ML080170137
| ML080170137 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/17/2008 |
| From: | Julio Lara Engineering Branch 3 |
| To: | Pardee C Exelon Generation Co |
| References | |
| IR-08-007 | |
| Download: ML080170137 (16) | |
See also: IR 05000237/2008007
Text
January 17, 2008
Mr. Charles G. Pardee
Chief Nuclear Officer and
Senior Vice President
Exelon Generation Company, LLC
4300 Winfield Road
Warrenville IL 60555
SUBJECT: DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3
NRC INSPECTION REPORT 05000237/2008007; 05000249/2008007
Dear Mr. Pardee:
On December 18, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed
an inspection at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed inspection
report documents the inspection findings, which were discussed on December 18, 2007, with
Mr. J. Ellis 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.
Based on the results of this inspection, two NRC identified findings of very low safety
significance (Green) were identified. All of these issues involved violations of NRC
requirements. However, because of the very low safety significance and because they
were entered into your corrective action program, the NRC is treating these violations as
Non-Cited Violations (NCVs) consistent with Section VI.A.1. of the NRC Enforcement Policy.
If you contest the subject or severity of an NCV, you should provide a response within 30 days
of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to
the Regional Administrator, Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352;
the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at the Dresden Nuclear Power Station.
C. Pardee -2-
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 Agencywide Documents Access and Management System (ADAMS),
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public
Electronic Reading Room).
Sincerely,
/RA by A. M. Stone Acting For/
Julio F. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
Docket Nos. 50-237; 50-249;72-037
Enclosure: 1. Notice of Violation
2. Inspection Report 05000237/2008007; 05000249/2008007
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Dresden Nuclear Power Station
Plant Manager - Dresden Nuclear Power Station
Regulatory Assurance Manager - Dresden Nuclear Power Station
Chief Operating Officer and Senior Vice President
Senior Vice President - Midwest Operations
Senior Vice President - Operations Support
Vice President - Licensing and Regulatory Affairs
Director - Licensing and Regulatory Affairs
Manager Licensing - Clinton, Dresden, and Quad Cities
Associate General Counsel
Document Control Desk - Licensing
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer
Chairman, Illinois Commerce Commission
C. Pardee -2-
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 Agencywide Documents Access and Management System (ADAMS),
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public
Electronic Reading Room).
Sincerely,
Julio F. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
Docket Nos. 50-237; 50-249;72-037
Enclosure: 1. Notice of Violation
2. Inspection Report 05000237/2008007; 05000249/2008007
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Dresden Nuclear Power Station
Plant Manager - Dresden Nuclear Power Station
Regulatory Assurance Manager - Dresden Nuclear Power Station
Chief Operating Officer and Senior Vice President
Senior Vice President - Midwest Operations
Senior Vice President - Operations Support
Vice President - Licensing and Regulatory Affairs
Director - Licensing and Regulatory Affairs
Manager Licensing - Clinton, Dresden, and Quad Cities
Associate General Counsel
Document Control Desk - Licensing
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer
Chairman, Illinois Commerce Commission
DOCUMENT NAME: G:\DRS\Work in Progress\DRESDEN 2008 007 DRS RCD.doc
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
NAME RDaley: ls JLara
DATE 01/17/08 01/17/08
OFFICIAL RECORD COPY
Inspection Report to Mr. Charles Pardee from Mr. Julio Lara dated January 17, 2008.
SUBJECT: DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3
NRC INSPECTION REPORT 05000237/2008007; 05000249/2008007
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No.(s): 50-237; 50-249
License No.(s): DPR-19; DPR-25
Report No: 05000237/2008007; 05000249/2008007
Licensee: Exelon Generation Company
Facility: Dresden Nuclear Power Station, Units 2 and 3
Location: Morris, IL 60450
Dates: December 3, 2007 through December 18, 2007
Inspectors: R. C. Daley, Senior Reactor Engineer
Approved by: J. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
SUMMARY OF FINDINGS
IR 05000237/2008007; 05000249/2008007; 12/03/2007 - 12/18/2007; Exelon
Generation Company, Dresden Nuclear Power Station, Units 2 and 3, Component Design Bases
Inspection Followup
This report covers a 2-week period of routine inspections by Region III inspectors. Two Green
findings, involving Non-Cited Violations, (NCVs) were identified. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual
Chapter 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 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Events
- SL IV. The inspectors identified a Severity Level IV NCV associated with the failure to
perform a safety evaluation in accordance with 10 CFR 50.59. Specifically, the licensee
failed to perform a safety evaluation when non-conservatively changing the design basis
loading for the emergency diesel generators (EDG) in the design calculation. This
resulted in the expected loading during a design basis accident no longer being bounded
by the EDG endurance testing requirements contained in the Technical Specifications
(TS). Because the licensee did not also evaluate the effect on the existing endurance test
loading requirements, the testing no longer adequately verified the capability of the EDG
to power its predicted loading during a LOOP/LOCA. This adverse change increased the
probability of a malfunction of equipment important to safety EDG during a
LOOP/LOCA event.
Because the issue affected the NRC=s ability to perform its regulatory function, this
finding was evaluated using the traditional enforcement process. The finding was
determined to be more than minor because the inspectors could not reasonably determine
that the change in EDG loading, which adversely affected equipment important to safety,
would not have ultimately required NRC approval. The finding was determined to be of
very low safety significance (Green) based on the results of the SDP Phase 1 screening
worksheet. Specifically, the licensee was eventually able to demonstrate through an
engineering evaluation, that the EDG loads would not exceed the bounding values
contained in the endurance test criteria. The inspectors determined that there was no
cross cutting aspect to this issue. (Section 1R21.b.1)
- Green. The inspectors identified an NCV for the failure to meet the requirements
contained in SR 3.8.1.15. Specifically, the testing that the licensee performed to meet SR 3.8.1.15 did not test to a power factor as close to the accident load power factor as
1 Enclosure
possible. These testing methods did not demonstrate the capability of the EDG to support
the emergency core cooling system loading and were non-conservative. The issue was
entered into the licensee=s corrective action program.
2 Enclosure
The issue was more than minor because it was associated with the Mitigating System
Cornerstone attribute of AEquipment Performance,@ and affected the cornerstone objective of
ensuring the availability and reliability of the EDGs. Specifically, the licensee=s testing
methods for SR 3.8.1.15 did not demonstrate the capability of the EDG to support ECCS
loading and was non-conservative. This finding was of very low safety significance, because
the inspectors answered ANo@ to all five questions under the Mitigating Systems Cornerstone
column of the Phase 1 worksheet. Specifically, the licensee subsequently performed the
required testing in SR 3.8.1.15 to the expected power factor, and the EDGs performed
satisfactorily. The inspectors determined that there was no cross-cutting aspect to this issue.
(Section 1R21.b.2)
B. Licensee-Identified Violations
No findings of significance were identified.
3 Enclosure
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Mitigating Systems
1R21 Component Design Basis Inspection (71111.21)
a. Inspection Scope
During this inspection period, inspectors completed followup inspection activities
concerning URI 05000237/249/2002006-02. This URI contained two issues regarding
whether the licensees Technical Specifications (TS) surveillance provided reasonable
assurance of the emergency diesel generators (EDG) capability to carry design basis loads
and whether operating the EDG at the reactive load for only 10 minutes of the 24-hour run
met the supporting regulatory analysis and intent of the surveillance requirement.
b. Findings
b.1 Failure to Perform a 10 CFR 50.59 Evaluation for Exceeding Continuous Rating on the EDG
Introduction: The inspectors identified a Severity Level IV Non-Cited Violation (NCV) of
10 CFR 50.59 having very low safety significance (Green) for the failure to perform an
adequate safety evaluation in accordance with 10 CFR 50.59. Specifically, the licensee failed
to perform a safety evaluation when non-conservatively changing the design basis loading for
the plant EDGs in their design calculation.
Description: The inspectors identified that the licensee=s calculated design basis loads for a
Loss of Offsite Power with a Loss of Coolant Accident (LOOP/LOCA) exceeded the
continuous rating of the EDG. Dresden Technical Specification (TS) Surveillance
Requirement (SR) 3.8.1.15 requires the following:
Verify each DG [diesel generator] operating within the power factor limit operates for
24-hours:
a. For $ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded $ 2730 kW [kiloWatt] and # 2860 kW (105 percent
to 110 percent of continuous rating); and
b. For the remaining hours of the test loaded $ 2340 kW and # 2600 kW
(90 percent to 110 percent of continuous rating)
The intent of this endurance test is to demonstrate that each EDG is in operational readiness
to assume the design-basis (LOOP/LOCA) loads even when the redundant EDG has failed.
An endurance test (24-hour run) is considered to be a reasonable duration to ascertain if the
EDG capability continues to remain intact for a potentially long-term operation that would be
4 Enclosure
needed to bring the plant to a safe shutdown following a design basis event. The test
challenges whether the fuel system will continue to supply fuel in order to keep up with the
maximum and varying load demand, the excitation system will produce sufficient magnetic
field to maintain voltage, and the voltage regulator will maintain the voltage within
acceptable limits. This can be achieved only when the diesel engine is loaded to its expected
design basis loading conditions, and when the generator is producing sufficient voltage and
current that reflect design basis accident loading. This endurance run verifies the EDG
capability should a LOOP/LOCA occur.
Based on the above considerations, as well as review of pertinent regulatory guidance
documents, the inspectors concluded that the surveillance test loading should bound
the accident loading. However, at Dresden, anticipated LOOP/LOCA loading exceeded
the upper limit of the endurance test requirement (2600 kW). In the case of the No. 2 EDG,
anticipated loading could be as high as 2851 kW for extended periods of time during
such postulated events.
Originally, the licensees design basis calculations and the facilitys Final Safety
Analysis Report (FSAR) predicted manual EDG loading to be below 2600 kW. However,
during the life of the plant, the design basis EDG loading calculations were revised such that
predicted LOOP/LOCA loading increased. During that time period, Calculation Number
71317-33-19-2, Revision 17, was processed which changed the design and license
basis predicted loading such that loads exceeded 2600 kW. Prior to this revision, the
endurance testing fully demonstrated the capability of the EDGs during a LOOP/LOCA;
however, once loads were predicted to be greater than 2600 kW, the testing performed under
SR 3.8.1.15 no longer adequately demonstrated the capability of the EDG to power
its predicted loading during a LOOP/LOCA. The inspectors concluded that without
changing the testing to reflect the new predicted loading, the probability of the EDGs
being incapable of carrying predicted loads during a LOOP/LOCA was increased. The
design basis reliability of the EDG had been reduced. At the time of the calculation revision,
the licensee did not perform an evaluation pursuant to 10 CFR 50.59. The inspectors
determined that had the licensee performed such an evaluation, the change would have
required a license amendment because the probability of a malfunction of equipment
important to safety (i.e., EDG) would have been increased.
The licensee provided the inspectors with historical documents that appeared to show that the
NRC had accepted that the auto-connected loads could exceed the continuous rating of the
EDG. However, the inspectors could not find any definitive information that would lead
them to conclude that the NRC had previously approved that manual loading of the EDG
could exceed the continuous rating. The distinction between the two is important, because
auto-connected loads are only at their maximum during the initial stages of the event. Shortly
after the event initiation, the auto-connected loads become much less. This initial spike in
loads is clearly bounded by the EDG testing, since 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the endurance run is performed
between 105 percent and 110 percent of the EDG continuous rating. However, manually
connected loads can last for the duration of the event. Therefore, if the manually connected
loads exceeded the continuous rating, the 24-hour endurance run, if performed at the
5 Enclosure
continuous rating, would not bound the actual loads that would be running for the duration of
the event.
This issue was addressed by the licensee in a Dresden Engineering White Paper. In
the White Paper, the licensee reassessed the loading values for the EDGs removing
conservatisms and loads that would not be used later (10 minutes into the event) in the
LOOP/LOCA event. By doing this, the licensee was able to conclude that even though
initial EDG loading may exceed 2600 kW, the loading at 10+ minutes would be sharply
6 Enclosure
reduced below the 2600 kW value. Since SR 3.8.1.15 tests for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at levels above
2600 kW that bound the initial loading of the EDGs, the inspectors found this to be
acceptable. The licensee planned to incorporate the results of this analysis into a
calculation revision. The licensee entered the issue into their corrective action program as
condition report (CR) 485889.
Analysis: The inspectors determined that the licensees failure to perform a safety
evaluation in accordance with 10 CFR 50.59 for changes to their design and license
basis was a performance deficiency warranting a significance determination. Specifically,
Calculation 7317-33-19-2, Revision 17, was processed in 1994 which changed design
and license basis predicted loading for the EDGs such that loads exceeded 2600 kW.
Because the licensee did not evaluate this change pursuant to 10 CFR 50.59, the TS EDG
surveillance testing in SR 3.8.1.15 no longer adequately verified the capability of the EDG
to power its predicted loading during a LOOP/LOCA. This adverse change increased the
probability of a malfunction of equipment important to safety (i.e., EDG) during a
LOOP/LOCA event.
Following discussions, a subsequent licensee evaluation showed that loading would have
remained below 2600 kW at time = 10+ minutes. However, this did not change the impact
of the initial change to the design and license basis loading. The changes to the loading
calculation changed the accident analyses such that exceeding 2600 kW on the EDGs was
determined to be acceptable with no further regulatory reviews performed. The inspectors
concluded that the calculation revision resulted in a change to the analyses that would not
have been acceptable under 10 CFR 50.59 and would have required either approval from
the NRC or a license amendment for a change to the loading requirement for SR 3.8.1.5.
Because violations of 10 CFR 50.59 are considered to be violations that potentially
impede or impact the regulatory process, they are dispositioned using the traditional
enforcement process instead of the significance determination process (SDP). The
finding was determined to be more than minor because the inspectors could not reasonably
determine that the changes to the licensees design basis would not have ultimately
required NRC prior approval.
The inspectors completed a significance determination of the underlying technical issue
using NRC=s inspection manual chapter (IMC) 0609, Appendix A, ASignificance
Determination of Reactor Inspection Findings for At-Power Situations,@ and answered
ANo@ to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet.
Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low
safety significance (Green). In accordance with the Enforcement Policy, the violation was
therefore classified as a Severity Level IV violation. The inspectors determined that there
was no cross cutting aspect to this issue.
Enforcement: Title 10 CFR 50.59(d)(1) states, in part, that the licensee shall maintain
records of changes in the facility, of changes in procedures, and of tests and experiments.
7 Enclosure
These records must include a written evaluation which provides a basis for the
determination that the change, test, or experiment does not require a license amendment.
Contrary to the above, the licensee failed to perform a safety evaluation when increasing
the design basis load on the EDGs while revising Calculation 7317-33-19-2, Revision 17,
in 1994. Since the EDG load testing required by the facilitys TS was no longer bounding
8 Enclosure
after this change, the probability of a malfunction of equipment important to safety (i.e.,
EDGs) was affected. In accordance with the Enforcement Policy, this violation of the
requirements of 10 CFR 50.59 was classified as a Severity Level IV Violation because the
underlying technical issue was of very low safety significance. Because this violation was
of very low safety significance, was not repetitive or willful, and it was entered into the
licensee=s corrective action program (CR 485889), this violation is being treated as an NCV
consistent with VI.A.1 of the NRC Enforcement Policy (NCV). (NCV 05000237/2008007-
b.2 Failure to Meet the EDG Power Factor Testing Requirements
Introduction: The inspectors identified an NCV having a very low safety significance
(Green) for the failure to meet the EDG power factor testing requirements contained in
TS SR 3.8.1.15. Specifically, the testing that the licensee performed to meet SR 3.8.1.15
did not test to a power factor (pf) as close to the accident load power factor as possible.
These testing methods did not demonstrate the capability of the EDG to support ECCS
loading and were non-conservative.
Description: Dresden Technical Specification (TS) Surveillance Requirement (SR)
3.8.1.15 requires the following:
Verify each DG [diesel generator] operating within the power factor limit operates for
24-hours:
a. For $ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded $ 2730 kW and # 2860 kW (105 percent to 110 percent
of continuous rating); and
b. For the remaining hours of the test loaded $ 2340 kW and # 2600 kW
(90 percent to 110 percent of continuous rating).
This SR is modified by a note which states, AIf grid conditions do not permit, the power
factor limit is not required to be met. Under this condition, the power factor shall be
maintained as close to the limit as practicable.@
The Bases for this TS SR states, AIn order to ensure that the DG is tested under load
conditions that are as close to design conditions as possible, testing must be performed at a
power factor as close to the accident load power factor as practicable. When synchronized
with offsite power, the power factor limit is # 0.85. This power factor is chosen to bound
the actual worst case inductive loading that the DG could experience under design basis
accident conditions.@
The licensee developed surveillance procedure DOS 6600-12, Diesel Generator Tests
Endurance and Margin/Full Load Rejection/ECCS/Hot Restart, to demonstrate compliance
with SR 3.8.1.15. In this surveillance test, the diesel is connected to the grid and operated
for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a load between 2730 and 2860 kW for the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of the test.
9 Enclosure
Sometime during this 22-hour period, the power factor is adjusted by increasing KVARs to
a band of 1550 to 1600 (0.83 - 0.86 pf) if possible, keeping the voltage on the emergency
bus less than 4300 volts. This is held for only 10 minutes before returning to the " 300
kVAR band. The inspectors were concerned, because this testing did not meet the intent of
the TS SR because it did not bound the actual worst case inductive loading that the
10 Enclosure
EDG could experience under design basis accident conditions. The licensee only tested
to 0.85 pf for 10 minutes during the performance of SR 3.8.1.15. While the TS did
provide provisions to accommodate cases when the plant conditions cannot allow
generation of sufficient kVAR to match design basis kVAR loading, the licensee=s testing
methodology did not follow these provisions. Prior to EDG testing, the licensee did not
perform any evaluation as to the condition of the grid, with respect to whether or not the
power factor limit can be achieved. Rather, regardless of whether the grid conditions may
support testing at the power factor limit, the licensee had established a testing practice
that only tests at this limit for 10 minutes.
The inspectors determined that the licensee=s testing methods did not demonstrate the
capability of the EDG to support ECCS loading and was non-conservative. The licensee
entered this issue into their corrective action program as CR 485889.
Analysis: The inspectors determined that the licensees failure to meet the EDG power
factor testing requirements contained in TS SR 3.8.1.15 was a performance deficiency
warranting a significance determination. Specifically, the testing that the licensee
performed to meet SR 3.8.1.15 did not test the EDG at a power factor as close to the
accident load power factor as possible. The issue was more than minor because it was
associated with the Mitigating System Cornerstone attribute of AEquipment Performance,@
and affected the cornerstone objective of ensuring the availability and reliability of the
EDGs. Specifically, the licensee=s testing methods for SR 3.8.1.15 did not demonstrate
the capability of the EDG to support ECCS loading and was non-conservative.
The finding screened as having very low significance (Green) using IMC 0609,
Appendix A, ASignificance Determination of Reactor Inspection Findings for the
At-Power Situations,@ because the inspectors answered ANo@ to all five questions under
the Mitigating Systems Cornerstone column of the Phase 1 worksheet. Specifically, the
licensee subsequently performed the required testing in SR 3.8.1.15 to the expected
power factor, and the EDGs performed satisfactorily. The inspectors determined that
there was no cross cutting aspect to this issue.
Enforcement: Dresden SR 3.8.1.15 states, in part, that the licensee is required to verify
each EDG operates within the power factor limit for 24-hours. This testing methods for
this SR should demonstrate the capability of the EDG to support ECCS loading. This
capability includes the operating at the power factor that would be expected during the
design basis accident (LOCA) event.
Contrary to the above, the testing that the licensee performed to meet SR 3.8.1.15 did not
test to a power factor as close to the accident load power factor as possible. The
inspectors determined that the licensee=s testing methods did not demonstrate the
capability of the EDG to support ECCS loading and was non-conservative. Because this
failure to comply with the requirements in SR 3.8.1.15 was determined to be of very low
safety significance and because it was entered in the licensee=s corrective action program
11 Enclosure
as CR 485889, this violation is being treated as an NCV, consistent with Section VI.A.1
of the NRC Enforcement Policy. (NCV 05000237/2008007-02; 05000249/2008007-02)
12 Attachment
4. OTHER ACTIVITIES (OA)
4OA5 Other Activities
.1 (Closed) Unresolved Item (URI) 0500237/249/2002006-02: Non-Conservative
Emergency Diesel Generator Testing
This URI contained two issues regarding whether the licensees Technical Specifications
(TS) surveillance provided reasonable assurance of the emergency diesel generators
(EDG) capability to carry design basis loads and whether operating the EDG at the
reactive load for only 10 minutes of the 24-hour run met the supporting regulatory
analysis and intent of the surveillance requirement. This URI had been previously
updated in inspection report 0500237/249/2005-009. Based on the information discussed
in Section 1R21.b of this report, two NCVs were identified. Therefore, this URI is
closed.
4OA6 Meetings
.1 Exit Meetings
The inspectors presented the inspection results to Mr. James Ellis and other licensee
members at the conclusion of the inspection on December 18, 2007. The inspectors
asked the licensee whether any materials examined during the inspection should be
considered proprietary. No proprietary information was identified.
ATTACHMENT: SUPPLEMENTAL INFORMATION
13 Enclosure
14 Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
J. Ellis, Regulatory Assurance Manager
Nuclear Regulatory Commission
M. Ring, Chief, Division of Reactor Projects, Branch 1
R. Schulz, Illinois Emergency Management Agency
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Opened and Closed
05000237/2008007-01; NCV Failure to Perform a 10 CFR 50.59 Evaluation for
05000249/2008007-01 Exceeding Continuous Rating on the EDG. (Section
1R21.b.1)05000237/2008007-02; NCV Failure to Meet the EDG Power Factor Testing
05000249/2008007-02 Requirements. (Section 1R21.b.2)
Closed
05000237/2002006-02; URI Non-Conservative Emergency Diesel Generator Testing.05000249/2002006-02 (Section 4OA5.1)
Discussed
None
1 Attachment
2 Attachment
LIST OF DOCUMENTS REVIEWED
Calculations
Calculation 7317-33-19-2; Diesel Generator Loading Under Design Basis Accident Condition;
dated January 21, 1993
Calculation 7317-33-19-2; Diesel Generator Loading Under Design Basis Accidents Condition;
dated January 28, 1994
Condition Reports
CR 485889, Op Eval 06-002; EDG Endurance Test Operability Evaluation; dated
October 26, 2006
CR 140598; EDG Loading Values Removed from the UFSAR; dated January 22, 2003
3 Attachment
4 Attachment
LIST OF ACRONYMS USED
CR Condition Report
ECCS Emergency Core Cooling System
EDG Emergency Diesel Generator
IMC Inspection Manual Chapter
kW Kilowatts
kVAR Kilovolt Amps Reactive
LOCA Loss of Coolant Accident
LOOP Loss of Offsite Power
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
pf Power Factor
SDP Significance Determination Process
SR Surveillance Requirement
TS Technical Specification
URI Unresolved Item
5 Attachment