ML070720181
| ML070720181 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/22/2007 |
| From: | Widmann M Reactor Projects Region 2 Branch 6 |
| To: | Singer K Tennessee Valley Authority |
| References | |
| IR-06-002, IR-06-005 | |
| Download: ML070720181 (32) | |
See also: IR 05000327/2006005
Text
February 22, 2007
Tennessee Valley Authority
ATTN: Mr. Karl W. Singer
Chief Nuclear Officer and
Executive Vice President
6A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
SUBJECT:
ERRATA LETTER FOR SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED
INSPECTION REPORT 05000327/2006005, 05000328/2006005 AND
07200034/2006002
Dear Mr. Singer:
On December 31, 2005, the United States Nuclear Regulatory Commission (NRC) completed
an inspection at your Sequoyah Nuclear Plant, Units 1 and 2. The above inspection report was
issued without three inspection findings and the closeout of four unresolved items which were
discussed in a conference call between Mr. R. Schin of this office and Mr. D. Kulisek and other
members of the your staff on December 20, 2006. The purpose of this letter is to include those
items in the inspection report and to ask that you replace the enclosed revised pages in your
original document.
The three additional inspection findings were of very low safety significance and were
determined to involve violations of NRC requirements. However, because of their very low
safety significance and because they are entered into your corrective action program, the NRC
is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the
NRC Enforcement Policy. If you contest any of the additional NCVs in the enclosed revised
pages, you should provide a response within 30 days of the date of this errata letter, with the
basis for your denial, to the United States Nuclear Regulatory Commission, ATTN.: Document
Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator Region
II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission,
Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Sequoyah Nuclear
Plant.
In accordance with 10 CFR 2.390 of the NRCs 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 (PARS) component of
2
NRCs 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).
If you have any questions, please contact me at (404) 562-4550.
Sincerely,
/RA/
Malcolm T. Widmann, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos. 50-327, 50-328,72-034
Enclosure:
Errata pages for Inspection Report 05000327/2006005 and 05000328/2006005
and 07200034/2006002
cc: w/encl: (See page 3)
_________________________
OFFICE
RII:DRP
RII:DRP
RII:DRS
SIGNATURE
/RA/
/RA/
/RA By MThomas for/
NAME
LGarner
MWidmann
CPayne
DATE
2/22/07
2/22/07
2/22/07
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
3
cc w/encl:
Ashok S. Bhatnagar
Senior Vice President
Nuclear Operations
Tennessee Valley Authority
Electronic Mail Distribution
Preston D. Swafford
Senior Vice President
Nuclear Support
Tennessee Valley Authority
Electronic Mail Distribution
Larry S. Bryant, Vice President
Nuclear Engineering &
Technical Services
Tennessee Valley Authority
Electronic Mail Distribution
Randy Douet
Site Vice President
Sequoyah Nuclear Plant
Electronic Mail Distribution
General Counsel
Tennessee Valley Authority
Electronic Mail Distribution
John C. Fornicola, General Manager
Nuclear Assurance
Tennessee Valley Authority
Electronic Mail Distribution
Glenn W. Morris, Manager
Licensing and Industry Affairs
Sequoyah Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Beth A. Wetzel, Manager
Corporate Nuclear Licensing and
Industry Affairs
Tennessee Valley Authority
4X Blue Ridge
1101 Market Street
Chattanooga, TN 37402-2801
Robert H. Bryan, Jr., General Manager
Licensing and Industry Affairs
Sequoyah Nuclear Plant
Tennessee Valley Authority
4X Blue Ridge
1101 Market Street
Chattanooga, TN 37402-2801
David A. Kulisek, Plant Manager
Sequoyah Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Lawrence E. Nanney, Director
TN Dept. of Environment & Conservation
Division of Radiological Health
Electronic Mail Distribution
County Mayor
Hamilton County Courthouse
Chattanooga, TN 37402-2801
Ann Harris
341 Swing Loop
Rockwood, TN 37854
James H. Bassham, Director
Tennessee Emergency Management
Agency
Electronic Mail Distribution
4
Letter to Karl W. Singer from Malcolm T. Widmann dated February 22, 2007
SUBJECT:
ERRATA: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000327/2006005, 05000328/2006005 AND 07200034/2006002
Distribution w/encl:
Bob Pascarelli, NRR
B. Moroney, NRR
C. Evans (Part 72 Only)
L. Slack, RII EICS
OE Mail (email address if applicable)
RIDSNRRDIRS
PUBLIC
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R01
Adverse Weather Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R02
Evaluations of Changes, Tests or Experiments . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R04
Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R05
Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R07
Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R08
Inservice Inspection (ISI) Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R11
Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R12
Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1R13
Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . . . 12
1R15
Operability Evaluations
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R17
Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
1R19
Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1R20
Refueling and Other Outage Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1R22
Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
1EP6
Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
2OS1 Access Control To Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 20
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
4OA2 Identification & Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA6 Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
4OA7 Licensee Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
ATTACHMENT: SUPPLEMENTARY INFORMATION
Key Points of Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
List of Items Opened, Closed, and Discussed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3
List of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14
Enclosure
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-327, 50-328,72-034
License Nos:
Report No:
05000327/2006005 and 05000328/2006005 and
07200034/2006002
Licensee:
Tennessee Valley Authority (TVA)
Facility:
Sequoyah Nuclear Plant
Location:
Sequoyah Access Road
Soddy-Daisy, TN 37379
Dates:
October 1, 2006 - December 31, 2006
Inspectors:
J. Baptist, Acting Senior Resident Inspector
J. Diaz-Velez, Health Physicist (Section 2OS1)
F. Ehrhardt, Operations Engineer (Section 1R11.2)
L. Lake, Reactor Inspector (Section 1R08)
G. Laska, Senior Operations Examiner (Section 1R11.3)
D. Mas-Penaranda, Reactor Inspector (Sections 1R02, 1R17)
E. Michel, Reactor Inspector (Section 4OA5.3)
B. Miller, Reactor Inspector (Sections 1R08, 4OA5.2)
R. Moore, Senior Reactor Inspector (Section 4OA5.3)
S. Rose, Senior Operations Engineer (Section 1R11.3)
C. Smith Senior Reactor Inspector (Sections 1R02, 1R17)
M. Speck, Resident Inspector
C. Stancil, Resident Inspector (Section 1EP6)
Approved by:
M. Widmann, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000327/2006005, IR 05000328/2006005; IR 07200034/2006002; 10/01/2006 -
12/31/2006; Sequoyah Nuclear Plant, Units 1 & 2; Licensed Operator Requalification
Program.
The report covered a three-month period of inspection by resident inspectors and
announced inspections by 10 regional inspectors and one resident inspector from
another site. One NRC-identified Green finding, which was also a non-cited violation,
was identified. 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.
A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 55.53,
Conditions of Licenses for failure to certify the qualifications and status of licensed
operators were current and valid prior to their resumption of license duties. Specific
aspects of the requalification program that were not valid included plant tours that were
not completed with another licensed operator and not completing all shift functions in
positions to which the individuals will be assigned. The licensee entered the finding into
the corrective action program as PER No.112004.
The finding is greater than minor because it is associated with the human performance
attribute of the Mitigating Systems Cornerstone that affects the cornerstone objective of
ensuring the availability, reliability, and capability of operators to respond to initiating
events to prevent undesirable consequences that could pose a potential risk to
operations. The finding was evaluated using the Operator Requalification Human
Performance Significance Determination Process. Under this SDP, record deficiencies
can be either minor or of very low safety significance (Green). This finding was
determined to be Green because it was related to the program for maintaining active
licenses and more than 20% of the records reviewed had deficiencies. (Section 1R11.3).
B.
Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee, was
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation and corrective
action are listed in Section 4OA7.
Enclosure
29
performance indicator data submitted to the NRC to ensure it accurately reflected the
performance history of these systems.
b.
Findings and Observations
No findings of significance were identified. The licensee accurately documented the
baseline planned unavailability hours, the actual unavailability hours and the actual
unreliability information for the MSPI systems. No significant errors in the reported data
were identified, which resulted in a change to the indicated index color. No significant
discrepancies were identified in the MSPI basis document which resulted in: (1) a
change to the system boundary, (2) an addition of a monitored component, or (3) a
change in the reported index color.
.5
Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review
a.
Inspection Scope
The inspectors reviewed the interim report for the INPO plant assessment report of
Sequoyah conducted in July 2006. The inspectors reviewed the report to ensure that
issues identified were consistent with the NRC perspectives of licensee performance
and if any significant safety issues were identified that required further NRC follow-up.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
.1
Exit Meeting Summary
On January 3, 2007, the resident inspectors presented the inspection results to
Mr. R. Douet and other members of his staff, who acknowledged the findings. The
inspectors asked the licensee whether any of the material examined during the
inspection should be considered proprietary. No proprietary information was identified.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
TS 6.8.1 requires that written procedures shall be established, implemented, and
maintained covering the activities recommended in Appendix A of Regulatory
Guide 1.33, Revision 2, February 1978. Contrary to this, on November 28, 2006,
an AUO improperly implemented 0-GO-13,Reactor Coolant System Drain and
Fill Operations, Revision 54, Appendix AC by mispositioning an RCS loop 4 drain
valve. This revealed itself through the subsequent transfer of RCS inventory to
the Reactor Coolant Drain Tank and lowering of RCS pressurizer level. The
30
Enclosure
error was promptly corrected by operations staff and the event was identified in
the licensees corrective action program as PER 115534. This finding is of very
low safety significance because it did not challenge RCS inventory control by
exceeding available makeup capacity.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
J. Adams, Boric Acid
D. Bodine, Chemistry/Environmental Manager
R. Bruno, Training Manager
R. Douet, Site Vice President
B. Dungan, Outage and Site Scheduling Manager
J. Epperson, Licensed Operator Requal Lead
J. Goulart, ISI
K. Jones, Site Engineering Manager
Z. Kitts, Licensing Engineer
D. Kulisek, Plant Manager
G. Morris, Licensing and Industry Affairs Manager
T. Niessen, Site Quality Manager
M. A. Palmer, Radiation Protection Manager
M. H. Palmer, Operations Manager
K. Parker, Maintenance and Modifications Manager
J. Proffitt, (Acting) Site Licensing Supervisor
J. Reisenbuechler, Operations Training Manager
R. Reynolds, Site Security Manager
N. Thomas, Licensing Engineer
S. Tuthill, Chemistry Operations Manager
J. Whitaker, ISI
K. Wilkes, Emergency Preparedness Manager
NRC personnel:
R. Bernhard, Region II, Senior Reactor Analyst
D. Pickett, Project Manager, Office of Nuclear Reactor Regulation
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000327,328/2006005-01
Failure to Certify Qualifications and Status
of Licensed Operators Were Current and
Valid (Section 1R11.3)
Opened
Appendix R Manual Isolation Valve Failure
to Close Within the Required Time text
(Section 1R15)
Closed
05000327,328/2515/169
TI
Mitigating Systems Performance Index
Verification (Section 4OA5.4)
Attachment
A-2
Discussed
05000327, 328/2515/150
TI
Reactor Pressure Vessel Head and Vessel
Head Penetration Nozzles (NRC Order EA-
03-009) - Unit 2 (Section 4OA5.2)
05000327, 328/2515/166
TI
Pressurized Water Reactor Containment
Sump Blockage (NRC Generic Letter 2004-
02) - Unit 2 Section 4OA5.3)
Attachment
LIST OF ACRONYMS
ANSI
American National Standards Institute
abnormal operating procedures
alternate repair criteria
American Society of Mechanical Engineers
anticipated transient without scram
auxiliary unit operator
BACC
boric acid corrosion control
BMV
bare metal visual
cooling charging pump
CCPIT
cooling charging pump injection tank
CFR
Code of Federal Regulations
CR
condition report
control rod drive mechanism
chemical volume control system
DCN
design change notice
EDY
effective degradation years
essential raw cooling water
ETSS
examination technique specifications sheet
flow control valve
functional evaluation
FOSAR
foreign object search and recovery
high radiation
HUT
holdup tank
Institute of Nuclear power Operations
independent spent fuel storage installation
inservice inspection
materials reliability program
mitigating systems performance index
non-cited violation
NRC
U.S. Nuclear Regulatory Commission
outer diameter stress corrosion cracking
OPDP
operations department procedure
publically available records
PER
problem evaluation report
PER
protective action recommendation
power-operated relief valve
primary water stress corrosion cracking
reactor coolant pump
radiation protection
Attachment
A-15
RPVH
rated thermal power
radiation work permit
refueling water storage tank
significance determination process
safety evaluation report
safety injection
surveillance instructions
structure, system, or component
turbine driven auxiliary feedwater pump
TI
temporary instruction
TS
technical specification
Tennessee Valley Authority
updated final safety analysis report
UHI
upper head injection
unresolved item
ultrasonic testing
work orders
Enclosure
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-327, 50-328,72-034
License Nos:
Report No:
05000327/2006005 and 05000328/2006005 and
07200034/2006002
Licensee:
Tennessee Valley Authority (TVA)
Facility:
Sequoyah Nuclear Plant
Location:
Sequoyah Access Road
Soddy-Daisy, TN 37379
Dates:
October 1, 2006 - December 31, 2006
Inspectors:
J. Baptist, Acting Senior Resident Inspector
J. Diaz-Velez, Health Physicist (Section 2OS1)
F. Ehrhardt, Operations Engineer (Section 1R11.2)
L. Lake, Reactor Inspector (Section 1R08)
G. Laska, Senior Operations Examiner (Section 1R11.3)
D. Mas-Penaranda, Reactor Inspector (Sections 1R02, 1R17)
E. Michel, Reactor Inspector (Section 4OA5.3)
B. Miller, Reactor Inspector (Sections 1R08, 4OA5.2)
R. Moore, Senior Reactor Inspector (Section 4OA5.3)
S. Rose, Senior Operations Engineer (Section 1R11.3)
R. Schin, Senior Reactor Inspector (Sections 4OA5.5 - 4OA5.8)
C. Smith Senior Reactor Inspector (Sections 1R02, 1R17)
M. Speck, Resident Inspector
C. Stancil, Resident Inspector (Section 1EP6)
Approved by:
M. Widmann, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2a
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R01
Adverse Weather Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R02
Evaluations of Changes, Tests or Experiments . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R04
Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R05
Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R07
Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R08
Inservice Inspection (ISI) Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R11
Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R12
Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1R13
Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . . . 12
1R15
Operability Evaluations
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R17
Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
1R19
Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1R20
Refueling and Other Outage Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1R22
Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
1EP6
Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
2OS1 Access Control To Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 20
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
4OA2 Identification & Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA6 Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
4OA7 Licensee Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
ATTACHMENT: SUPPLEMENTARY INFORMATION
Key Points of Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
List of Items Opened, Closed, and Discussed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3
List of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14
Enclosure
SUMMARY OF FINDINGS
IR 05000327/2006005, IR 05000328/2006005; IR 07200034/2006002; 10/01/2006 -
12/31/2006; Sequoyah Nuclear Plant, Units 1 & 2; Licensed Operator Requalification
Program.
The report covered a three-month period of inspection by resident inspectors and
announced inspections by 11 regional inspectors and one resident inspector from
another site. Four NRC-identified Green findings, which were also non-cited violations,
were identified. 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.
A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 55.53,
Conditions of Licenses for failure to certify the qualifications and status of licensed
operators were current and valid prior to their resumption of license duties. Specific
aspects of the requalification program that were not valid included plant tours that were
not completed with another licensed operator and not completing all shift functions in
positions to which the individuals will be assigned. The licensee entered the finding into
the corrective action program as PER No.112004.
The finding is greater than minor because it is associated with the human performance
attribute of the Mitigating Systems Cornerstone that affects the cornerstone objective of
ensuring the availability, reliability, and capability of operators to respond to initiating
events to prevent undesirable consequences that could pose a potential risk to
operations. The finding was evaluated using the Operator Requalification Human
Performance Significance Determination Process. Under this SDP, record deficiencies
can be either minor or of very low safety significance (Green). This finding was
determined to be Green because it was related to the program for maintaining active
licenses and more than 20% of the records reviewed had deficiencies. (Section 1R11.3).
Green. The inspectors identified a non-cited violation of Unit 1 License Condition 16,
Fire Protection, and Unit 2 License Condition 13, Fire Protection, for failure to protect
certain equipment that was required for safe shutdown from fire damage. The
licensees Safe Shutdown Analysis for a fire in the Unit 1 480V Board Room 1B (Fire
Area FAA-095) relied on the fire not damaging at least two of the three Unit 1 battery
chargers located in the room plus one of the two Unit 1 inverters and one of the two
Unit 2 inverters located in the room. However, the battery chargers and inverters were
not separated or protected from fire damage as required by the License Conditions and
Fire Protection Program. The licensee entered the issue into the corrective action
program.
2b
Enclosure
This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the
availability and reliability of systems that mitigate initiating events to prevent undesirable
consequences and also involved a lack of required fire barriers or separation for
equipment relied upon for safe shutdown following a fire. The finding is of very low
safety significance because of the low frequency of fires that could damage two of the
three Unit 1 battery chargers, both Unit 1 inverters, or both Unit 2 inverters that were
located in the Unit 1 480V Board Room 1B concurrent with a failure of the sprinkler
system. (Section 4OA5.5)
Green. The inspectors identified a non-cited violation of Unit 1 License Condition 16,
Fire Protection, and Unit 2 License Condition 13, Fire Protection, for failure to protect
certain electrical cables for safe shutdown equipment from fire damage. The power
cables to Unit 1 vital inverter 1-II and Unit 2 vital inverter 2-II were routed through the
north end of the Unit 1 480V Board Room 1B (Fire Area FAA-095) without protection or
separation from fire damage as required by the License Conditions and Fire Protection
Program. The licensee entered the issue into the corrective action program and revised
the fire procedure to add local manual operator actions to mitigate the effects of fire
damage to the cables of concern.
This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the
availability and reliability of systems that mitigate initiating events to prevent undesirable
consequences and also involved a lack of required fire barriers or separation for
equipment relied upon for safe shutdown following a fire. The finding is of very low
safety significance because of the low frequency of fires that could damage the cables
of concern and also damage the redundant safe shutdown equipment. (Section
4OA5.6)
Green. The inspectors identified a non-cited violation of Unit 2 License Condition 13,
Fire Protection, for failure to maintain adequate lighting in the Unit 2 main steam valve
vault room to support time-critical operator actions required for post-fire safe shutdown.
The licensee entered the issue into the corrective action program and replaced the light
bulbs to restore the room lighting.
This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. It affected the
availability and reliability of systems that mitigate initiating events to prevent undesirable
consequences. The finding is of very low safety significance because of the low
frequency of fires that could lead to core damage if the operator actions in the Unit 2
main steam valve vault room were not performed in a timely manner. (Section 4OA5.7)
B.
Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee, was
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. The violation and corrective
actions are listed in Section 4OA7.
29
Enclosure
performance indicator data submitted to the NRC to ensure it accurately reflected the
performance history of these systems.
b.
Findings and Observations
No findings of significance were identified. The licensee accurately documented the
baseline planned unavailability hours, the actual unavailability hours and the actual
unreliability information for the MSPI systems. No significant errors in the reported data
were identified, which resulted in a change to the indicated index color. No significant
discrepancies were identified in the MSPI basis document which resulted in: (1) a
change to the system boundary, (2) an addition of a monitored component, or (3) a
change in the reported index color.
.5
(Closed) Unresolved Item (URI) 05000327,328/2005011-01, Reliance on 20-foot
Separation Zones for Fire Protection in Unit 1 480V Board Room 1B
a.
Inspection Scope
This in-office review followed up on URI 05000327,328/2005011-01, which had been
opened for NRC review of the licensing basis regarding use of 20-foot separation zones,
as specified in Appendix R,Section III.G.2 of 10 CFR 50, to protect safe shutdown
equipment from fire damage and the potential for the identified condition to adversely
affect safe shutdown.
b.
Findings
Introduction. A Green non-cited violation (NCV) of Unit 1 License Condition 16, Fire
Protection, and Unit 2 License Condition 13, Fire Protection, was identified for failure to
protect certain equipment that was required for safe shutdown from fire damage. The
licensees Safe Shutdown Analysis for a fire in the Unit 1 480V Board Room 1B (Fire
Area FAA-095) relied on the fire not damaging at least two of the three Unit 1 battery
chargers located in the room plus one of the two Unit 1 inverters and one of the two
Unit 2 inverters located in the room. However, the battery chargers and inverters were
not separated or protected from fire damage as required by the License Conditions and
Description. As described in Inspection Report (IR) 05000327,328/2005011, the NRC
had identified that the battery chargers and inverters in the Unit 1 480V Board Room 1B
(Fire Area FAA-095) were not separated or protected from fire damage as required by
10 CFR 50, Appendix R, Section III.G.2. One method prescribed by III.G.2 was
separation of equipment of redundant trains by a horizontal distance of more than 20
feet with no intervening combustibles or fire hazards. In addition, III.G.2 required that
fire detectors and an automatic fire suppression system be installed in the fire area.
The licensee had relied on 20-foot separation zones between each of the three Unit 1
battery chargers located in the room, a 20-foot separation zone between the two Unit 1
inverters located in the room, and a 20-foot separation zone between the two Unit 2
inverters located in the room. However, each 20-foot separation zone was not free of
intervening combustibles or fire hazards as required in that each 20-foot zone contained
30
Enclosure
energized 480V motor control centers (MCCs), nonqualified electrical cables in open
trays, and other electrical equipment including inverters. The MCCs, inverters, and non-
qualified cables in trays represented both ignition sources (fire hazards) and
combustibles in the form of insulated wires.
IR 05000327,328/2005011 also described the NRC-approved Deviation #11 to 10 CFR 50, Appendix R, Section III.G.2, regarding 20-foot separation zones in the auxiliary
building. (The Unit 1 480V Board Room 1B was located in the auxiliary building.)
Deviation #11 allowed 20-foot separation zones with intervening combustibles in the
form of cable trays provided that: 1) the cables had fuse and breaker coordination to
minimize the potential for fires initiating from cable faults and 2) extra sprinklers were
installed to compensate for cable trays partially blocking any sprinklers. The electrical
cables that were in open trays in the 20-foot separation zones in Unit 1 480V Board
Room 1B had sprinklers installed above and alongside them and the cables had fuse
and breaker coordination. The 480V MCCs that were in the 20-foot separation zones
also had sprinklers installed above them; however, the MCCs were not included in an
approved Deviation. Also, the MCCs represented much more significant ignition
sources (fire hazards) than the cable trays. In addition, some of the other electrical
equipment that was in one 20-foot separation zone (inverters in the 20-foot zone
between battery chargers on the south end of the room ) had no sprinklers above them.
After further in-office review of the licensing basis, the inspectors determined that strict
compliance with Appendix R to 10 CFR 50 is not a current requirement for Sequoyah
Units 1 and 2. Appendix R states that it applies to licensed nuclear power electric
generating stations that were operating prior to January 1, 1979. However, Sequoyah
Units 1 and 2 received their operating licenses after January 1, 1979. Prior to 1997, the
Sequoyah Unit 1 and Unit 2 License Conditions for Fire Protection had required that
TVA shall comply with Sections III.G, III.J, III.l, and III.O of Appendix R of 10 CFR 50,
except where the NRC has approved deviations. However, the Unit 1 and Unit 2
License Conditions for Fire Protection were changed in 1997 to no longer specifically
require compliance with Appendix R. The current License Conditions for Fire Protection
allow the licensee to make changes to the fire protection program if the changes do not
adversely affect post-fire safe shutdown.
During the inspection that is documented in IR 05000327,328/2005011, licensee
engineers had written an evaluation stating that the presence of MCCs and inverters in
the 20-foot separation zones in Unit 1 480V Board Room 1B did not adversely affect
safe shutdown and were acceptable as installed because there were sprinklers above
the cable trays and MCCs. However, after further in-office and onsite review, the
inspectors determined that the arrangement of MCCs with open cable trays directly
above them in the room created the potential for a fire initiating in an MCC section to
quickly involve cable trays, grow large enough to damage all of the equipment in the
room, and consequently to adversely affect safe shutdown. While there were sprinklers
above the MCCs and cable trays that could potentially extinguish a fire before it became
large, they were in a cross-zone preaction-type system that had a potential to fail. The
sprinkler piping was normally dry. Supplying water into the sprinkler piping involved
activation of at least two smoke detectors from different zones in the room and then
automatic opening of a valve in the fire water system. If the cross-zone detector circuit
31
Enclosure
failed or the automatic valve failed, then all of the sprinklers in the room would fail to
deliver water. The inspectors determined that the presence of 480V MCCs and
inverters (with open cable trays above them) in the 20-foot separation zones did not
comply with the approved Fire Protection Program and that this nonconforming
condition did adversely affect safe shutdown. Consequently, this condition represented
a violation of the Unit 1 and Unit 2 License Conditions for Fire Protection. When
informed of this determination, the licensee promptly entered the condition into the
corrective action program in Problem Evaluation Report (PER) 116718.
Analysis. This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. The finding
affected the availability and reliability of systems that mitigate initiating events to prevent
undesirable consequences and also involved a lack of required fire barriers for
equipment relied upon for safe shutdown following a fire. The finding is of very low
significance because of the low frequency of fires that could quickly grow large enough
to damage all of the equipment in the room, concurrent with a failure of the sprinkler
system.
The finding affected fire protection, so the Fire Protection Significance Determination
Process (SDP) (NRC Manual Chapter 0609, Appendix F) analysis was used. Because
the finding affected post-fire safe shutdown, represented a high degradation, and had a
duration of more than 30 days, the Fire Protection SDP Phase 1 analysis screened to
Phase 2. In the Phase 2 analysis, the same fire scenarios that affected this finding also
affected the finding described in the following Section 4OA5.6, so they were analyzed
together. In the Phase 2 analysis, about 40 of the 480V motor control center (MCC)
vertical sections in the room with multiple open cable trays directly above them could
initiate a fire that could create a hot gas layer that could damage everything in the room
before the fire brigade would arrive, if the automatic sprinkler system failed. With a
sprinkler system failure probability of 0.05, the finding screened to greater than Green
and an SDP Phase 3 was needed. In the Phase 3 analysis, two NRC Senior Reactor
Analysts conducted onsite inspection of the physical arrangement of target cables and
ignition sources and used more advanced analytical methods than those used in the
SDP Phase 2 analysis. The SDP Phase 3 analysis concluded the finding was of very
low safety significance (Green) because of the low frequency of fires that could quickly
grow large enough to damage all of the equipment in the room, concurrent with a failure
of the sprinkler system.
Enforcement. The Unit 1 and Unit 2 License Conditions for Fire Protection (16 and 13,
respectively) require that TVA implement and maintain in effect all provisions of the
approved fire protection program referenced in the Sequoyah Nuclear Plants Final
Safety Analysis Report and as approved in NRC Safety Evaluation Reports (SERs),
including the SERs contained in NUREG-011, Supplement 1, and NUREG-1232,
Volume 2. The License Conditions also state that TVA may make changes to the
approved fire protection program without prior approval by the Commission only if those
changes would not adversely affect the ability to achieve and maintain safe shutdown in
the event of a fire.
The SERs in NUREG-011 and NUREG-1232 accepted the Sequoyah fire protection
program based on meeting the criteria of Appendix A to BTP 9.5-1 and Sections III.G,
III.J, III.l, and III.O of Appendix R. BTP 9.5-1 and Section III.G of Appendix R require
32
Enclosure
that where cables or equipment that could prevent operation or cause maloperation of
systems necessary to achieve and maintain hot shutdown conditions are located within
the same fire area outside of primary containment, the cables shall be separated from
circuits of redundant trains or protected from fire damage by one of three specified
means.
Contrary to the above requirements, the Unit 1 battery chargers, Unit 1 inverters, and
Unit 2 inverters in Unit 1 480V Board Room 1B (Fire Area FAA-095) were not separated
from circuits of redundant trains or protected from fire damage by one of the three
specified means and thus could adversely affect safe shutdown. These electrical
components that were relied on for safe shutdown during a fire in Unit 1 480V Board
Room 1B had been unprotected for many years. Because this failure to protect safe
shutdown components is of very low safety significance and has been entered into the
licensees corrective action program as PER 116718, this violation is being treated as an
NCV, consistent with Section VI.A of the NRC Enforcement Policy. It is identified as
NCV 05000327,328/2006005-03 Inadequate 20-foot Separation Zones for Fire
Protection in Unit 1 480V Board Room 1B . URI 05000327,328/2005011-01 is closed.
.6
(Closed) Unresolved Item (URI) 05000327,328/2005011-02, Unprotected Power Cables
to Vital Inverters in the Unit 1 480V Board Room 1B
a.
Inspection Scope
This in-office review followed up on URI 05000327,328/2005011-02, which had been
opened for NRC review of the licensing basis regarding use of local manual operator
actions instead of physical protection or separation of cables as required by 10 CFR 50,
Appendix R,Section III.G.2.
b.
Findings
Introduction. A Green NCV of Unit 1 License Condition 16, Fire Protection, and Unit 2
License Condition 13, Fire Protection, was identified for failure to protect certain
electrical cables for safe shutdown equipment from fire damage. The alternating current
(AC) power cables to Unit 1 vital inverter 1-II and Unit 2 vital inverter 2-II were routed
through the north end of the Unit 1 480V Board Room 1B (Fire Area FAA-095) without
protection or separation from fire damage as required by the License Conditions and
Description. As described in IR 05000327,328/2005011, the NRC had identified that the
licensee had failed to adequately protect the AC power cables to Unit 1 vital inverter 1-II
and Unit 2 vital inverter 2-II in the north end of the Unit 1 480V Board Room 1B (Fire
Area FAA-095) from fire damage. When informed of this condition, the licensee
promptly entered the issue into their corrective action program in PER 91841 and
revised the fire procedure to add local manual operator actions to mitigate the effects of
fire damage to the cables of concern. However, this licensee corrective action relied on
local manual operator actions instead of using physical protection or separation of the
cables as required by 10 CFR 50, Appendix R, Section III.G.2
After further review of the licensing basis, the inspectors determined that strict
compliance with Appendix R to 10 CFR 50 is not a current requirement for Sequoyah
33
Enclosure
Units 1 and 2. Appendix R states that it applies to licensed nuclear power electric
generating stations that were operating prior to January 1, 1979. However, Sequoyah
Units 1 and 2 received their operating licenses after January 1, 1979. Prior to 1997, the
Sequoyah Unit 1 and Unit 2 License Conditions for Fire Protection had required that
TVA shall comply with Sections III.G, III.J, III.l, and III.O of Appendix R of 10 CFR 50,
except where the NRC has approved deviations. However, the Unit 1 and Unit 2
License Conditions for Fire Protection were changed in 1997 to no longer specifically
require compliance with Appendix R. The current License Conditions for Fire Protection
allow the licensee to make changes to the fire protection program if the changes do not
adversely affect post-fire safe shutdown. Consequently, since the added local manual
operator actions did not adversely affect safe shutdown, the licensee could rely on them
as corrective action for the identified condition.
The inspectors determined that the licensees failure to protect the AC power cables to
Unit 1 vital inverter 1-II and Unit 2 vital inverter 2-II in the north end of the Unit 1 480V
Board Room 1B (Fire Area FAA-095) from fire damage was not in compliance with the
License Conditions for Fire Protection and the licensees approved fire protection
program, which included design criteria described in 10 CFR 50, Appendix R, Section
III.G.2 and NRC Branch Technical Position (BTP) 9.5-1. Further, this condition
adversely affected post-fire safe shutdown in that it created the potential for one fire to
damage equipment that was relied on for safe shutdown during that fire.
Analysis. This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. The finding
affected the availability and reliability of systems that mitigate initiating events to prevent
undesirable consequences and also involved a lack of required fire barriers for
equipment relied upon for safe shutdown following a fire. The finding is of very low
significance because of the low frequency of fires that could damage the cables of
concern and also damage the redundant safe shutdown equipment which is located in
the same fire area.
The finding affected fire protection, so the Fire Protection Significance Determination
Process (SDP) (NRC Manual Chapter 0609, Appendix F) analysis was used. Because
the finding affected post-fire safe shutdown, represented a high degradation, and had a
duration of more than 30 days, the Fire Protection SDP Phase 1 analysis screened to
Phase 2. In the Phase 2 analysis, the same fire scenarios that affected this finding also
affected the finding described in the above Section 4OA5.5, so they were analyzed
together. In the Phase 2 analysis, about 40 of the 480V motor control center (MCC)
vertical sections in the room with multiple open cable trays directly above them could
initiate a fire that could create a hot gas layer that could damage everything in the room
before the fire brigade would arrive, if the automatic sprinkler system failed. With a
sprinkler system failure probability of 0.05, the finding screened to greater than Green
and an SDP Phase 3 was needed. In the Phase 3 analysis, two NRC Senior Reactor
Analysts conducted onsite inspection of the physical arrangement of target cables and
ignition sources and used more advanced analytical methods than those used in the
SDP Phase 2 analysis. The SDP Phase 3 analysis concluded the finding was of very
low safety significance (Green) because of the low frequency of fires that could damage
the cables of concern and also damage the redundant safe shutdown equipment which
is located in the same fire area.
34
Enclosure
Enforcement. The Unit 1 and Unit 2 License Conditions for Fire Protection (16 and 13,
respectively) require that TVA implement and maintain in effect all provisions of the
approved fire protection program referenced in the Sequoyah Nuclear Plants Final
Safety Analysis Report and as approved in NRC Safety Evaluation Reports (SERs),
including the SERs contained in NUREG-011, Supplement 1, and NUREG-1232,
Volume 2. The License Conditions also state that TVA may make changes to the
approved fire protection program without prior approval by the Commission only if those
changes would not adversely affect the ability to achieve and maintain safe shutdown in
the event of a fire.
The SERs in NUREG-011 and NUREG-1232 accepted the Sequoyah fire protection
program based on meeting the criteria of Appendix A to BTP 9.5-1 and Sections III.G,
III.J, III.l, and III.O of Appendix R. BTP 9.5-1 and Section III.G of Appendix R require
that where cables or equipment that could prevent operation or cause maloperation of
systems necessary to achieve and maintain hot shutdown conditions are located within
the same fire area outside of primary containment, the cables shall be separated from
circuits of redundant trains or protected from fire damage by one of three specified
means.
Contrary to the above requirements, the AC power cables to Unit 1 vital inverter 1-II and
Unit 2 vital inverter 2-II in the north end of the Unit 1 480V Board Room 1B (Fire Area
FAA-095) were not separated from circuits of redundant trains or protected from fire
damage by one of the three specified means and thus could adversely affect safe
shutdown. These cables had been unprotected for several years. Because this failure
to protect cables is of very low safety significance and has been entered into the
licensees corrective action program as PER 91841, this violation is being treated as an
NCV, consistent with Section VI.A of the NRC Enforcement Policy. It is identified as
NCV 05000327,328/2006005-04, Unprotected Power Cables to Vital Inverters in the
Unit 1 480V Board Room 1B. URI 05000327,328/2005011-02 is closed.
.7
(Closed) Unresolved Item (URI) 05000327,328/2005011-04, Appendix R Operator
Action to Throttle AFW in Main Steam Valve Vault Room
a.
Inspection Scope
This in-office review followed up on URI 05000327,328/2005011-04, which had been
opened for NRC review of the licensing basis for the post-fire operator action to throttle
AFW flow in the Unit 1 and 2 main steam valve vault rooms.
b.
Findings
Introduction. A Green NCV of Unit 2 License Condition 13, Fire Protection, was
identified for failure to maintain lighting in the Unit 2 main steam valve vault room. The
lighting was needed to support the post-fire time-critical operator action to throttle AFW
flow in the room.
Description. As described in IR 05000327,328/2005011, the NRC had identified that the
Unit 2 main steam valve vault room was completely dark. All of the normal lights were
out because the light bulbs were burned out and the installed Appendix R emergency
lights were off because normal power was available. When informed of this condition,
35
Enclosure
the licensee had promptly entered the issue into their corrective action program in PER
91899 and replaced the light bulbs to restore the normal lighting. The inspectors had
determined that the operator actions in the Unit 2 main steam valve vault room were
feasible with lighting, but were not feasible for one operator to reliably accomplish in
complete darkness. Additionally, the inspectors had questioned the acceptability of the
licensees reliance on the local manual operator actions without obtaining NRC approval
for a Deviation from the requirements of 10 CFR 50, Appendix R, Section III.G.2.
After further review of the licensing basis, the inspectors determined that strict
compliance with Appendix R to 10 CFR 50 was not a current requirement for Sequoyah
Units 1 and 2. Appendix R states that it applies to licensed nuclear power electric
generating stations that were operating prior to January 1, 1979. However, Sequoyah
Units 1 and 2 received their operating licenses after January 1, 1979. Prior to 1997, the
Sequoyah Unit 1 and Unit 2 License Conditions for Fire Protection had required that
TVA shall comply with Sections III.G, III.J, III.l, and III.O of Appendix R of 10 CFR 50,
except where the NRC has approved deviations. However, the Unit 1 and Unit 2
License Conditions for Fire Protection were changed in 1997 to no longer specifically
require compliance with Appendix R. The current License Conditions for Fire Protection
allow the licensee to make changes to the fire protection program if the changes do not
adversely affect post-fire safe shutdown. Consequently, if the local manual operator
actions in the main steam valve vault room were feasible and reliable, then they would
not adversely affect safe shutdown and the licensee could rely on them without needing
NRC review and approval.
The time-critical local manual actions to throttle AFW in the main steam valve vault room
were required in AOP-N.08, Appendix R Fire Safe Shutdown, Rev. 7, and in AOP-C.04,
Shutdown From Auxiliary Control Room, Rev. 8. The inspectors had determined that
the manual operator actions could be considered feasible if the installed lighting was
working. However, with no installed lights working, the manual operator actions were
not determined to be feasible. Consequently, the licensees failure to maintain the
lighting to support the operator actions created a condition that could adversely affect
safe shutdown during certain fires.
After the licensee replaced the light bulbs in the Unit 2 main steam valve vault room,
then both the Unit 2 and Unit 1 main steam valve vault rooms were lighted. With the
rooms lighted, the inspectors considered that the manual operator actions in the rooms
were feasible and would not adversely affect safe shutdown.
Analysis. This finding is of greater than minor safety significance because it affected the
objectives of the Mitigating Systems Cornerstone of Reactor Safety. The finding
affected the availability and reliability of systems that mitigate initiating events to prevent
undesirable consequences and also involved a lack of required fire protection for
equipment relied upon for safe shutdown following a fire. The finding is of very low
safety significance because of the low frequency of fires that could lead to core damage
if the operator actions in the Unit 2 main steam valve vault room were not performed in a
timely manner.
The finding affected fire protection, so the Fire Protection Significance Determination
Process (SDP) (NRC Manual Chapter 0609, Appendix F) analysis was used. Because
the finding affected post-fire safe shutdown, represented a high degradation, and had a
36
Enclosure
duration of more than 30 days, the Fire Protection SDP Phase 1 analysis screened to
Phase 2. In the Phase 2 analysis, because licensee mitigation of a fire in almost every
area of the plant involved reliance on a manual action to throttle AFW in the Unit 2 main
steam valve vault room, and no credit was given for the manual action, the finding did
not screen to Green and an SDP Phase 3 analysis was needed.
A regional Senior Reactor Analyst performed the Phase 3 analysis, including onsite
inspection and consideration of fires initiating in all areas of the plant. The Phase 3
analysis determined that the mitigation of fires in some areas of the plant relied on local
manual throttling of the motor driven auxiliary feedwater pump flow in the main steam
valve vault room; however, delay in performing that action would not lead to core
damage. Fires in other areas of the plant relied on local manual throttling of turbine
driven auxiliary feedwater pump flow in the main steam valve vault room; however, the
frequency of those fires was low. Also, the existence of Flamastic 77 (a flame spread
retardant) would slow the growth of fires in those areas. The SDP Phase 3 analysis
concluded the finding was of very low safety significance (Green) because of the low
frequency of fires that could lead to core damage if the operator actions in the Unit 2
main steam valve vault room were not performed in a timely manner.
Enforcement. The Unit 2 License Condition 13 for Fire Protection requires that TVA
implement and maintain in effect all provisions of the approved fire protection program
referenced in the Sequoyah Nuclear Plants Final Safety Analysis Report and as
approved in NRC SERs, including the SERs contained in NUREG-011, Supplement 1,
and NUREG-1232, Volume 2. The License Condition also states that TVA may make
changes to the approved fire protection program without prior approval by the
Commission only if those changes would not adversely affect the ability to achieve and
maintain safe shutdown in the event of a fire.
The fire protection program included the safe shutdown methodology that relied on the
time-critical local manual operator actions to throttle emergency feedwater in the main
steam valve vault room. Those local manual operator actions had been in place for
many years and had been documented in NRC Inspection Report 05000327,328/1988-
024, which was referenced in the SER in NUREG-1232, Volume 2.
Contrary to the above requirements, the licensee did not maintain the lighting in the Unit
2 main steam valve vault room to support the time-critical local manual operator actions
in that room. Because this failure to maintain room lighting is of very low safety
significance and has been entered into the licensees corrective action program as PER
91899, this violation is being treated as an NCV, consistent with Section VI.A of the
NRC Enforcement Policy. It is identified as NCV 05000328/2006005-05, Failure to
Maintain Lighting for Time-Critical Local Manual Actions for Post-Fire Safe Shutdown.
URI 05000327,328/2005011-04 is closed.
.8
(Closed) Unresolved Item (URI) 05000327,328/2005011-05, Reliance on Local Manual
Operator Actions for Appendix R Fires
a.
Inspection Scope
This in-office review followed up on URI 05000327,328/2005011-05, which had been
opened for NRC review of the licensing basis related to reliance on local manual
37
Enclosure
operator actions that had not been specifically approved by the NRC for mitigating 10 CFR 50, Appendix R, Section III.G.2 fires.
b.
Findings
Introduction. The inspectors determined that licensee reliance on local manual operator
actions, that had not been approved by the NRC for mitigating 10 CFR 50, Appendix R,
Section III.G.2 fires, was not prohibited by the licensing basis. Consequently, this URI
did not represent a finding.
Description. As described in IR 05000327,328/2005011, the NRC had identified that the
licensee relied on many local manual operator actions, that had not been approved by
the NRC as a Deviation, to mitigate 10 CFR 50, Appendix R, Section III.G.2 fires.
After further review of the licensing basis, the inspectors determined that strict
compliance with Appendix R to 10 CFR 50 was not a current requirement for Sequoyah
Units 1 and 2. Appendix R states that it applies to licensed nuclear power electric
generating stations that were operating prior to January 1, 1979. However, Sequoyah
Units 1 and 2 received their operating licenses after January 1, 1979. Prior to 1997, the
Sequoyah Unit 1 and Unit 2 License Conditions for Fire Protection had required that
TVA shall comply with Sections III.G, III.J, III.l, and III.O of Appendix R of 10 CFR 50,
except where the NRC has approved deviations. However, the Unit 1 and Unit 2
License Conditions for Fire Protection were changed in 1997 to no longer specifically
require compliance with Appendix R. The current License Conditions for Fire Protection
allow the licensee to make changes to the fire protection program if the changes do not
adversely affect post-fire safe shutdown. Consequently, if the local manual operator
actions were feasible and reliable, then they would not adversely affect safe shutdown
and the licensee could rely on them without needing NRC review and approval.
With the exception of the action to locally control AFW pump flow in the Unit 2 main
steam valve vault room (described above in Section 4OA5.7), the inspectors had found
that the local manual actions that were reviewed were all feasible. Therefore, the
licensee could rely on them without obtaining NRC review and approval. URI
05000327,328/2005011-05 is closed.
.9
Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review
a.
Inspection Scope
The inspectors reviewed the interim report for the INPO plant assessment report of
Sequoyah conducted in July 2006. The inspectors reviewed the report to ensure that
issues identified were consistent with the NRC perspectives of licensee performance
and if any significant safety issues were identified that required further NRC follow-up.
b. Findings
No findings of significance were identified.
38
Enclosure
4OA6 Meetings, Including Exit
Exit Meeting Summary
On January 3, 2007, the resident inspectors presented the inspection results to
Mr. R. Douet and other members of his staff, who acknowledged the findings. The
inspectors asked the licensee whether any of the material examined during the
inspection should be considered proprietary. No proprietary information was identified.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
TS 6.8.1 requires that written procedures shall be established, implemented, and
maintained covering the activities recommended in Appendix A of Regulatory
Guide 1.33, Revision 2, February 1978. Contrary to this, on November 28, 2006,
an AUO improperly implemented 0-GO-13,Reactor Coolant System Drain and
Fill Operations, Revision 54, Appendix AC by mispositioning an RCS loop 4 drain
valve. This revealed itself through the subsequent transfer of RCS inventory to
the Reactor Coolant Drain Tank and lowering of RCS pressurizer level. The
error was promptly corrected by operations staff and the event was identified in
the licensees corrective action program as PER 115534. This finding is of very
low safety significance because it did not challenge RCS inventory control by
exceeding available makeup capacity.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
J. Adams, Boric Acid
D. Bodine, Chemistry/Environmental Manager
R. Bruno, Training Manager
R. Douet, Site Vice President
B. Dungan, Outage and Site Scheduling Manager
J. Epperson, Licensed Operator Requal Lead
J. Goulart, ISI
K. Jones, Site Engineering Manager
Z. Kitts, Licensing Engineer
D. Kulisek, Plant Manager
G. Morris, Licensing and Industry Affairs Manager
T. Niessen, Site Quality Manager
M. A. Palmer, Radiation Protection Manager
M. H. Palmer, Operations Manager
K. Parker, Maintenance and Modifications Manager
J. Proffitt, (Acting) Site Licensing Supervisor
J. Reisenbuechler, Operations Training Manager
R. Reynolds, Site Security Manager
N. Thomas, Licensing Engineer
S. Tuthill, Chemistry Operations Manager
J. Whitaker, ISI
K. Wilkes, Emergency Preparedness Manager
NRC personnel:
R. Bernhard, Region II, Senior Reactor Analyst
D. Pickett, Project Manager, Office of Nuclear Reactor Regulation
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Inability to Perform Required Actions of
AOP-N.08, Appendix R Fire Safe Shutdown
(Section 1R15)
Opened and Closed
05000327,328/2006005-01
Failure to Certify Qualifications and Status
of Licensed Operators Were Current and
Valid (Section 1R11.3)
A-2
Attachment
05000327,328/2006005-03
Inadequate 20-foot Separation Zones for
Fire Protection in Unit 1 480V Board Room
1B (Section 4OA5.5)
05000327,328/2006005-04
Unprotected Power Cables to Vital Inverters
in the Unit 1 480V Board Room 1B (Section
4OA5.6)05000328/2006005-05
Failure to Maintain Lighting for Time-Critical
Local Manual Actions for Post-Fire Safe
Shutdown (Section 4OA5.7)
Closed
05000327,328/2515/169
TI
Mitigating Systems Performance Index
Verification (Section 4OA5.4)
05000327,328/2005011-01
Reliance on 20-foot Separation Zones for
Fire Protection in Unit 1 480V Board Room
1B (Section 4OA5.5)
05000327,328/2005011-02
Unprotected Power Cables to Vital Inverters
in the Unit 1 480V Board Room 1B (Section
4OA5.6)
05000327,328/2005011-04
Appendix R Operator Action to Throttle
AFW in Main Steam Valve Vault Room
(Section 4OA5.7)
05000327,328/2005011-05
Reliance on Local Manual Operator Actions
for Appendix R Fires (Section 4OA5.8)
Discussed
05000327, 328/2515/150
TI
Reactor Pressure Vessel Head and Vessel
Head Penetration Nozzles (NRC Order EA-
03-009) - Unit 2 (Section 4OA5.2)
05000327, 328/2515/166
TI
Pressurized Water Reactor Containment
Sump Blockage (NRC Generic Letter 2004-
02) - Unit 2 Section 4OA5.3)
Attachment
LIST OF ACRONYMS
alternating currrent
ANSI
American National Standards Institute
abnormal operating procedures
alternate repair criteria
American Society of Mechanical Engineers
anticipated transient without scram
auxiliary unit operator
BACC
boric acid corrosion control
BMV
bare metal visual
cooling charging pump
CCPIT
cooling charging pump injection tank
CFR
Code of Federal Regulations
CR
condition report
control rod drive mechanism
chemical volume control system
DCN
design change notice
EDY
effective degradation years
essential raw cooling water
ETSS
examination technique specifications sheet
flow control valve
functional evaluation
FOSAR
foreign object search and recovery
high radiation
HUT
holdup tank
Institute of Nuclear power Operations
IR
inspection report
independent spent fuel storage installation
inservice inspection
motor control center
materials reliability program
mitigating systems performance index
non-cited violation
NRC
U.S. Nuclear Regulatory Commission
outer diameter stress corrosion cracking
OPDP
operations department procedure
publically available records
PER
problem evaluation report
PER
protective action recommendation
power-operated relief valve
primary water stress corrosion cracking
reactor coolant pump
A-15
Attachment
radiation protection
RPVH
rated thermal power
radiation work permit
refueling water storage tank
significance determination process
safety evaluation report
safety injection
surveillance instructions
structure, system, or component
turbine driven auxiliary feedwater pump
TI
temporary instruction
TS
technical specification
Tennessee Valley Authority
updated final safety analysis report
UHI
upper head injection
unresolved item
ultrasonic testing
work orders