pollock
Plant General Manager
P. Furio
Licencing Supervisor
J. Mills
System Engineering General Supervisor
M. Simpson
Supervisor, Electrical and Controls System Engineering
A. Julka
Director Fleet PRA
J. Stone
Principal Engineer
A. Simpson
Sr. Licensing Engineer
L. Larragoite
Director Fleet Licensing
M. Flaherty
Manager - Engineering
S. Loeper
EDG system manager
R. Stark
Design Engineer
J. Boggs
Design Engineer
A. Julka
Director, PRA
J. Wynn
System Engineer
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Failure to Adequately Control the Design of the
Setpoints for 1A EDG Feeder Breaker for
Essential EDG Support Systems (Section 2)
Closed
None
Discussed
None
A-2
LIST OF DOCUMENTS REVIEWED
Procedures and Documents
STP O-004A-1, Unit 1A Train Engineered Safety Features Test, Revision 25
FTE-29, Acceptance Test and Calibration of Amptectors, Revision 8
York 507-40053, Direct Digital Control System for 0C-AHU-1 & 0C-AHU-2
Calculations and Studies
Calculation D-E-94-001, Bechtel Design Calculation for SACM Diesel Project, Revision 7
Calvert Cliffs Nuclear Power Plant Risk Evaluation for the 1A DG MCC Feeder Breaker Trip on
3/24/2006 and 7/25/2006
ES 200400160, Equivalency Evaluation for New Westector Trip Devices in Place of
Westinghouse Amptectors, Revision 0
ES 200600156, Engineering Evaluation for Breaker Short Time Over-current Settings 52-1703
(of MCC 123) and 52-0703 (MCC 023)
Condition Reports/Evaluations
IRE-013-237, 1MCC123 Feeder Breaker tripped, 3/27/2006
Action Item Tracking (AIT) IR200600072, Apparent Cause for 1MCC123 Feeder Breaker
Tripping, Revision 1
Drawings
61001SH0001, Electrical Main Single Line Diagram
61-027-E, Diesel Generator Project Single Line Diagram, Diesel Generator 1A, SH1, Rev. 2
61-085-C, Diesel Generator Project Schematic Diagram, 1A Emergency DG HVAC System
Preheat Duct Heater DH-4, Sh 106, Revision 2
62429SH0001, HVAC System P & ID, Diesel Generator Building 1, Revision 5
60-626-B, Logic Diagram, DG Building 2, Switchgear Room Air Handling Unit 0C-AHU-2,
SH004, Revision 1
60-626-B, Logic Diagram, DG Room Exhaust Fans 0C-F-1 Thru OC-F-4, Sh 48, Revision 1
61085SH0077, Schematic Diagram Diesel Generator 1A Building Supply Fan F-10, Revision 5
Work Orders (WO)
WO 0200402727, Inspect 52-0703 (MCC 023 Feed) Per FTE-077
WO 1200403460, Inspect 52-1703 (MCC 123 Feed) Per FTE-077
WO 0200002152, Inspect 52-0703 (MCC 023 Feed) Per FTE-077 and Calibrate the Amptector
per FTE 29
WO 1200601442, Replace Amptector 52-1703 and Calibrate the Amptector per FTE 29
A-3
LIST OF ACRONYMS
Agency Document Access and Management System
Action Item Tracking
Apparent Violation
CR
Condition Report
Direct Current
DS
Diesel Generator
Heating, ventilating and air-conditioning
IMC
Inspection Manual Chapter
LER
Licensee Event Report
Motor Control Center
NRC
Nuclear Regulatory Commission
Publicly Available Records System
Significance Determination Process
Senior Reactor Analyst
Surveillance Test Procedure
TS
Technical Specification
Updated Final Safety Analysis Report
Work Order
B-1
MOTOR CONTROL CENTER (MCC) LOADS
1MCC123 supplies power to all essential 480 VAC components for the operation of the
1A EDG at Calvert Cliffs Unit 1. The MCC is fed from the 4.16kV safety-related bus 11
that is powered from the 1A EDG output or the offsite power supply. A step down
transformer reduces the voltage to 480 VAC and supplies the power to 1MCC123 and
another MCC (MCC124). The loads connected to the 480 VAC 1MCC123 are as
follows:
a.
Six radiator fans (40 HP with an estimated inrush current of 247.5 amps each)
b.
Two ventilation fans (F10 &F12, 40 HP with an estimated inrush current of 283
amps each)
c.
Two fuel oil feed pump motors (P11 and P12, 5 HP with an estimated inrush
current of 12.76 amps each
d.
One battery room exhaust fan (F8, 1.5 HP with an estimated inrush current of
20.37 amps)
e.
One battery charger (estimated inrush current of 6.17 amps)
f.
One distribution panel (1P23 with estimated inrush current of 1.44 amps)
g.
Four room ventilation fans (F1, F2, F3, and F4, 20 HP with an estimated inrush
current of 148.1 amps each)
h.
One intake duct heater (DH-4, 70 KW with steady state current of 79 amps)
i.
One battery room unit heater with fan (estimated inrush current of 21 amps)
j.
Two pre-lube pumps (not expected to be running during EDG operation)
The operation of the above components is such that when the 1A EDG starts and
energizes the 4.16kV Safety Bus 17, it in turn energizes the 480 VAC 1MCC123, items
a through f and at least one fan of item g starts immediately (regardless of the diesel
room temperature). The other three fans of item g will start, sequentially, when the
room temperature exceeds 85°F, 95°F, and 105°F. If the diesel room temperature is
above 105°F when the 1MCC123 is energized then all four room fans listed in item "g"
will start simultaneously. Item i is set at 78°F to maintain the battery room at or above
the temperature setting.
Item h is controlled through two silicon controlled rectifiers (SCRs), which responds to
a thermister (a resistance temperature sensor) in the outdoor air inlet duct, and is set at
40°F. It is controlled with a control band of 10°F that is fully heated at 35°F and has no
heat at 45°F. However, the current design will always energize the heater initially (for
about 2 seconds) when power is restored to the bus. This heater design led to an overall
increase in the in-rush current the breaker will experience when power is restored to the
bus.
[NRC FORM 665S
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NRC INSPECTION REPORT: Calvert Cliffs Unit 1
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