ML101240946
| ML101240946 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/04/2010 |
| From: | Ryan Lantz NRC/RGN-IV/DRP/RPB-D |
| To: | Ridenoure R Southern California Edison Co |
| References | |
| FOIA/PA-2011-0221, FOIA/PA-2011-0157 IR-10-002 | |
| Download: ML101240946 (77) | |
See also: IR 05000361/2010002
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
R E GI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
May 4, 2010
Mr. Ross T. Ridenoure
Senior Vice President and
Chief Nuclear Officer
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION - NRC INTEGRATED
INSPECTION REPORT 05000361/2010002 and 05000362/2010002
Dear Mr. Ridenoure:
On March 24, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your San Onofre Nuclear Generating Station, Units 2 and 3 facility. The enclosed integrated
inspection report documents the inspection findings, which were discussed on March 23, 2010,
with you, and other members of your staff.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents 12 NRC identified findings and two self-revealing findings of very low
safety significance (Green). All of these findings were determined to involve violations of NRC
requirements. Additionally, three licensee-identified violations, which were determined to be of
very low safety significance, are listed in this report. However, because of the very low safety
significance and because they are entered into your corrective action program, the NRC is
treating these findings as noncited violations, consistent with Section VI.A.1 of the NRC
Enforcement Policy. If you contest the violations or the significance of the noncited violations,
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, U.S. Nuclear
Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas,
76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, D.C. 20555-0001; and the NRC Resident Inspector at the San Onofre Nuclear
Generating Station facility. In addition, if you disagree with the characterization of any finding in
this report, you should provide a response within 30 days of the date of this inspection report,
with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC
Southern California Edison Company
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Resident Inspector at San Onofre Nuclear Generating Station. The information you provide will
be considered in accordance with Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letter, and its
enclosure, will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records component of 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).
Sincerely,
/RA/ Donald B. Allen for
Ryan E. Lantz, Chief
Project Branch D
Division of Reactor Projects
Docket Nos.
50-361; 50-362
Enclosure:
NRC Inspection Report 05000361/2010002 and 05000362/2010002
w/Attachment: Supplemental Information
cc w/Enclosure:
Chairman, Board of Supervisors
County of San Diego
1600 Pacific Highway, Room 335
San Diego, CA 92101
Gary L. Nolff
Assistant Director-Resources
City of Riverside
3900 Main Street
Riverside, CA 92522
Mark L. Parsons
Deputy City Attorney
City of Riverside
3900 Main Street
Riverside, CA 92522
Southern California Edison Company
- 3 -
Gary H. Yamamoto, P.E., Chief
Division of Drinking Water and
Environmental Management
1616 Capitol Avenue, MS 7400
P.O. Box 997377
Sacramento, CA 95899-7377
Michael L. DeMarco
San Onofre Liaison
San Diego Gas & Electric Company
8315 Century Park Ct. CP21C
San Diego, CA 92123-1548
Director, Radiological Health Branch
State Department of Health Services
P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414
The Mayor of the City of San Clemente
100 Avenida Presidio
San Clemente, CA 92672
James D. Boyd, Commissioner
California Energy Commission
1516 Ninth Street (MS 34)
Sacramento, CA 95814
Douglas K. Porter, Esquire
Southern California Edison Company
2244 Walnut Grove Avenue
Rosemead, CA 91770
Albert R. Hochevar
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Steve Hsu
Department of Health Services
Radiologic Health Branch
MS 7610, P.O. Box 997414
Sacramento, CA 95899-7414
Southern California Edison Company
- 4 -
R. St. Onge
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Chief, Technological Hazards Branch
FEMA Region IX
1111 Broadway, Suite 1200
Oakland, CA 94607-4052
Southern California Edison Company
- 5 -
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Chuck.Casto@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRP Deputy Director (Anton.Vegel@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Greg.Warnick@nrc.gov)
Resident Inspector (John.Reynoso@nrc.gov)
Branch Chief, DRP/D (Ryan.Lantz@nrc.gov)
Senior Project Engineer, DRP/D (Don.Allen@nrc.gov)
SONGS Administrative Assistant (Heather.Hutchinson@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Randy.Hall@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
ROP Reports
DRS/TSB STA (Dale.Powers@nrc.gov)
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)
Regional State Liaison Officer (Bill.Maier@nrc.gov)
NSIR/DPR/EP (Eric.Schrader@nrc.gov)
File located: R:\\Reactors\\Songs\\2010\\SO2010-002-RP-GGW.doc
SUNSI Rev Compl.
- Yes No
- Yes No
Reviewer Initials
Publicly Avail
- Yes No
Sensitive
Yes ; No
Sens. Type Initials
RI:DRP
RI:DRP
SRI:DRP
C:DRS/OB
C:DRS/EB1
JReynoso
JJosey
GWarnick
MHaire
TFarnholtz
/DAllen for E/ /DAllen for E/
/RA/
/DAllen for/
/RML for/
5/3 /10
5/3 /10
5/3/10
4/29/10
4/29/10
C:DRS/EB2
C:DRS/PSB1
C:DRS/PSB2
C:DRP
NO'Keefe
MShannon
GWerner
RLantz
/RA/
/JLarsen for/
/RA/
/DAllen for/
4/28/10
4/29/10
4/29/10
4/30/10
OFFICIAL RECORD COPY
T=Telephone
E=E-mail
F=Fax
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Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket:
50-361, 50-362
License:
Report:
05000361/2010002 and 05000362/2010002
Licensee:
Southern California Edison Co. (SCE)
Facility:
San Onofre Nuclear Generating Station, Units 2 and 3
Location:
5000 S. Pacific Coast Hwy
San Clemente, California
Dates:
January 1, 2010 through March 24, 2010
Inspectors:
D. Allen, Senior Project Engineer
P. Elkmann, Senior Emergency Preparedness Inspector
J. Josey, Resident Inspector
J. Reynoso, Resident Inspector
B. Rice, Reactor Engineer
W. Schaup, Project Engineer
G. Warnick, Senior Resident Inspector
Approved By:
Ryan E. Lantz
Chief, Project Branch D
Division of Reactor Projects
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Enclosure
SUMMARY OF FINDINGS
IR 05000361/2010002, 05000362/2010002; 01/01/2010 - 03/24/2010; San Onofre Nuclear
Generating Station, Units 2 & 3; Integ Resid & Reg Report; Fire Prot, Maint Effect, Maint Risk &
Em Work, Op Eval, Postmaint Test, Ref Outages, Id.& Res.of Prob, Event F/U
The report covered a 3-month period of inspection by resident inspectors and an announced
baseline inspection by a regional based inspector. Fourteen Green noncited violations of
significance were identified. The significance of most findings is indicated by their color (Green,
White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination
Process. Findings for which the significance determination process does not apply may be
Green or be assigned a severity level after NRC management review. The NRC's program for
overseeing the safe operation of commercial nuclear power reactors is described in NUREG-
1649, Reactor Oversight Process, Revision 4, dated December 2006.
A.
NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Initiating Events
Green. The inspectors identified three examples of a noncited violation of
Technical Specification 5.5.1.1.d, for the failure of contractor and station
personnel to properly implement the requirements of a station fire protection
procedure for control of hot work activities. Specifically, between January 4 and
March 17, 2010, three examples were identified where contractor and station
personnel failed to properly implement the requirements of procedure
SO123-XV-1.41, Control of Ignition Sources, Revision 14, Steps 6.2.1
and 6.4.1.3. Specifically, contractor and station personnel failed to ensure that
combustible materials were covered or removed from the ignition source.
Following the inspectors identification of each example, the licensee immediately
stopped the hot work activities and restored compliance with the requirements of
procedure SO123-XV-1.41. This issue was entered into the licensees corrective
action program as Nuclear Notifications NNs 200729747, 200746059 and
200835830.
The finding is greater than minor because if left uncorrected, the practice of
conducting hot work in a manner that allows uncontrolled combustibles to be
within the procedurally specified exclusion area would have the potential to lead
to a more significant safety concern, in that, it could result in a fire in or near risk
important equipment. The finding is associated with the Initiating Events
Cornerstone. The inspectors determined that Manual Chapter 0609, Appendix F,
Fire Protection Significance Determination Process, does not address the
potential risk significance of shutdown fire protection findings, and Appendix G,
Shutdown Operations Significance Determination Process, does not address
fire protection findings, and therefore could not be applied to shutdown plant
conditions. Because of this, the inspectors used Manual Chapter 0609,
Appendix M, Significance Determination Process Using Qualitative Criteria.
The NRC management review was performed by using the Manual Chapter
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Enclosure
0609, Appendix F, Phase 1 Worksheet, to establish a bounding analysis. Using
the bounding analysis, the finding is determined to have very low safety
significance because the finding represented a low degradation rating, in that, it
did not have any significant effect on the likelihood that a fire might occur, or that
a fire which does occur might not be promptly suppressed. This finding had a
crosscutting aspect in the area of human performance associated with work
practices, in that, the licensee failed to define and effectively communicate
expectations regarding procedural compliance and personnel following
procedures H.4(b) (Section 1R05).
Green. The inspectors identified a noncited violation of 10 CFR 50.65(a)(4),
Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants, involving multiple instances where operations and work control
personnel failed to adequately assess and implement appropriate risk
management activities. Specifically, between February 18, and February 23,
2010, operations and work control personnel failed to adequately assess and
manage the increase in risk associated with maintenance activities in the
electrical switchyard. Following the inspectors identification of the findings, the
licensee adequately assessed and managed the increase in risk for the
maintenance activities. This issue was entered into the licensees corrective
action program as Nuclear Notifications NNs 200801929 and 200805635.
The finding is greater than minor since it was similar to both more than minor
Examples 7.e and 7.f in NRC Inspection Manual Chapter 0612, Appendix E,
Examples of Minor Issues, because when the activities were correctly assessed
plant procedures required risk management actions to be taken. The finding is
associated with the Initiating Events Cornerstone. The inspectors determined
that the licensee does not maintain a shutdown probabilistic risk analysis model,
and as such, an incremental core damage probability cannot be estimated for the
plant conditions that existed at the time of the performance deficiency. For this
reason, the inspectors determined that Manual Chapter 0609, Appendix K,
Maintenance Risk Assessment and Risk Management Significance
Determination Process, Flowchart 2, could not be used to determine the risk
significance the finding. Using the qualitative review process of Manual Chapter
0609, Appendix M, Significance Determination Process Using Qualitative
Criteria, the finding is determined to have very low safety significance because
the finding did not result in any additional loss of defense in depth systems. This
finding has a crosscutting aspect in the area of human performance associated
with the work practices because the licensee failed to define and effectively
communicate expectations regarding procedural compliance and that personnel
follow procedures H.4(b) (Section 1R13).
Cornerstone: Mitigating Systems
Green. The inspectors identified a noncited violation of 10 CFR 50.65(b)(2)(ii) for
the licensees failure to appropriately scope the steam driven auxiliary feedwater
pump trench eductor in the maintenance rule monitoring program. Specifically,
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Enclosure
from the inception of the facilities monitoring program through March 2010, the
licensee failed to properly scope the steam drive auxiliary feedwater pump trench
educator. The eductors prevent water from accumulating in the trench because
water in contact with the pumps steam supply piping would cause condensation
of the steam in the pipe. Condensation would cause the turbine to over speed,
which would render the pump incapable of performing its specified safety
function. This issue was entered into the licensees corrective action program as
Nuclear Notification NN 200765185.
The finding is greater than minor because it is associated with the equipment
performance attribute of the Mitigating Systems Cornerstone and directly affected
the cornerstone objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Using the Manual Chapter 0609, Significance Determination Process, Phase 1
Worksheets, the finding is determined to have very low safety significance
because the finding: (1) is not a design or qualification issue confirmed not to
result in a loss of operability or functionality; (2) did not represent an actual loss
of safety function of the system or train; (3) did not result in the loss of one or
more trains of nontechnical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating
event. The inspectors determined that since the scoping of the systems had
occurred more than 2 years in the past, and the opportunity to reevaluate system
scoping had not occurred recently, that the finding did not represent current plant
performance and therefore did not have a crosscutting aspect associated with it
(Section 1R12).
Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the
licensees failure to properly implement procedure requirements to ensure that
applicable risk significant operating experience was entered into the corrective
action program for timely evaluation. Specifically, on December 17, 2009, the
operating experience review committee failed to properly implement the
requirements of procedure SO23-XV-40, Sharing Industry Information,
Revision 1. An industry operating experience report review determined the
operating experience was not applicable and was distributed as information only;
not requiring any action. The same industry operating experience was later
determined to be applicable by the probabilistic risk assessment group, and
interim compensatory measures were initiated on February 10, 2010, to address
the issues. This issue was entered into the licensees corrective action program
as Nuclear Notifications NN 200805879.
The finding is greater than minor because it is associated with the procedure
quality attribute of the Mitigating Systems Cornerstone and affects the associated
cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Using the Manual Chapter 0609, Significance Determination Process, Phase 1
Worksheets, the finding is determined to have very low safety significance
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Enclosure
because the finding: (1) is not a design or qualification issue confirmed not to
result in a loss of operability or functionality; (2) did not represent an actual loss
of safety function of the system or train; (3) did not result in the loss of one or
more trains of nontechnical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating
event. This finding has a crosscutting aspect in the area of human performance
associated with decision-making because the operating experience review
committee did not use a systematic process when making a safety significant
decision, to ensure safety is maintained and obtaining interdisciplinary inputs and
reviews on risk-significant decisions H.1(a) (Section 1R13).
Green. The inspectors identified two examples of a noncited violation of
10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawing,
for the failure of operations personnel to follow procedures to approve and
document operability determinations using adequate or technically correct
information. Specifically, on January 15, and January 22, 2010, operations
personnel failed to follow procedure SO123-XV-52, Functionality Assessments
and Operability Determinations, Revision 14, in that, the documented bases for
operability for degraded conditions did not adequately support the basis for an
operability position taken by the licensee. Following the inspectors identification
of the issues, operations personnel performed new operability determinations to
provide adequate bases for operability. This issue was entered into the
licensees corrective action program as Nuclear Notifications NNs 200765208
and 200753880.
The finding is greater than minor because, if left uncorrected, inadequate
operability determinations would have the potential to lead to a more significant
safety concern. Specifically, the failure to recognize that risk significant
equipment is in a potentially inoperable condition and as such, may not be able
to perform its specified safety function would not be recognized and accounted
for by operators. The finding is associated with the Mitigating Systems
Cornerstone. Using the Manual Chapter 0609, Significance Determination
Process, Phase 1 Worksheets, the finding is determined to have very low safety
significance because the finding: (1) is not a design or qualification issue
confirmed not to result in a loss of operability or functionality; (2) did not
represent an actual loss of safety function of the system or train; (3) did not result
in the loss of one or more trains of nontechnical specification equipment; and (4)
did not screen as potentially risk significant due to a seismic, flooding, or severe
weather initiating event. This finding has a crosscutting aspect in the area of
problem identification and resolution associated with the corrective action
program because the licensee failed to thoroughly evaluate problems such that
the resolutions addressed causes and extent of conditions as necessary P.1(c)
(Section 1R15).
Green. A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, was identified for failure of
maintenance planning personnel to develop and specify an adequate
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Enclosure
postmaintenance test in the work instructions used to perform maintenance on
the backup nitrogen regulator for the component cooling water surge tank.
Specifically, on October, 25, 2009, Maintenance Order MO 800335873 did not
specify postmaintenance testing instructions that would verify that nitrogen
supply valve PCV 5403 would perform satisfactorily in service, following
calibration, and properly control surge tank pressure during changes in surge
tank levels. This issue was entered into the licensees corrective action program
as Nuclear Notifications NNs 200766430 and 200887764.
The finding is greater than minor because it is associated with the procedure
quality attribute of the Mitigating Systems Cornerstone and affects the associated
cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Furthermore, the finding is similar to more than minor example 3.i in NRC
Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in
that, an extensive engineering evaluation was required to verify that the
component cooling water system remained capable of performing its safety
function during a design basis earthquake. Using the Manual Chapter 0609,
Appendix G, Shutdown Operations Significance Determination Process,
Phase 1 guidance, the finding is determined to have very low safety significance
because the finding did not result in an increase in the likelihood of a loss of
reactor coolant system inventory, degrade the ability to add reactor coolant
system inventory, or degrade the ability to recover decay heat removal. This
finding has a crosscutting aspect in the area of human performance associated
with work practices because maintenance planning personnel failed to follow
procedures to develop adequate work instructions to perform maintenance on
safety-related equipment H.4(b) (Section 1R19).
Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure
of licensee personnel to follow procedure SO123-XV-50.CAP-1, Writing Nuclear
Notifications for Problem Identification and Resolution, Revision 2, and enter
conditions adverse to quality into the corrective action program. Specifically,
between January 4 and March 14, 2010, the inspectors identified multiple
instances, including two programs, where licensee personnel were aware of the
existence of conditions adverse to quality, but failed to appropriately enter them
into the corrective action program without being prompted by the inspectors.
This issue was entered into the licensees corrective action program as Nuclear
Notifications NNs 200778816 and 200780926.
The finding is greater than minor because it was similar to more than minor
example 3.j in NRC Manual Chapter 0612, Appendix E, Examples of Minor
Issues, in that programmatic deficiencies were identified associated with this
issue that would have the potential to lead to more significant safety concerns if
left uncorrected. Specifically, contractor and licensee personnels failure to enter
conditions adverse to quality into the station corrective action program could
result in the licensees failure to recognize that risk significant equipment is in a
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Enclosure
degraded or nonconforming condition, and as such, may not be able to perform
its specified safety function. The finding is associated with the Mitigating
Systems Cornerstone. Using the Manual Chapter 0609, Significance
Determination Process, Phase 1 Worksheets, the finding is determined to have
very low safety significance because the finding: (1) is not a design or
qualification issue confirmed not to result in a loss of operability or functionality;
(2) did not represent an actual loss of safety function of the system or train; (3)
did not result in the loss of one or more trains of non-technical specification
equipment; and (4) did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. This finding has a crosscutting
aspect in the area of problem identification and resolution associated with the
corrective action program because the licensee failed to implement a corrective
action program with a low threshold for identifying issues. This also includes
identifying such issues completely, accurately, and in a timely manner
commensurate with their safety significance P.1(a) (Section 4OA2).
Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure
of maintenance personnel to follow Work Order 800195196 and provide
appropriate oversight to transmission and distribution personnel while performing
work in the electrical switchyard. Specifically, on February 26, 2010,
maintenance personnel failed to follow Work Order 800195196, and procedure
SO123-XV-15.3, Temporary System Alteration and Restoration, Revision 17, to
provide appropriate oversight of transmission and distribution personnel who
were performing work in the plant switchyard, which resulted in the over torquing
of nine bolts on the reserve auxiliary transformer circuit breakers. The licensee
corrected the over torqued bolt condition. This issue was entered into the
licensees corrective action program as Nuclear Notifications NNs 200803364
and 200811993.
The finding is greater than minor because circumventing procedural
requirements, if left uncorrected, would have the potential to lead to a more
significant safety concern, in that, more risk significant equipment could be
rendered inoperable without the knowledge and approval of appropriate
management or control room personnel. The finding is associated with the
Mitigating Systems Cornerstone. Using the Manual Chapter 0609, Significance
Determination Process, Phase 1 Worksheets, the finding is determined to have
a very low safety significance because the finding: (1) is not a design or
qualification issue confirmed not to result in a loss of operability or functionality;
(2) did not represent an actual loss of safety function of the system or train;
(3) did not result in the loss of one or more trains of nontechnical specification
equipment; and (4) did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. This finding has a crosscutting
aspect in the area of human performance associated with work practices
because maintenance personnel failed to ensure supervisory and management
oversight of work activities, including contractors, such that nuclear safety was
supported H.4(c) (Section 4OA2).
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Enclosure
Green. A self-revealing noncited violation of Technical Specification 5.5.1.1 was
identified for the failure of operations personnel to follow procedures for operating
the component cooling water system. Specifically, on January 27, 2010,
operations personnel failed to follow the requirements of procedure SO123-2-17,
Component Cooling Water System Operation, Revision 31, while performing a
planned drain down of the component cooling water surge tanks. Operations
personnel failed to maintain the surge tank pressure, in accordance with
procedure SO23-2-17, such that, component cooling water surge tank pressure
was permitted to go low out of the expected operating range. As a result of this
low surge tank pressure, operators declared the component cooling water and
shutdown cooling train A systems inoperable. This issue was entered into the
licensees corrective action program as Nuclear Notification NN 200771367.
The finding is greater than minor because the continued failure to follow
procedures when operating safety-related plant equipment, if left uncorrected,
would have the potential to lead to a more significant safety concern. The finding
is associated with the Mitigating Systems Cornerstone. Using the Manual
Chapter 0609, Appendix G, Shutdown Operations Significance Determination
Process, Phase 1 guidance, the finding is determined to have very low safety
significance because the finding did not result in an increase in the likelihood of a
loss of reactor coolant system inventory, degrade the ability to add reactor
coolant system inventory, or degrade the ability to recover decay heat removal.
This finding has a crosscutting aspect in the area of human performance
associated with work practices because operations personnel failed to use
proper human error prevention techniques and proceeded in the face of
unexpected circumstances when operating the component cooling water system
H.4(a) (Section 4OA3).
Cornerstone: Barrier Integrity
Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated
with the licensees failure to adequately implement procedures SO123-I-3.7,
Refueling Foreign Material Exclusion Control, Revision 6, and SO123-I-1.18,
Foreign Material Exclusion, Revision 14. Specifically, between January 12,
2010, and February 23, 2010, multiple occasions were identified during Refueling
Outage U2C16, where licensee personnel failed to implement appropriate foreign
material exclusion controls in areas designated as Zone 1 foreign material
exclusion areas. This issue was entered into the licensees corrective action
program as Nuclear Notifications NNs 200760484, 200742082, 200743834 and
200805961.
The finding is greater than minor because it is associated with the human
performance attribute of the Barrier Integrity Cornerstone and affects the
cornerstone objective of providing reasonable assurance that physical barriers
protect the public from radionuclide releases caused by accidents or events.
Furthermore, the programmatic deficiencies that were identified associated with
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Enclosure
this issue would have the potential to lead to a more significant safety concern, if
left uncorrected. Specifically, licensee personnels continued failure to implement
appropriate foreign material exclusion controls would result in degradation and
adverse impacts on materials and systems associated with the spent fuel pool or
the reactor cavity. Using the Manual Chapter 0609, Appendix G, Shutdown
Operations Significance Determination Process, Phase 1 guidance, the finding
is determined to have very low safety significance because the finding did not
result in an increase in the likelihood of a loss of reactor coolant system
inventory, degrade the ability to add reactor coolant system inventory, or degrade
the ability to recover decay heat removal. This finding had a crosscutting aspect
in the area of human performance associated with work practices because the
licensee failed to define and effectively communicate expectations regarding
procedural compliance which resulted in a failure to follow procedure by licensee
personnel H.4(b) (Section 1R20).
Cornerstone: Occupational Radiation Safety
Green. The inspectors identified a noncited violation of Technical Specification 5.8.3 for the failure of radiation protection personnel to appropriately barricade
and conspicuously post an area that was accessible to personnel that could have
resulted in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Specifically, from
February 2004 through March 17, 2010, the radiation personnel failed to
appropriately barricade and conspicuously post the access ladder to the upper
refueling cavity when it was being used as the means to control access to an
individual high radiation area in the lower cavity where the maximum measured
radiation dose rate was 2.8 rem per hour. The inspectors determined that the
ladder was not appropriately barricaded and conspicuously posted, and as such
the controls the licensee had in place were easily circumvented. On March 17,
2010, radiation protection personnel appropriately barricaded and conspicuously
posted the access ladder to the upper refueling cavity. This issue was entered
into the licensees corrective action program as Nuclear Notifications
NNs 200793188 and 200837345.
The finding is greater than minor because it is associated with the program and
process attribute of the Radiation Safety Cornerstone and directly affected the
associated cornerstone objective of ensuring the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation. Using Manual Chapter 0609,
Appendix C, Occupational Radiation Safety Significance Determination
Process, this finding is determined to have very low safety significance because
it did not involve: (1) an ALARA planning or work control issue, (2) an
overexposure, (3) a substantial potential for overexposure, or (4) an impaired
ability to assess dose. The inspectors determined that since the licensee had not
recently re-evaluated the locked high radiation area controls associated with this
ladder; this finding did not represent current plant performance, and therefore,
did not have a crosscutting aspect associated with it (Section 1R20).
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Enclosure
Other Findings
SL-IV. The inspectors identified a noncited violation of 10 CFR 50.72,
Immediate Notification Requirements for Operating Nuclear Power Reactors,
for the licensees failure to notify the NRC Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
following discovery of an event meeting the reportability criteria as specified.
Specifically, on December 23, 2009, the licensee failed to notify the NRC
Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the discovery of an event or condition that
resulted in a condition where the spent fuel pool cooling system was prevented
from fulfilling its safety function of residual heat removal with the complete core
off loaded. This issue was entered into the licensees corrective action program
as Nuclear Notification NN 200733257.
The finding is greater than minor because the NRC relies on licensees to identify
and report conditions or events meeting the criteria specified in regulations in
order to perform its regulatory function, and when this is not done the regulatory
function is impacted. The inspectors reviewed this issue in accordance with
Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through
this review, the inspectors determined that traditional enforcement was
applicable to this issue because the NRC's regulatory ability was affected. The
inspectors determined that this finding was not suitable for evaluation using the
significance determination process, and as such, was evaluated in accordance
with the NRC Enforcement Policy. The finding was reviewed by NRC
management and because the violation was determined to be of very low safety
significance, was not repetitive or willful, and was entered into the corrective
action program, this violation is being treated as a Severity Level IV noncited
violation consistent with the NRC Enforcement Policy. This finding has a
crosscutting aspect in the area of problem identification and resolution
associated with the corrective action program because the licensee failed to
thoroughly evaluate problems such that the resolutions addressed causes and
extent of conditions as necessary. This includes properly classifying, prioritizing,
and evaluating for operability and reportability conditions adverse to quality
P.1(c) (Section 4OA2).
SL-IV. The inspectors identified a noncited violation of 10 CFR 50.73, Licensee
Event Report System, associated with the failure of nuclear regulatory affairs
personnel to submit a licensee event report within 60 days following discovery of
an event meeting the reportability criteria as specified. Specifically, nuclear
regulatory affairs personnel failed to submit a licensee event report within 60
days following discovery of a complete loss of spent fuel pool cooling event that
occurred on February 13, 2007. This issue was entered into the licensees
corrective action program as Nuclear Notifications NNs 200740135 and
200733257.
The finding is greater than minor because the NRC relies on licensees to identify
and report conditions or events meeting the criteria specified in regulations in
order to perform its regulatory function, and when this is not done the regulatory
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Enclosure
function is impacted. The inspectors reviewed this issue in accordance with
Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through
this review, the inspectors determined that traditional enforcement was
applicable to this issue because the NRC's regulatory ability was affected. The
inspectors determined that this finding was not suitable for evaluation using the
significance determination process, and as such, was evaluated in accordance
with the NRC Enforcement Policy. The finding was reviewed by NRC
management and because the violation was determined to be of very low safety
significance, was not repetitive or willful, and was entered into the corrective
action program, this violation is being treated as a Severity Level IV noncited
violation consistent with the NRC Enforcement Policy. Since the inadequate
reportability determination had been made in 2007, and the licensees
reportability program has undergone significant revision since this time, the
inspectors determined that this was not reflective of current licensee performance
and therefore did not have a crosscutting aspect associated with it (Section
4OA2).
SL-IV. The inspectors identified a noncited violation of 10 CFR 50.59, Changes,
Test, and Experiments, for the failure of licensing personnel to obtain a technical
specification license amendment for a change made to the technical specification
bases concerning the emergency chilled water system. Specifically, in 1996,
licensing personnel implemented a technical specification bases change for
Limiting Condition for Operation 3.7.10, Emergency Chilled Water, which
changed the intent and application of the technical specification, and added
wording which allowed a period of time for required support systems to be
inoperable without declaring the emergency chillers inoperable. This issue was
entered into the licensees corrective action program as Nuclear Notifications
NNs 200747320 and 200758329.
The finding is greater than minor because the failure to follow the requirements of
10 CFR 50.59 and receive prior NRC approval for changes in licensed actions
impacted the NRCs regulatory ability. The inspectors reviewed this issue in
accordance with Inspection Manual Chapter 0612 and the NRC Enforcement
Manual. Through this review, the inspectors determined that traditional
enforcement was applicable to this issue because the NRC's regulatory ability
was affected. The inspectors determined that this finding was not suitable for
evaluation using the significance determination process, and as such, was
evaluated in accordance with the NRC Enforcement Policy. The finding was
reviewed by NRC management and because the violation was determined to be
of very low safety significance, was not repetitive or willful, and was entered into
the corrective action program, this violation is being treated as a Severity
Level IV noncited violation consistent with the NRC Enforcement Policy. Since
the bases change was made in 1996, the inspectors determined that this was not
reflective of current licensee performance and therefore did not have a
crosscutting aspect associated with it (Section 4OA2).
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Enclosure
B.
Licensee-Identified Violations
Violations of very low safety significance, which were identified by the licensee, have
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. These violations and
corrective action tracking numbers are listed in Section 4OA7.
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Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 2 began the inspection period shutdown for a scheduled refueling outage (U2C16) and
steam generator replacement, and remained there for the duration of the inspection period.
Unit 3 began the inspection period at full power. Between March 4 and March 10, 2010, the unit
reduced power to 50 percent for fuel conservation, and remained there for the duration of the
inspection period.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
Readiness for Impending Adverse Weather Conditions
a.
Inspection Scope
Since coastal flooding with potential tornados and high winds were forecast in the vicinity
of the facility for January 20 through January 22, 2010, the inspectors reviewed the
licensees overall preparations/protection for the expected weather conditions. On
January 20, 2010, the inspectors walked down the Unit 3 auxiliary feedwater structure
and the off site power distribution system because their safety-related functions could be
affected or required as a result of high winds or tornado-generated missiles or the loss of
offsite power. The inspectors evaluated the licensee staffs preparations against the
sites procedures and determined that the staffs actions were adequate. During the
inspection, the inspectors focused on plant-specific design features and the licensees
procedures used to respond to specified adverse weather conditions. The inspectors
also toured the plant grounds to look for any loose debris that could become missiles
during a tornado. The inspector's evaluated operator staffing and accessibility of
controls and indications for those systems required to control the plant. Additionally, the
inspectors reviewed the Updated Final Safety Analysis Report and performance
requirements for systems selected for inspection, and verified that operator actions were
appropriate as specified by plant-specific procedures. The inspectors also reviewed a
sample of corrective action program items to verify that the licensee identified adverse
weather issues at an appropriate threshold and dispositioned them through the
corrective action program in accordance with station corrective action procedures.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one readiness for impending adverse weather
condition sample as defined in IP 71111.01-05.
b.
Findings
No findings of significance were identified.
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Enclosure
1R04 Equipment Alignments (71111.04)
.1
Partial Walkdowns
a.
Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
February 23, 2009, Unit 2, containment alignment for integrated leakage rate test
March 10, 2010, Unit 3, auxiliary feedwater pump MP-141 alignment
March 11, 2010, Unit 2, emergency diesel generator train A
March 22, 2010, Unit 2, saltwater cooling train A
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could affect the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Final Safety Analysis Report, technical specification
requirements, administrative technical specifications, outstanding work orders, corrective
action documents, and the impact of ongoing work activities on redundant trains of
equipment in order to identify conditions that could have rendered the systems incapable
of performing their intended functions. The inspectors also walked down accessible
portions of the systems to verify system components and support equipment were
aligned correctly and operable. The inspectors examined the material condition of the
components and observed operating parameters of equipment to verify that there were
no obvious deficiencies. The inspectors also verified that the licensee had properly
identified and resolved equipment alignment problems that could cause initiating events
or impact the capability of mitigating systems or barriers and entered them into the
corrective action program with the appropriate significance characterization. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four partial system walkdown samples as
defined by IP 71111.04-05.
b.
Findings
No findings of significance were identified.
.2
Semi-Annual Complete Walkdown
a.
Inspection Scope
Between January 22, 2010, and March 24, 2010, the inspectors performed a complete
system alignment inspection of the Unit 2 safety injection system to verify the functional
capability of the system. The inspectors selected this system because it was considered
both safety-significant and risk-significant in the licensees probabilistic risk assessment.
The inspectors walked down the system to review mechanical and electrical equipment
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Enclosure
line ups, electrical power availability, system pressure and temperature indications, as
appropriate, component labeling, component lubrication, component and equipment
cooling, hangers and supports, operability of support systems, and to ensure that
ancillary equipment or debris did not interfere with equipment operation. The inspectors
reviewed a sample of past and outstanding work orders to determine whether any
deficiencies significantly affected the system function. In addition, the inspectors
reviewed the corrective action program database to ensure that system equipment-
alignment problems were being identified and appropriately resolved. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one complete system walkdown sample as
defined by IP 71111.04-05.
b.
Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
Quarterly Fire Inspection Tours
a.
Inspection Scope
The inspectors conducted fire protection walkdowns that were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
January 4, 2010, Units 2 and 3, hot work activities in the saltwater cooling pipe
tunnel
January 14, 2010, Unit 2, auxiliary feedwater pump tunnel
February 9, 2010, Unit 2, safety equipment building rooms 2 through 5 and 15
February 10, 2010, Unit 3, penetration building
The inspectors reviewed areas to assess if licensee personnel had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant; effectively maintained fire detection and suppression capability; maintained
passive fire protection features in good material condition; and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire
protection equipment, systems, or features, in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to affect equipment that could initiate or mitigate a plant
transient, or their impact on the plants ability to respond to a security event. Using the
documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
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Enclosure
fire detectors and sprinklers were unobstructed, that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four quarterly fire-protection inspection samples
as defined by IP 71111.05-05.
b.
Findings
Introduction. The inspectors identified three examples of a Green noncited violation of
Technical Specification 5.5.1.1.d, for the failure of contractor and station personnel to
properly implement the requirements of a station fire protection procedure for control of
hot work activities.
Description. On January 4, 2010, while performing a fire protection walk down of the
Unit 2 salt water cooling tunnel the inspectors noted contract personnel, who were being
supervised by station personnel, performing what appeared to be hot work activities on
the salt water cooling piping. The inspectors noted that the activities were producing
sparks and the sparks were coming in contact with unprotected combustible materials.
The inspectors inquired about this activity and were informed that a portion of the work,
grinding activities, had been classified as hot work and as such a flame permit was
associated with it and a fire watch was present. The inspectors reviewed the flame
permit and noted that it required all combustible material within 35 feet of the activity to
removed or covered. When the inspectors pointed this out to the fire watch they were
informed by the station personnel that were present, including supervisors, that the
evolution that was producing the sparks that were coming in contact with the
unprotected combustibles was flapper wheeling activities and was not subject to hot
work controls. The inspectors pointed out that the grinding was a hot work activity that
was in progress and required all materials to be removed or covered within 35 feet.
The inspectors questioned this response concerning the flapper wheel activities and
reviewed station procedure SO123-XV-1.41, Control of Ignition Sources, Revision 14,
to validate what they had been told. During this review the inspectors noted that
Section 6.2.1 stated, in part, For sanding and flapper wheel activities, all
flammable/combustible material shall be removed from within the area where the field of
sparks would be expected to spread from this activity, and if relocation is impractical
then shield all combustibles. As such, the inspectors determined that the procedure
had not been appropriately followed for either activity. Also, the personnel who were
performing the work, supervising the work, and performing fire watch duties were not
familiar with the procedural requirements for the activities being performed. Nuclear
Notification NN 200729747 was initiated to document the inspectors concerns.
On January 14, 2010, the inspectors observed work activities in the Unit 2 auxiliary
feedwater tunnel, and noted that welders were conducting hot work activities with
unprotected combustibles within 35 feet of the work area. The inspectors noted that the
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Enclosure
flame permit for the activity identified that all combustible material within 35 feet of the
activity either had to be removed or covered. When the inspectors pointed this out to the
fire watch and welders, the activity was stopped. The licensee initiated Nuclear
Notification NN 200746059 to capture this concern, and conducted a human
performance error review board. During this review, the licensee determined that the fire
watch and the welders had failed to follow the requirements of procedure
SO123-XV-1.41.
On March 16, 2010, the inspectors passed through the turbine building and noted sparks
coming from the overhead. Upon further investigation, the inspectors noted that the
sparks were coming from work activities occurring on the level above and the sparks
were coming in contact with unprotected combustible materials. The inspectors noted
that a fire watch was posted in the area and inquired of the adequacy of the work site.
The fire watches initial response was that this area was more than 35 feet away from
the work area therefore it was not an issue. The inspectors were not satisfied with this
response and requested that a supervisor come to the area. During discussions with the
supervisor, the inspectors learned that the activities that were occurring above were
flapper wheeling activities, and that the work area was supposed to be completely
enclosed. The inspectors also determined that the work group was not familiar with the
procedural requirements associated with flapper wheel activities. As such, the
inspectors determined that the licensee had failed to follow procedure SO123-XV-1.41
for flapper wheel activities and remove or cover all flammable/combustible material from
within the area where the field of sparks would be expected to spread. The licensee
initiated Nuclear Notification NN 200835830 to capture this concern.
Analysis. The failure to follow the requirements of a station fire protection procedure for
control of hot work activities was a performance deficiency. The finding is greater than
minor because if left uncorrected, the practice of conducting hot work in a manner that
allows uncontrolled combustibles to be within the procedurally specified exclusion area
would have the potential to lead to a more significant safety concern, in that, it could
result in a fire in or near risk important equipment. The finding is associated with the
Initiating Events Cornerstone. The inspectors determined that Manual Chapter 0609,
Appendix F, Fire Protection Significance Determination Process, does not address the
potential risk significance of shutdown fire protection findings, and Appendix G,
Shutdown Operations Significance Determination Process, does not address fire
protection findings, and therefore could not be applied to shutdown plant conditions.
Because of this, the inspectors used Manual Chapter 0609, Appendix M, Significance
Determination Process Using Qualitative Criteria. The NRC management review was
performed by using the Manual Chapter 0609, Appendix F, Phase 1 Worksheet, to
establish a bounding analysis. Using the bounding analysis, the finding is determined to
have very low safety significance because the finding represented a low degradation
rating, in that, it did not have any significant effect on the likelihood that a fire might
occur, or that a fire which does occur might not be promptly suppressed. This finding
had a crosscutting aspect in the area of human performance associated with work
practices, in that, the licensee failed to define and effectively communicate expectations
regarding procedural compliance and personnel follow procedures H.4(b).
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Enclosure
Enforcement. Technical Specification 5.5.1.1.d requires, in part, that written procedures
be established, implemented, and maintained covering Fire Protection Program
implementation. The Fire Protection Program was implemented, in part, by procedure
SO123-XV-1.41, Control of Ignition Sources, Revision 14. Procedure SO123-XV-1.41,
Steps 6.2.1 and 6.4.1.3, required that combustible materials be covered or removed
from the ignition sources. Contrary to the above, between January 4 and March 17,
2010, three examples were identified where contractor and station personnel failed to
properly implement the requirements of procedure SO123-XV-1.41, Steps 6.2.1 and
6.4.1.3. Specifically, contractor and station personnel failed to ensure that combustible
materials were covered or removed from the ignition source. Following the inspectors
identification of each example, the licensee immediately stopped the hot work activities
and restored compliance with the requirements of procedure SO123-XV-1.41. Because
this finding is of very low safety significance and has been entered into the licensees
corrective action program as Nuclear Notifications NNs 200729747, 200746059 and
200835830, this violation is being treated as a noncited violation, consistent with Section
VI.A of the NRC Enforcement Policy: NCV 05000361/2010002-01, Failure to Implement
Fire Protection Plan Requirements Related to Hot Work Activities.
1R06 Flood Protection Measures (71111.06)
a.
Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report, the flooding analysis,
and plant procedures to assess seasonal susceptibilities involving internal flooding;
reviewed the Updated Final Safety Analysis Report and corrective action program to
determine if licensee personnel identified and corrected flooding problems; inspected
underground bunkers/manholes to verify the adequacy of sump pumps, level alarm
circuits, cable splices subject to submergence, and drainage for bunkers/manholes;
verified that operator actions for coping with flooding can reasonably achieve the desired
outcomes; and walked down the one area listed below to verify the adequacy of
equipment seals located below the flood line, floor and wall penetration seals, watertight
door seals, common drain lines and sumps, sump pumps, level alarms, and control
circuits, and temporary or removable flood barriers. Specific documents reviewed during
this inspection are listed in the attachment.
March 15, 2010, Unit 3, auxiliary feedwater pump house
These activities constitute completion of one flood protection measures inspection
sample as defined by IP 71111.06-05.
b.
Findings
No findings of significance were identified.
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Enclosure
1R11 Licensed Operator Requalification Program (71111.11)
a.
Inspection Scope
On March 9, 2010, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification examinations to verify that operator
performance was adequate, evaluators were identifying and documenting crew
performance problems, and training was being conducted in accordance with licensee
procedures. The inspectors evaluated the following areas:
Licensed operator performance
Crews clarity and formality of communications
Crews ability to take timely actions in the conservative direction
Crews prioritization, interpretation, and verification of annunciator alarms
Crews correct use and implementation of abnormal and emergency procedures
Control board manipulations
Oversight and direction from supervisors
Crews ability to identify and implement appropriate technical specification
actions and emergency plan actions and notifications
The inspectors compared the crews performance in these areas to pre-established
operator action expectations and successful critical task completion requirements.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one quarterly licensed-operator requalification
program sample as defined in IP 71111.11.
b.
Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a.
Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems:
March 4, 2010, Units 2 and 3, instrument air system
March 24, 2010, Unit 3, auxiliary feedwater system
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Enclosure
The inspectors reviewed events caused by ineffective equipment maintenance that
resulted in valid or invalid automatic actuations of engineered safeguards systems and
independently verified the licensee's actions to address system performance or condition
problems in terms of the following:
Implementing appropriate work practices
Identifying and addressing common cause failures
Scoping of systems in accordance with 10 CFR 50.65(b)
Characterizing system reliability issues for performance
Charging unavailability for performance
Trending key parameters for condition monitoring
Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)
Verifying appropriate performance criteria for structures, systems, and
components classified as having an adequate demonstration of performance
through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as
requiring the establishment of appropriate and adequate goals and corrective
actions for systems classified as not having adequate performance, as described
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of two quarterly maintenance effectiveness
samples as defined in IP 71111.12-05.
b.
Findings
Introduction. The inspectors identified a Green noncited violation of 10 CFR
50.65(b)(2)(ii) for the licensees failure to appropriately scope the steam driven auxiliary
feedwater pumps trench eductor in the maintenance rule monitoring program.
Description. On January 21, 2010, operations personnel observed that water had come
in contact with the steam line mud leg in the Unit 3 steam driven auxiliary feedwater
pump steam supply trench during heavy rains. Operations personnel declared the
auxiliary feedwater pump inoperable in accordance with procedure SO23-2-4, Auxiliary
Feedwater System Operation, Revision 27, until the piping could be blown down and
the pump run for 30 minutes to verify that the piping was dried out. The licensee entered
this issue into their corrective action program as Nuclear Notification NN 200758566.
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Enclosure
The inspectors reviewed the maintenance rule functional failure evaluation associated
with Nuclear Notification NN 200758566. The inspectors noted that the licensee had
concluded that this event was not a functional failure of the eductor. The licensees
evaluation focused on the performance criteria of the auxiliary feedwater pump, and did
not appear to consider appropriate criteria for the trench eductor. The basis for the
conclusion was a calculation that had been performed to demonstrate that water in
contact with the steam line mud leg did not make the auxiliary feedwater pump
The eductors were installed in 1986 and were used to remove water from the steam
supply trench to prevent adverse affects on the auxiliary feedwater pump. Trench water
in contact with the pumps steam supply piping would cause condensation of the steam
in the pipe causing the potential for the turbine to over speed, which would render the
pump incapable of performing it specified safety function.
The inspectors observed that the trench eductor was not connected to the auxiliary
feedwater system, but that it was a support system installed to facilitate the auxiliary
feedwater pump being able to perform its specified safety function. The inspectors
questioned the adequacy of evaluating a failure of the eductor to perform its function,
preventing water from accumulating in the trench, against the performance criteria of the
auxiliary feedwater system, which was to provide a reliable source of feedwater to steam
generators during normal and emergency conditions. Through discussions with the
licensees maintenance rule coordinator, the inspectors determined that the eductors
were not scoped in the stations maintenance rule monitoring program. The
maintenance rule coordinator informed the inspectors that the eductors were not scoped
in the maintenance rule monitoring program because their failure could not directly
cause the failure of the auxiliary feedwater pump, and the station was not required to
consider hypothetical failures that resulted from system interdependencies that have not
been previously seen. The inspectors determined that the licensee had developed a
narrow interpretation of what directly meant based on a narrow interpretation of some
examples from NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants.
Through more reviews, the inspectors noted that the licensee determined that the
eductors had been installed to assist in removing any accumulated water in the trench,
to limit buildup of water to ensure that condensate does not accumulate in the steam
lines and cause an overspeed trip of the turbine. Furthermore, this had been done
based on past plant experience dealing with water causing condensation in the steam
piping. Therefore, the inspectors determined that the licensee had inappropriately
interpreted 10 CFR 50.65(b)(2)(ii), with regard to nonsafety-related structures, systems
and components whose failure could prevent safety-related structures, systems, and
components from fulfilling their safety-related function, and had failed to appropriately
scope the eductors for both Units 2 and 3 in the stations maintenance rule monitoring
program.
Analysis. The failure to properly scope the auxiliary feedwater trench eductors in the
maintenance rule monitoring program was a performance deficiency. The finding is
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Enclosure
greater than minor because it is associated with the equipment performance attribute of
the Mitigating Systems Cornerstone and directly affected the cornerstone objective of
ensuring the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. Using the Manual Chapter 0609,
Significance Determination Process, Phase 1 Worksheets, the finding is determined to
have very low safety significance because the finding: (1) is not a design or qualification
issue confirmed not to result in a loss of operability or functionality; (2) did not represent
an actual loss of safety function of the system or train; (3) did not result in the loss of one
or more trains of nontechnical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
The inspectors determined that since the scoping of the systems had occurred more
than 2 years in the past, and the opportunity to reevaluate system scoping had not
occurred recently, that the finding did not represent current plant performance and
therefore did not have a crosscutting aspect associated with it.
Enforcement. Title 10 CFR 50.65(b)(2)(ii) requires, in part, that the scope of the
monitoring program specified in paragraph (a)(1) of this section shall include nonsafety
related structures, systems and components whose failure could prevent safety-related
structures, systems, and components from fulfilling their safety-related function.
Contrary to the above, from the inception of the facilities monitoring program through
March 2010, the licensee failed to properly scope the steam drive auxiliary feedwater
pump trench eductor into the maintenance rule monitoring program. Because this
violation is of very low safety significance and has been entered into the licensees
corrective action program as Nuclear Notification NN 200765185, this violation is being
treated as a noncited violation consistent with Section VI.A of the NRC Enforcement
Policy: NCV 05000361;05000362/2010002-02, Failure to Appropriately Scope Auxiliary
Feedwater Pump Trench Eductors in the Maintenance Rule Monitoring Program.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a.
Inspection Scope
The inspectors reviewed licensee personnel's evaluation and management of plant risk
for the maintenance and emergent work activities affecting risk-significant and safety-
related equipment listed below to verify that the appropriate risk assessments were
performed prior to removing equipment for work:
January 13-14, 2010, Units 2 and 3, use of non-conservative technical
specifications for new fuel movement related to proposed change number
PCN 593
January 20, 2010, Unit 2, proposed cavity drain down activities during inclement
weather
February 3, 2010, Unit 2, diving operations in the intake area
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Enclosure
February 10-12, 2009, Units 2 and 3, safety monitor model change interim
measures to address uncertainty associated with manual operation of motor
operated valves
February 17, 2010, Units 2 and 3, mobile crane use in the electrical switchyard
The inspectors selected these activities based on potential risk significance relative to
the reactor safety cornerstones. As applicable for each activity, the inspectors verified
that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)
and that the assessments were accurate and complete. When licensee personnel
performed emergent work, the inspectors verified that the licensee personnel promptly
assessed and managed plant risk. The inspectors reviewed the scope of maintenance
work, discussed the results of the assessment with the licensee's probabilistic risk
analyst or shift technical advisor, and verified plant conditions were consistent with the
risk assessment. The inspectors also reviewed the technical specification requirements
and inspected portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of six maintenance risk assessments and
emergent work control inspection samples as defined by IP 71111.13-05.
b.
Findings
1. Operating Experience Review
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees
failure to properly implement procedure requirements to ensure that applicable risk
significant operating experience was entered into the corrective action program for timely
evaluation.
Description. On December 17, 2009, an industry operating experience report was
reviewed by the operating experience review committee regarding lessons learned from
the industry related to the expected differential pressure across locally operated valves,
which must be considered when evaluating the ability of operators to change valve
position in accident conditions. The review determined the operating experience was
not applicable and was distributed as information only; not requiring any action. On
February 10, 2010, the probabilistic risk assessment group initiated interim
compensatory measures for the safety monitor model used to assess the risk associated
with on-line work activities. The interim actions were taken following the probabilistic risk
assessment groups recognition that the industry operating experience report had a
potential impact and were conservatively used to address uncertainty associated with
the manual operation of auxiliary feedwater motor operated valves under the differential
pressures expected during accident conditions.
On February 11, 2010, the inspectors questioned the timeliness of the risk significant
operating experience report evaluation that took several months to be properly assessed
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Enclosure
by the probabilistic risk assessment group. On February 23, 2010, based on prompting
by the inspectors, the licensee initiated Nuclear Notification NN 200805879 to
investigate the timeliness of their operating experience review of the event involving the
expected differential pressure across locally operated valves which could impact risk
significant components. The evaluation identified the initial industry operating
experience review failed to recognize the applicability of the operating experience or the
potential risk significant impact that needed further analysis. As such, this information
was not entered into the corrective action program, and therefore, not directed to
appropriate subject matter experts or communicated to the affected station groups in a
timely manner as required by procedure SO23-XV-40, Sharing Industry Information,
Revision 1. The evaluation also concluded the operating experience review committee
lacked a knowledge basis to recognize the potential implications, and instead of using a
systematic approach, depended upon distribution to other departments and personnel to
assess the need for entry into the corrective action program for evaluation of the impact
to risk-significant and safety-significant activities.
Analysis. The failure to properly implement procedure requirements to ensure adequate
review of applicable industry operating experience was a performance deficiency. The
finding is greater than minor because it is associated with the procedure quality attribute
of the Mitigating Systems Cornerstone and affects the associated cornerstone objective
to ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. Using the Manual Chapter 0609,
Significance Determination Process, Phase 1 Worksheets, the finding is determined to
have very low safety significance because the finding: (1) is not a design or qualification
issue confirmed not to result in a loss of operability or functionality; (2) did not represent
an actual loss of safety function of the system or train; (3) did not result in the loss of one
or more trains of nontechnical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
This finding has a crosscutting aspect in the area of human performance associated with
decision-making because the operating experience review committee did not use a
systematic process when making a safety significant decision, to ensure safety is
maintained and obtaining interdisciplinary inputs and reviews on risk-significant
decisions H.1(a).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires that activities affecting quality shall be prescribed by
documented instructions, procedures or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with those instructions,
procedures, and drawings. Procedure SO23-XV-40, Sharing Industry Information,
Revision 1, required actions to ensure a review of industry operating experience for
applicability and the need for timely evaluation in the corrective action program.
Contrary to the above, on December 17, 2009, the operating experience review
committee failed to properly implement the requirements of procedure SO23-XV-40.
Specifically, an industry operating experience report review determined the operating
experience was not applicable and was distributed as information only; not requiring any
action. The same industry operating experience was later determined to be applicable
by the probabilistic risk assessment group, and interim compensatory measures were
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Enclosure
initiated on February 10, 2010, to address the issues. Because this finding is of very low
safety significance and has been entered into the licensees corrective action program
as Nuclear Notifications NN 200805879, this violation is being treated as a noncited
violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000361;05000362/2010002-03, Failure to Enter Operating Experience into Corrective Action
Program for Timely Evaluation.
2. Risk Assessment for Switchyard Activities
Introduction. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4),
Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants, involving multiple instances where operations and work control personnel failed
to adequately assess and implement appropriate risk management activities for work in
the stations electrical switchyard.
Description. On February 17, 2010, the licensee determined that the station had failed
to perform an adequate risk assessment for proposed crane activities in the switchyard
with regard to Unit 3, which was operating at full power. Before allowing the activities to
commence the licensee performed the required risk assessment, and classified the work
as a high risk activity in the switchyard for Unit 3, and commenced the crane activity.
The inspectors subsequently reviewed the risk assessment on February 18, 2010.
During their review, the inspectors determined that this assessment had been performed
only for Unit 3, as identified under the additional requirements section, which stated;
maintain requirements per procedure SO23-5-1.8.1, Shutdown Nuclear Safety,
Revision 23, on Unit 2. Based on this, the inspectors questioned how the activities being
performed in the switchyard had been assessed with regard to Unit 2, which was
shutdown in Mode 5 at the time.
The inspectors reviewed procedure SO23-5-1.8.1, and noted that the following:
The stated objective of the procedure was to provide guidelines for controlling
evolutions and activities while in Mode 5 and 6 to ensure that Shutdown Safety
Functions are maintained Operable, Functional, or Available as required to
support the station philosophy of Defense in Depth
Section 6.1.1 defined electrical power availability as a Shutdown Safety Function
Attachment 1, Definitions, Section 1.8 defined a high risk evolution as; Outage
activities, plant configurations, or conditions during shutdown where the plant is
more susceptible to an event causing the loss of a shutdown safety function.
Section 6.11, Control of High Risk Evolutions, provided specific guidance on
evaluating these evolutions and establishing required risk management actions
As a result, the inspectors determined that; an adequate risk assessment had not been
performed for Unit 2, and the requirements of Section 6.11 of procedure SO23-5-1.8.1
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Enclosure
had not been implemented with respect to implementing required risk management
actions for the on-going crane activities in the switchyard.
The inspectors presented this information indicating a failure to adequately assess risk
associated with the crane activities and implement appropriate risk management actions,
relative to Unit 2 to the licensee. During discussions with station personnel, the
inspectors were informed that the station believed that the Defense in Depth planning
sheets were the stations risk assessment for Unit 2, and since they had not removed any
of the identified systems from service they were within their analysis. The inspectors
pointed out that procedure SO23-5-1.8.1, Section 6.1.1.3 stated, in part:
The selected safety function fulfillment plans are recorded in the Defense in
Depth planning sheets. These are tables which document the pre-planned safety
function fulfillment plan methods, safety function protection plan, or other
contingency plans for each safety function.
Accordingly, the inspectors identified that the crane activities had not been assessed
and incorporated into the stations defense in depth strategy, and as such, the Defense in
Depth planning sheets were not an appropriate risk assessment for this activity.
The licensee determined that an appropriate risk assessment had not been performed,
and when one was performed, risk management actions were identified as required by
procedure SO23-5-1.8.1. On February 19, 2010, the licensee initiated Nuclear
Notification NN 200801929 to document the issue and implement corrective actions.
Subsequently, on February 23, 2010, the inspectors questioned why operations
personnel were allowing work on a support system for a Unit 2 emergency diesel
generator while switchyard work was still in progress. While investigating this concern,
the licensee determined that the crane had been removed from the switchyard on
February 19, 2010. This resulted in the risk management actions for the Unit 2
emergency diesel generators being discontinued. However, there was a failure to
recognize and properly assess a man-lift that was staged for use in the switchyard. Use
of the man-lift would also require risk management actions for the Unit 2 emergency
diesel generators. Subsequently, the licensee was able to determine that the man-lift
had not been used from February 19 through 23, 2010. The licensee initiated Nuclear
Notification NN 200805635 to document this issue.
Analysis. The failure to perform an adequate risk assessment and implement
appropriate risk management actions was a performance deficiency. The finding is
greater than minor since it was similar to both more than minor examples 7.e and 7.f in
NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues,
because when the activities were correctly assessed plant procedures required risk
management actions to be taken. The finding is associated with the Initiating Events
Cornerstone. The inspectors determined that the licensee does not maintain a
shutdown probabilistic risk analysis model, and as such, an incremental core damage
probability cannot be estimated for the plant conditions that existed at the time of the
performance deficiency. For this reason, the inspectors determined that Manual
Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management
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Enclosure
Significance Determination Process, Flowchart 2, could not be used to determine the
risk significance the finding. Using the qualitative review process of Manual
Chapter 0609, Appendix M, Significance Determination Process Using Qualitative
Criteria, the finding is determined to have very low safety significance because the
finding did not result in any additional loss of defense in depth systems. This finding has
a crosscutting aspect in the area of human performance associated with the work
practices because the licensee failed to define and effectively communicate expectations
regarding procedural compliance and that personnel follow procedures H.4(b).
Enforcement. Title 10 CFR 50.65(a)(4), states in part, that before performing
maintenance activities (including but not limited to surveillance, postmaintenance testing,
and corrective and preventive maintenance), the licensee shall assess and manage the
increase in risk that may result from the proposed maintenance activities. Contrary to
the above, between February 18, and February 23, 2010, operations and work control
personnel failed to adequately assess and manage the increase in risk associated with
maintenance activities in the electrical switchyard. Following the inspectors
identification of the findings, the licensee adequately assessed and managed the
increase in risk for the maintenance activities. Because this finding is of very low safety
significance and has been entered into the licensees corrective action program as
Nuclear Notifications NNs 200801929 and 200805635, this violation is being treated as a
noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000361/2010002-04, Failure to Assess and Manage Risk for Electrical Switchyard
Impacting Maintenance.
1R15 Operability Evaluations (71111.15)
a.
Inspection Scope
The inspectors reviewed the following issues:
January 3-5, 2010, Unit 2, inspectors identified various seismic issues associated
with the gap required between containment interior and exterior structures
requiring various evaluations and Unit 3 at power entry
January 13, 2010, Unit 2, operability impact of through wall piping flaws found on
emergency core cooling system Train A piping
January 19, 2010, Unit 3, operability impact of a through wall piping flaw on the
common emergency core cooling system mini-flow line
January 22, 2010, Unit 2, operability impact due to suspected growth of through
wall piping flaws previously identified on emergency core cooling system Train A
piping
February 2, 2010, Unit 3, intake structure integrity
February 4, 2010, Unit 3, through wall flaw indication on emergency core cooling
system piping
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Enclosure
February 9-10, 2009, Units 2 and 3, seat leak requirements for component
cooling water pump discharge valves
February 12-14, 2010, Unit 2, safety related battery 2B007 surveillance results
indicate battery at 85 percent of service life
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that technical specification operability was
properly justified and the subject component or system remained available such that no
unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the technical specifications and Updated
Final Safety Analysis Report to the licensees evaluations, to determine whether the
components or systems were operable. Where compensatory measures were required
to maintain operability, the inspectors determined whether the measures in place would
function as intended and were properly controlled. The inspectors determined, where
appropriate, compliance with bounding limitations associated with the evaluations.
Additionally, the inspectors also reviewed a sampling of corrective action documents to
verify that the licensee was identifying and correcting any deficiencies associated with
operability evaluations. Specific documents reviewed during this inspection are listed in
the attachment.
These activities constitute completion of eight operability evaluations inspection samples
as defined in IP 71111.15-05.
b.
Findings
Introduction. The inspectors identified two examples of a Green noncited violation of
10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawing, for the
failure of operations personnel to follow procedures to approve and document operability
determinations using adequate or technically correct information.
Description. The inspectors reviewed the operability determinations documented in
Nuclear Notifications NNs 200745284 and 200760570, to verify the evaluation adequacy
and compliance with procedure SO123-XV-52, Functionality Assessments and
Operability Determinations, Revision 14. Nuclear Notification NN 200745284 was
written on January 14, 2010, to document a through wall pipe leak on the Unit 3
emergency core cooling system miniflow common discharge line. During their review,
the inspectors noted that the licensee had classified the flaw as a pinhole leak, based on
the visible appearance of the flaw at the time of discovery, and had developed an
immediate operability determination based on this characterization. However, at 12
midnight on January 15, 2010, as part of their prompt operability determination data
gathering, the licensee had performed nondestructive examination testing and
discovered that the flaw was actually a 0.5 inch linear flaw, and this was reported to
operations personnel at 00:45 a.m. Operations personnel believed that this new
classification was bounded by the original immediate operability determination.
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Enclosure
However, the inspectors noted that NRC Inspection Manual Part 9900 guidance,
Operability Determinations, Paragraph 4.6, Timing of Operability Determinations,
states, in part, If, at any time, information is developed that negates a previous
determination that there is a reasonable expectation that the structures, systems and
components is operable, the licensee should declare the structures, systems and
components inoperable. As such the inspectors determined that this new information,
the characterization of the flaw as a linear indication versus a pinhole, should have
resulted in a new immediate operability determination being performed. The inspectors
communicated their concerns to operations personnel. The licensee performed a new
immediate operability determination, and initiated Nuclear Notification NN 200753880 to
capture this issue in their corrective action program.
Nuclear Notification NN 200760570 was initiated to document an increase in flaw size
for previously identified flaws on the Unit 3 train A emergency core cooling system
suction header, identified during augmented inspections on January 22, 2010. As a
result of this new condition being identified, the licensee performed an immediate
operability determination using; the calculated growth rates, the calculated maximum
allowed flaw size, and the systems mission time of 120 days.
The inspectors determined that the licensees operability determination was inadequate.
Specifically, their use of a 120 day mission time did not adequately address the flaw
growth rate in relation to the calculated maximum allowed flaw size. Specifically, the
calculated flaw growth rate would exceed the maximum allowed flaw size before the
systems 120 day mission time would be completed. The inspectors informed the
licensee of their concerns. The licensee performed a new operability determination to
provide adequate bases for operability, and initiated Nuclear Notification NN 200765208
to capture this issue in their corrective action program.
Analysis. The failure to follow procedures to approve an adequate basis for operability
was a performance deficiency. The finding is greater than minor because, if left
uncorrected, inadequate operability determinations would have the potential to lead to a
more significant safety concern. Specifically, the failure to recognize that risk significant
equipment is in a potentially inoperable condition and as such, may not be able to
perform its specified safety function would not be recognized and accounted for by
operators. The finding is associated with the Mitigating Systems Cornerstone. Using
the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets,
the finding is determined to have very low safety significance because the finding: (1) is
not a design or qualification issue confirmed not to result in a loss of operability or
functionality; (2) did not represent an actual loss of safety function of the system or train;
(3) did not result in the loss of one or more trains of nontechnical specification
equipment; and (4) did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. This finding has a crosscutting aspect in the
area of problem identification and resolution associated with the corrective action
program because the licensee failed to thoroughly evaluate problems such that the
resolutions addressed causes and extent of conditions as necessary P.1(c).
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Enclosure
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions or drawings of a type appropriate to the circumstances and
shall be accomplished in accordance with these instructions and drawings.
Procedure SO123-XV-52, Functionality Assessments and Operability Determinations,
Revision 14, required that operations personnel make a definitive statement of
operability and the basis for the statement. Contrary to the above, on January 15, and
January 22, 2010, operations personnel failed to follow procedure SO123-XV-52, in that,
the documented bases for operability for degraded conditions did not adequately support
the basis for an operability position taken by the licensee. Because this finding is of very
low safety significance and has been entered into the licensees corrective action
program as Nuclear Notifications NNs 200765208 and 200753880, this violation is being
treated as a noncited violation consistent with Section VI.A of the NRC Enforcement
Policy: NCV 05000362/2010002-05, Failure to Follow Procedure Results in an
Inadequate Operability Determination.
1R19 Postmaintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed the following postmaintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
January 29, 2010, Unit 2, retest of 2PCV-5403, nitrogen pressure control valve
train A for component cooling water surge tank
February 5, 2010, Unit 2, functional testing of spliced resistance temperature
detectors to reactor coolant system loop 2 hot leg channel B narrow range
February 5, 2010, Unit 2, boration dilution controls system preoperational testing
March 3, 2010, Unit 2, containment integrated leak rate test
The inspectors selected these activities based upon the structure, system, or
component's ability to affect risk. The inspectors evaluated these activities for the
following (as applicable):
The effect of testing on the plant had been adequately addressed; testing was
adequate for the maintenance performed
Acceptance criteria were clear and demonstrated operational readiness; test
instrumentation was appropriate
The inspectors evaluated the activities against the technical specifications, the Updated
Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and
various NRC generic communications to ensure that the test results adequately ensured
that the equipment met the licensing basis and design requirements. In addition, the
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Enclosure
inspectors reviewed corrective action documents associated with postmaintenance tests
to determine whether the licensee was identifying problems and entering them in the
corrective action program and that the problems were being corrected commensurate
with their importance to safety. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of four postmaintenance testing inspection
samples as defined in IP 71111.19-05.
b.
Findings
Introduction. A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, was identified for failure of
maintenance planning personnel to develop and specify an adequate postmaintenance
test in the work instructions used to perform maintenance on the backup nitrogen
regulator for the component cooling water surge tank.
Description. On January 27, 2010, both component cooling water surge tank levels
were lowered, using procedure SO23-2-17, Component Cooling Water System
Operation, Revision 31. The component cooling water surge tanks were required to
have a nitrogen pressure between 33-40 psig to remain operable. Pressure in
component cooling water surge tanks trains A and B were maintained with nitrogen
supply valves PCV 5403 and PCV 5404, respectively. The valves were designed to
regulate pressure at 38 +/-1 psig when properly calibrated. During the evolution,
operations personnel failed to follow procedure SO23-2-17 to monitor nitrogen pressure
such that it could be maintained while lowering level, since they incorrectly assumed the
nitrogen supply valves were properly calibrated and would automatically maintain surge
tank nitrogen pressure in the required range. However, nitrogen supply valve PCV 5403
did not function as expected and failed to maintain nitrogen surge tank pressure in the
acceptable range for operability. The performance deficiencies associated with this
event are documented as NCV 05000361/2010002-14 of this report.
Nuclear Notification NN 200771367 was initiated to evaluate the event. The evaluation
determined that nitrogen supply valve PCV 5403 did not have the correct setpoints and
was improperly calibrated. Instrument and control maintenance technicians last
completed a maintenance calibration on the valve on October 25, 2009, using procedure
SO123-II-9.176, Pressure Reducing Regulators - Calibration, Revision 2. During this
maintenance, technicians failed to follow the requirements of procedure SO123-II-9.176
to properly calibrate the pressure control valve which resulted in the pressure control
valve not properly maintaining nitrogen pressure in the surge tank as the volume in the
surge tank was lowered on January 27, 2010.
The inspectors reviewed the maintenance history for nitrogen supply valve PCV 5403,
including Maintenance Order MO 800335873, which implemented maintenance
procedure SO123-II-9.176 to perform the calibration. The maintenance procedure
contained a section for restoration and return to service following the calibration. The
inspectors observed that the maintenance procedure SO123-II-9.176, Section 6.4,
Restoration and Return to Service, did not require any postmaintenance or functional
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Enclosure
test to ensure the nitrogen supply valve would properly maintain pressure following the
calibration when returned to service. The inspectors also observed that Maintenance
Order MO 800335873 did not specify any other test or verification that would ensure that
nitrogen supply valve PCV 5403 was capable of performing its design function following
the maintenance activity.
Procedure SO123-I-1.7, Work Order Preparation and Processing, Revision 30,
Attachment 5, Step 1.1, contained instructions for the determination of adequate
postmaintenance test requirements for maintenance activities. Procedure SO123-I-1.7,
Step 1.1.1 stated, in part, that if the maintenance procedure did not list any test
requirements, then refer to procedure SO23-I-1.25, Post Maintenance Testing,
Revision 0, for guidelines in determining adequate testing requirements. Procedure
SO23-I-1.25, Attachment 4, described a functional test as a test or verification to ensure
that the component, equipment, or subsystem that was affected by the maintenance
activity was completely capable of performing its design function. Further, it stated that
functional tests or checks, such as verification that calibrations have been satisfactorily
completed, should be considered where specific test guides have not been provided.
Following this review, the inspectors concluded that Maintenance Order MO 800335873
did not specify adequate postmaintenance testing as required by procedures
SO123-I-1.7 and SO23-I-1.25.
The inspectors communicated their observations to licensee personnel, and verified that
their concerns were captured in Nuclear Notifications NNs 200766430 and 200887764.
An engineering analysis was required to demonstrate that the component cooling water
system train A remained operable during the period from October 25, 2009, to
January 27, 2010. The engineering evaluation determined that the system would have
been able to fulfill all its intended safety functions as defined in the Updated Final Safety
Analysis Report, Section 9.2.2.2. Following the improper calibration determination, on
January 28, 2010, nitrogen supply valve PCV 5403 was re-calibrated and an adequate
postmaintenance test was performed.
Analysis. The failure to establish work instructions to include adequate postmaintenance
test requirements to verify equipment operability following maintenance was a
performance deficiency. The finding is greater than minor because it is associated with
the procedure quality attribute of the Mitigating Systems Cornerstone and affects the
associated cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Furthermore, the finding is similar to more than minor example 3.i in NRC Inspection
Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that, an extensive
engineering evaluation was required to verify that the component cooling water system
remained capable of performing its safety function during a design basis earthquake.
Using the Manual Chapter 0609, Appendix G, Shutdown Operations Significance
Determination Process, Phase 1 guidance, the finding is determined to have very low
safety significance because the finding did not result in an increase in the likelihood of a
loss of reactor coolant system inventory, degrade the ability to add reactor coolant
system inventory, or degrade the ability to recover decay heat removal. This finding has
a crosscutting aspect in the area of human performance associated with work practices
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Enclosure
because maintenance planning personnel failed to follow procedures to develop
adequate work instructions to perform maintenance on safety-related equipment
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
procedures, or drawings. Maintenance Order MO 800335873 established the
instructions to perform a calibration for a safety-related pressure reducing regulator.
Contrary to the above, on October, 25, 2009, Maintenance Order MO 800335873 did not
include adequate testing required to demonstrate that the component cooling water
system remained operable following maintenance. Specifically, Maintenance Order
MO 800335873 did not specify postmaintenance testing instructions that would verify
that nitrogen supply valve PCV 5403 would perform satisfactorily in service, following
calibration, and properly control surge tank pressure during changes in surge tank
levels. Because this finding is of very low safety significance and has been entered into
the licensees corrective action program as Nuclear Notifications NNs 200766430 and
200887764, this violation is being treated as a noncited violation consistent with
Section VI.A of the NRC Enforcement Policy: NCV 05000361/2010002-06, Failure to
Perform an Adequate Postmaintenance Test.
1R20 Refueling and Other Outage Activities (71111.20)
a.
Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the Unit 2
refueling outage (U2C16) and steam generator replacement, including activities
associated with a stuck reactor vessel head alignment pin, conducted January 26-28,
2010, to confirm that licensee personnel had appropriately considered risk, industry
experience, and previous site-specific problems in developing and implementing a plan
that assured maintenance of defense-in-depth. During the refueling outage, the
inspectors observed portions of the shutdown and cooldown processes and monitored
licensee controls over the outage activities listed below.
Configuration management, including maintenance of defense-in-depth, is
commensurate with the outage safety plan for key safety functions and
compliance with the applicable technical specifications when taking equipment
out of service
Clearance activities, including confirmation that tags were properly hung and
equipment appropriately configured to safely support the work or testing
Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error
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Enclosure
Status and configuration of electrical systems to ensure that technical
specifications and outage safety-plan requirements were met, and controls over
switchyard activities
Monitoring of decay heat removal processes, systems, and components
Verification that outage work was not impacting the ability of the operators to
operate the spent fuel pool cooling system
Reactor water inventory controls, including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss
Controls over activities that could affect reactivity
Maintenance of secondary containment as required by the technical
specifications
Refueling activities, including fuel handling and sipping to detect fuel assembly
leakage
Licensee identification and resolution of problems related to refueling outage
activities
Specific documents reviewed during this inspection are listed in the attachment.
Refueling Outage U2C16 was still in progress at the end of this inspection period.
Consequently, these activities constitute only a partial completion of one refueling outage
and other outage inspection sample as defined in IP 71111.20-05.
b.
Findings
1. Foreign Material Exclusion Area Controls
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of
licensee personnel to follow procedures associated with foreign material exclusion
controls in areas designated as Zone 1 foreign material exclusion areas, on multiple
occasions, during Refueling Outage U2C16.
Description. On January 13, 2010, while performing core reload operations, station
personnel identified foreign material in the bottom of the reactor cavity. Refueling
personnel decided that since this material was not in the way of the current assemblies
being loaded that the reload could continue and the material recovered at a more
convenient time in the future. Refueling personnel generated Nuclear Notification
NN 200743228 to capture this issue in the corrective action program.
The inspectors reviewed this nuclear notification as well as procedure SO123-I-1.18,
Foreign Material Exclusion Control, Revision 14. During this review the inspectors
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Enclosure
noted that Attachment 5, Foreign Material Exclusion Controls, Section 13, Recovery
from Loss of FME Control, required, in part, to promptly stop all work in the immediate
area, not take any action that could cause further migration of the foreign material,
recover the foreign material if it can be easily retrieved, or generate a Notification which
should evaluate whether the associated work can resume before recovering the foreign
material. The inspectors determined that the actions of refueling personnel following the
identification of foreign material in the reactor cavity were contrary to the requirements of
procedure SO123-I-1.18. The inspectors informed the licensee of their observations,
and the licensee entered this issue into their corrective action program as Nuclear
Notification NN 200743834. Subsequently, the licensee determined that refueling
personnel had failed to reference procedure SO123-I-1.18 when foreign material had
been discovered on January 13, 2010.
During subsequent observations of the licensees activities in and around other Zone 1
foreign material exclusion areas (areas which required the highest level of foreign
material exclusion controls) the inspectors identified four additional instances where
licensee personnel failed to appropriately implement procedural requirements associated
with Zone 1 foreign material exclusion controls. Specifically:
January 12, 2010, station personnel were instructed to enter the Zone 1 foreign
material exclusion area around the spent fuel pool wearing anti-contamination
clothing, booties and gloves, and then remove the clothing and place it in the
trash bag in the area without entering it in the foreign material exclusion log so
that it could be tracked
January 22, 2010, the inspectors identified an instance where the foreign
material exclusion area watch logged material being brought out of the Zone 1
foreign material exclusion area around the reactor refueling cavity that had not
been logged into the area, which represented a loss of foreign material exclusion
controls
January 22, 2010, the inspectors identified a nylon rope in the Zone 1 foreign
material exclusion area around the reactor refueling cavity being used to restrain
material that had frayed ends that were not adequately covered
February 23, 2010, the inspectors identified that the Zone 1 foreign material
exclusion area around the Unit 3 spent fuel pool had material in it that was not
being tracked and controlled as required
The inspectors concluded that not all of these examples of the licensees failure to follow
procedure SO123-I-3.7, Refueling Foreign Material Exclusion Control, directly resulted
in the introduction of foreign material into a critical system. They were, however,
indicative of a programmatic issue associated with the licensees proper implementation
of the foreign material exclusion control program. The inspectors informed the licensee
of their observations, and the licensee entered this issue into their corrective action
program as Nuclear Notifications NNs 200742082, 200760484, and 200805961.
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Analysis. The failure of licensee personnel to follow procedures for the control of foreign
material was a performance deficiency. The finding is greater than minor because it is
associated with the human performance attribute of the Barrier Integrity Cornerstone and
affects the cornerstone objective of providing reasonable assurance that physical
barriers protect the public from radionuclide releases caused by accidents or events.
Furthermore, the programmatic deficiencies that were identified associated with this
issue would have the potential to lead to a more significant safety concern, if left
uncorrected. Specifically, licensee personnels continued failure to implement
appropriate foreign material exclusion controls would result in degradation and adverse
impacts on materials and systems associated with the spent fuel pool or the reactor
cavity. Using the Manual Chapter 0609, Appendix G, Shutdown Operations
Significance Determination Process, Phase 1 guidance, the finding is determined to
have very low safety significance because the finding did not result in an increase in the
likelihood of a loss of reactor coolant system inventory, degrade the ability to add reactor
coolant system inventory, or degrade the ability to recover decay heat removal. This
finding had a crosscutting aspect in the area of human performance associated with
work practices because the licensee failed to define and effectively communicate
expectations regarding procedural compliance which resulted in a failure to follow
procedure by licensee personnel H.4(b).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
procedures, or drawings. Contrary to the above, between January 12, 2010, and
February 23, 2010, the inspectors identified several examples where the licensee failed
to adequately implement foreign material exclusion controls as required by procedure
SO123-I-1.18. Because this finding is of very low safety significance and has been
entered into the licensees corrective action program as Nuclear Notifications
NNs 200760484, 200742082, 200743834 and 200805961, this violation is being treated
as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000361/2010002-07, Failure to Adequately Implement Foreign Material
Exclusion Controls.
2. Controls for Locked High Radiation Area
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.8.3 for the failure of radiation protection personnel to appropriately
barricade and conspicuously post an area that was accessible to personnel that could
have resulted in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Description. On February 12, 2010, while touring the Unit 2 containment building, the
inspectors noted that the ladder that provided access to the upper refueling cavity was
being used to control access to a locked high radiation area in the lower refueling cavity.
The inspectors noted that the ladder had a safety cage around it, a swing door to restrict
access inside of the safety cage, and the locked high radiation sign was attached to the
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swing door of the safety cage. However, there was nothing on the back side of the
ladder to either restrict access or denote it as a locked high radiation area.
The inspectors questioned the adequacy of the posting and access control method being
used by the licensee. Specifically, the placement of the sign on the swing door was such
that it was not clearly visible if the ladder was approached from the back side, and the
inspectors concluded that its placement was confusing as to where the locked high
radiation area actually was. The inspectors also questioned whether the back side of
the ladder was appropriately barricaded and conspicuously posted in a way to prevent
access. The inspectors informed the licensee of their concerns. The licensee initiated
Nuclear Notification NN 200793188 to capture this concern in the corrective action
program.
The licensees initial determination was that the posting was adequate and the back side
of the ladder was sufficiently controlled. The inspectors questioned this determination
and initiated discussions with the NRC Office of Nuclear Reactor Regulation.
The inspectors determined that the posting and method of barricading the ladder was
inadequate. Specifically, the controls the licensee had in place were easily
circumvented, and as such, the inspectors determined that the licensee had failed to
appropriately control access to the lower refueling cavity where there was an area where
the maximum measured radiation dose rate was 2.8 rem per hour. On March 17, 2010,
radiation protection personnel appropriately barricaded and conspicuously posted the
access ladder to the upper refueling cavity.
Analysis. The failure to appropriately barricade and conspicuously post areas that are
accessible to personnel that could result in radiation doses greater than 1.0 rem in 1
hour was a performance deficiency. The finding is greater than minor because it is
associated with the program and process attribute of the Radiation Safety Cornerstone
and directly affected the associated cornerstone objective of ensuring the adequate
protection of the worker health and safety from exposure to radiation from radioactive
material during routine civilian nuclear reactor operation. Using Manual Chapter 0609,
Appendix C, Occupational Radiation Safety Significance Determination Process, this
finding is determined to have very low safety significance because it did not involve:
(1) an ALARA planning or work control issue, (2) an overexposure, (3) a substantial
potential for overexposure, or (4) an impaired ability to assess dose. The inspectors
determined that since the licensee had not recently re-evaluated the locked high
radiation area controls associated with this ladder; this finding did not represent current
plant performance, and therefore, did not have a crosscutting aspect associated with it.
Enforcement. Technical Specifications 5.8.3 states, in part, that individual high radiation
areas that are accessible to personnel that could result in radiation doses greater than
1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas where no enclosure exists to enable
locking and where no enclosure can be reasonably constructed, the individual area shall
be barricaded and conspicuously posted. Contrary to the above, from February 2004
through March 17, 2010, the radiation personnel failed to appropriately barricade and
conspicuously post the access ladder to the upper refueling cavity when it was being
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Enclosure
used as the means to control access to an individual high radiation area in the lower
cavity where the maximum measured radiation dose rate was 2.8 rem per hour.
Because this violation is of very low safety significance and it was entered into the
licensees corrective action program as Nuclear Notifications NNs 200793188 and
200837345, this violation is being treated as a noncited violation consistent with
Section VI.A of the NRC Enforcement Policy: NCV 05000361/2010002-08, Failure to
Appropriately Control Access to a Locked High Radiation Area.
1R22 Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report, procedure
requirements, and technical specifications to ensure that the seven surveillance activities
listed below demonstrated that the systems, structures, and/or components tested were
capable of performing their intended safety functions. The inspectors also verified that
licensee personnel identified and implemented any needed corrective actions associated
with the surveillance testing.
January 22, 2010, Unit 2, high pressure and low pressure safety injection open
check valve inservice test results review
February 11, 2010, Unit 3, salt water cooling pump P113 comprehensive full flow
test
March 3, 2010, Unit 2, local leak rate test penetration 19
March 9, 2010, Unit 2, inservice valve test of pressurizer spray valve MU976
March 10, 2010, Unit 3, reactor power calibration surveillance
March 16, 2010, Unit 3, containment spray pump in-service and valve test
March 22, 2010, Unit 2, low pressure safety injection pump MP016
The inspectors witnessed test performance and/or reviewed test performance
documentation to verify that the significant surveillance test attributes were adequate to
address the following:
Prevention of preconditioning
Evaluation of testing impact on the plant
Clear acceptance criteria and procedure guidance
Adequacy of test equipment
Adequacy of documentation of test results and data
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Adequacy of jumper/lifted lead controls
Testing frequency and method demonstrated technical specification operability
Test equipment removal
Restoration of plant systems
Fulfillment of ASME Code requirements
Updating of performance indicator data
Engineering evaluations, root causes, and bases for returning tested systems,
structures, and components not meeting the test acceptance criteria were correct
Reference setting data
Annunciators and alarms setpoints.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of seven surveillance testing inspection samples
as defined in IP 71111.22-05.
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a.
Inspection Scope
The inspectors performed an in-office review of the San Onofre Nuclear Generating
Station Emergency Plan, Revision 28, submitted by the licensee December 17, 2009.
This revision updated letters of agreement with offsite authorities, updated the letter of
agreement with the Institute of Nuclear Power Operations, and updated the site policy
regarding the responsibilities of the shift manager.
This revision was compared to its previous revision, to the criteria of NUREG-0654,
Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1, and to the standards in
10 CFR 50.47(b) to determine if the revision adequately implemented the requirements
of 10 CFR 50.54(q). This review was not documented in a safety evaluation report and
did not constitute approval of licensee-generated changes; therefore, this revision is
subject to future inspection.
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Enclosure
These activities constitute completion of one sample as defined in Inspection Procedure
71114.04-05.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1
Data Submission Issue
a.
Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the Fourth
Quarter 2009 performance indicators for any obvious inconsistencies prior to its public
release in accordance with Inspection Manual Chapter 0608, Performance Indicator
Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b.
Findings
No findings of significance were identified.
.2
Unplanned Scrams per 7000 Critical Hours (IE01)
a.
Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams per 7000 Critical
Hours performance indicator for Units 2 and 3 for the period from the first quarter 2009
through the fourth quarter 2009. To determine the accuracy of the performance indicator
data reported during those periods, performance indicator definitions and guidance
contained in NEI Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 6, was used. The inspectors reviewed the licensees operator
narrative logs, issue reports, event reports and NRC Inspection reports for the period of
January 1, 2009, through December 31, 2009, to validate the accuracy of the submittals.
The inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified. Specific documents reviewed are
described in the attachment to this report.
These activities constitute completion of two unplanned scrams per 7000 critical hours
samples as defined in Inspection Procedure 71151-05.
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Enclosure
b.
Findings
No findings of significance were identified.
.3
Unplanned Scrams with Complications (IE02)
a.
Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams with
Complications performance indicator for Units 2 and 3 for the period from the first
quarter 2009 through the fourth quarter 2009. To determine the accuracy of the
performance indicator data reported during those periods, performance indicator
definitions and guidance contained in NEI Document 99-02, Regulatory Assessment
Performance Indicator Guideline, Revision 6, was used. The inspectors reviewed the
licensees operator narrative logs, issue reports, event reports and NRC integrated
inspection reports for the period of January 1, 2009, through December 31, 2009, to
validate the accuracy of the submittals. The inspectors also reviewed the licensees
issue report database to determine if any problems had been identified with the
performance indicator data collected or transmitted for this indicator and none were
identified. Specific documents reviewed are described in the attachment to this report.
These activities constitute completion of two unplanned scrams with complications
samples as defined in Inspection Procedure 71151-05.
b.
Findings
No findings of significance were identified.
.4
Unplanned Power Changes per 7000 Critical Hours (IE03)
a.
Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Transients per 7000
Critical Hours performance indicator Units 2 and 3 for the period from the first quarter
2009 through the fourth quarter 2009. To determine the accuracy of the performance
indicator data reported during those periods, performance indicator definitions and
guidance contained in NEI Document 99-02, Regulatory Assessment Performance
Indicator Guideline, Revision 6, was used. The inspectors reviewed the licensees
operator narrative logs, issue reports, event reports and NRC integrated inspection
reports for the period of January 1, 2009, through December 31, 2009, to validate the
accuracy of the submittals. The inspectors also reviewed the licensees issue report
database to determine if any problems had been identified with the performance
indicator data collected or transmitted for this indicator and none were identified.
Specific documents reviewed are described in the attachment to this report.
These activities constitute completion of two unplanned transients per 7000 critical hours
samples as defined in Inspection Procedure 71151-05.
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Enclosure
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical
Protection
.1
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. The inspectors reviewed attributes that included: the complete and
accurate identification of the problem; the timely correction, commensurate with the
safety significance; the evaluation and disposition of performance issues, generic
implications, common causes, contributing factors, root causes, extent of condition
reviews, and previous occurrences reviews; and the classification, prioritization, focus,
and timeliness of corrective. Minor issues entered into the licensees corrective action
program because of the inspectors observations are included in the attached list of
documents reviewed.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure, they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b.
Findings
No findings of significance were identified.
.2
Daily Corrective Action Program Reviews
a.
Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees corrective action program. The inspectors
accomplished this through review of the stations daily corrective action documents.
The inspectors performed these daily reviews as part of their daily plant status
monitoring activities and, as such, did not constitute any separate inspection samples.
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b.
Findings
No findings of significance were identified.
.3
Selected Issue Follow-up Inspection
a.
Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors recognized a corrective action item documenting the issues listed below. The
inspectors considered the following during the review of the licensees actions: (1)
complete and accurate identification of the problem in a timely manner; (2) evaluation
and disposition of operability/reportability issues; (3) consideration of extent of condition,
generic implications, common cause, and previous occurrences; (4) classification and
prioritization of the resolution of the problem; (5) identification of root and contributing
causes of the problem; (6) identification of corrective actions; and (7) completion of
corrective actions in a timely manner.
January 5, 2010, Unit 2, reportability review associated with the loss of spent fuel
pool cooling event that occurred on December 23, 2009
February 14, 2010, Unit 2, main transformer and unit auxiliary transformer
breaker trips following attempted start of reactor coolant pump motor M004 as
documented in Nuclear Notification NN 200794912
February 26, 2010, Unit 2, inadequate oversight of transmission and distribution
personnel who were performing work in the plant switchyard per Work Order 800195196
These activities constitute completion of three in-depth problem identification and
resolution samples as defined in IP 71152-05.
b.
Findings
1. Missed Eight Hour Report
Introduction. The inspectors identified a Severity Level IV noncited violation of
10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power
Reactors, for the licensees failure to notify the NRC Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
following discovery of an event meeting the reportability criteria as specified.
Description. On December 23, 2009, Unit 2 was in refueling outage U2C16 with; all fuel
off-loaded to the spent fuel pool, train A of saltwater cooling in service, train B was out of
service and drained for maintenance, spent fuel pool cooling was in service and
providing residual heat removal, and component cooling water was in service providing
cooling to spent fuel pool cooling. At approximately 10:00 a.m., operations personnel
received the saltwater cooling train A low flow and component cooling water heat
exchanger differential pressure high alarms. They noted flow rapidly lowering and heat
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Enclosure
exchanger differential pressure rising. Based on the observed plant conditions,
operations personnel entered abnormal operating instruction SO23-13-7, Loss of
Component Cooling Water/Saltwater Cooling, Revision 14. This procedure directed
operations personnel to secure both the saltwater cooling and the component cooling
water pumps, and line up for reverse flow of the saltwater cooling heat exchanger, based
on the observed indications. Due to this action, operations personnel entered Licensee
Controlled Specification 3.7.106, Spent Fuel Pool Operation, Condition B, and initiated
procedure SO23-3-2.11, Spent Fuel Pool Operations, Revision 26, Attachment 17, to
monitor spent fuel pool temperature due to the loss of spent fuel pool cooling.
Approximately one and one half hours later, reverse flow of the heat exchanger was
initiated and verified to be satisfactory and the abnormal operating instruction was
exited.
On January 5, 2010, the resident inspectors reviewed the licensees followup of this
event. During their review, the inspectors noted that the licensee had concluded the
event was caused by debris entering the system through a failed pump suction screen.
The licensee had also concluded that this event was not reportable to the NRC. This
decision had been made based on the licensees determination that the Technical
Specifications for component cooling water, 3.7.7, and salt water cooling, 3.7.8, were
only applicable in Modes 1-4, and when in Modes 5 and 6, the operability requirements
are determined by the systems they support, and Unit 2 was defueled and, therefore,
outside of all defined Modes. Therefore component cooling water and salt water
cooling were not required to be OPERABLE by any Technical Specification, and as such
not reportable.
The inspectors questioned the licensees reportability conclusion. The inspectors noted
that the applicability of Licensee Controlled Specification 3.7.106 was At all times with
irradiated fuel in the spent fuel pool, and as such, this specification was not mode
dependant. The inspectors also determined that this required the component cooling
water and salt water cooling systems be in operation as support systems for the spent
fuel pool cooling system to be operable. Furthermore, the inspectors noted that
procedure S023-5-1.8.1, Shutdown Nuclear Safety, Revision 23, classified the spent
fuel pool cooling system as providing the safety function fulfillment plan by providing
residual heat removal with the core off loaded to the spent fuel pool. As such, this event
prevented the fulfillment of the safety function of structures or systems that are needed
to remove residual heat when the salt water cooling and component cooling water
pumps were secured, and should have been reported to the NRC as such.
The inspectors informed the licensee of their concerns. The licensee initiated Nuclear
Notification NN 200733257 to address this concern. Subsequently, the licensee
determined that this event did represent an event that prevented the fulfillment of the
safety function of structures or systems that are needed to remove residual heat, and
submitted a late 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report and Licensee Event Report 05000361/2009-004-00, Both
Trains of Spent Fuel Pool Cooling Inoperable Results in a Loss of Safety Function.
Analysis. The failure to make an applicable non-emergency 8-hour event notification
report within the required time frame was a performance deficiency. The finding is
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Enclosure
greater than minor because the NRC relies on licensees to identify and report conditions
or events meeting the criteria specified in regulations in order to perform its regulatory
function, and when this is not done the regulatory function is impacted. The inspectors
reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC
Enforcement Manual. Through this review, the inspectors determined that traditional
enforcement was applicable to this issue because the NRC's regulatory ability was
affected. The inspectors determined that this finding was not suitable for evaluation
using the significance determination process, and as such, was evaluated in accordance
with the NRC Enforcement Policy. The finding was reviewed by NRC management and
because the violation was determined to be of very low safety significance, was not
repetitive or willful, and was entered into the corrective action program, this violation is
being treated as a Severity Level IV noncited violation consistent with the NRC
Enforcement Policy. This finding has a crosscutting aspect in the area of problem
identification and resolution associated with the corrective action program because the
licensee failed to thoroughly evaluate problems such that the resolutions addressed
causes and extent of conditions as necessary. This includes properly classifying,
prioritizing, and evaluating for operability and reportability conditions adverse to quality
Enforcement. Title 10 CFR 50.72, Immediate Notification Requirements for Operating
Nuclear Power Reactors, requires, in part, that the licensee shall notify the NRC
Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after discovery of a nonemergency event described in
paragraph (b)(3)(v). Title 10 CFR 50.72(b)(3)(v)(B) requires, in part, any event or
condition that at the time of discovery could have prevented the fulfillment of the safety
function of structures or systems that are needed to remove residual heat shall be
reported within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of discovery. Contrary to the above, on December 23, 2009, the
licensee failed to notify the NRC Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the discovery of
an event or condition that resulted in a condition where the spent fuel pool cooling
system was prevented from fulfilling its safety function of residual heat removal with the
complete core off loaded. This finding was determined to be applicable to traditional
enforcement because the failure to report conditions or events meeting the criteria
specified in regulations affects the NRCs regulatory ability. The finding was evaluated in
accordance with the NRC's Enforcement Policy. The finding was reviewed by NRC
management and because the violation was of very low safety significance, was not
repetitive or willful, and was entered into the corrective action program as Nuclear
Notification NN 200733257, this violation is being treated as a Severity Level IV noncited
violation, consistent with the NRC Enforcement Policy: NCV 05000361/2010002-09,
Failure to Notify the NRC Within Eight Hours of a Non-Emergency Event.
2. Missed Licensee Event Report
Introduction. The inspectors identified a Severity Level IV noncited violation of
10 CFR 50.73, Licensee Event Report System, associated with the failure of nuclear
regulatory affairs personnel to submit a licensee event report within 60 days following
discovery of an event meeting the reportability criteria as specified.
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Enclosure
Description. During their review of a recent issue involving the loss of spent fuel pool
cooling, documented as NCV 05000361/2010002-09 in this report, the inspectors
became aware of another instance where spent fuel pool cooling had been lost.
Specifically, on February 13, 2007, Unit 2 was operating at 100 percent, with train A
spent fuel pool cooling pump 2P009 out of service for maintenance, and train B pump
2P010 in service providing cooling. At approximately 12:49 p.m., pump 2P010 tripped
on over current, which resulted in a complete loss of spent fuel pool cooling. Based on
this plant condition, operations personnel entered abnormal operating instruction SO23-
13-23, Loss of Spent Fuel Pool Cooling, Revision 10, and entered Licensee Controlled
Specification 3.7.106, Spent Fuel Pool Operation. Approximately 78 minutes later
operators restored pump 2P010 to service, which restored spent fuel pool cooling.
The licensee entered this issue into their corrective action program as Action Request
AR 070200583, and performed a reportability evaluation. Through this evaluation,
regulatory affairs personnel concluded this event was not reportable because the
conditions of Licensee Controlled Specification 3.7.106 were satisfied. Specifically,
spent fuel pool cooling had been lost for 78 minutes and specification 3.7.106 had a 6
hour action statement.
The inspectors questioned the licensees reportability conclusion. Specifically, the
inspectors noted that the Updated Final Safety Analysis Report, Section 3.1.6.2,
Criterion 61 - Fuel Storage and Handling and Radioactivity Control, identified that the
spent fuel pool cooling system provides cooling to remove residual heat from the spent
fuel pool, and Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System, stated that
the system was designed to provide continuous cooling for the spent fuel pool. As such,
the inspectors determined that this event represented a condition that alone prevented
the fulfillment of the safety function of the spent fuel pool cooling system that was
needed to remove residual heat.
The inspectors informed the licensee of their concerns. The licensee initiated Nuclear
Notification NN 200740135 to address this concern. Subsequently, the licensee
determined that this event did represent a condition that alone prevented the fulfillment
of the safety function of the spent fuel pool cooling system that was needed to remove
residual heat, and submitted a Licensee Event Report 05000361/2007-007-00,
Inoperable SFP Cooling Pumps Results in Loss of Safety Function.
Analysis. The failure to submit a required licensee event report within 60 days following
an event requiring a report to the NRC was a performance deficiency. The finding is
greater than minor because the NRC relies on licensees to identify and report conditions
or events meeting the criteria specified in regulations in order to perform its regulatory
function, and when this is not done the regulatory function is impacted. The inspectors
reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC
Enforcement Manual. Through this review, the inspectors determined that traditional
enforcement was applicable to this issue because the NRC's regulatory ability was
affected. The inspectors determined that this finding was not suitable for evaluation
using the significance determination process, and as such, was evaluated in accordance
with the NRC Enforcement Policy. The finding was reviewed by NRC management and
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Enclosure
because the violation was determined to be of very low safety significance, was not
repetitive or willful, and was entered into the corrective action program, this violation is
being treated as a Severity Level IV noncited violation consistent with the NRC
Enforcement Policy. Since the inadequate reportability determination had been made in
2007, and the licensees reportability program has undergone significant revision since
this time, the inspectors determined that this was not reflective of current licensee
performance and therefore did not have a crosscutting aspect associated with it.
Enforcement. Title 10 CFR 50.73, Licensee Event Report System, requires, in part,
that a licensee shall submit a licensee event report for any event of the type described in
paragraph (a)(1) within 60 days after the discovery of the event.
Title 10 CFR 50.73(a)(2)(v)(B) requires, in part, that licensees report any event or
condition that alone could have prevented the fulfillment of the safety function of
structures or systems that are needed to remove residual heat. Contrary to the above,
nuclear regulatory affairs personnel failed to submit a licensee event report within 60
days following discovery of a complete loss of spent fuel pool cooling event that
occurred on February 13, 2007. This finding was determined to be applicable to
traditional enforcement because the failure to report conditions or events meeting the
criteria specified in regulations affects the NRCs regulatory ability. The finding was
evaluated in accordance with the NRC's Enforcement Policy. The finding was reviewed
by NRC management and because the violation was of very low safety significance, was
not repetitive or willful, and was entered into the corrective action program as Nuclear
Notification NN 200740135, this violation is being treated as a Severity Level IV noncited
violation, consistent with the NRC Enforcement Policy: NCV 05000361/2010006-10,
Failure to Report a Safety System Functional Failure.
3. Technical Specification Bases Change
Introduction. The inspectors identified a Severity Level IV noncited violation of
10 CFR 50.59, Changes, Test, and Experiments, for the failure of licensing personnel
to obtain a technical specification license amendment for a change made to the technical
specification bases concerning the emergency chilled water system.
Description. While performing a review of an event on Unit 2 involving the loss of spent
fuel pool cooling, documented as NCV 05000361/2010002-09 in this report, the
inspectors noted a concern associated with Units 2 and 3 emergency chillers. The
inspectors noted that Units 2 and 3 share two emergency chillers, ME-335 and ME-336,
between the two units, and one chiller would normally be lined up to be operated from
the Unit 2 component cooling water system and one chiller would be lined up to be
operated from the Unit 3 component cooling water system. On December 23, 2009,
emergency chiller ME-336 was lined up to Unit 2 and emergency chiller ME-335 was
lined up to Unit 3, when Unit 2 experienced a clogging event of the only operable train of
salt water cooling system which resulted in a loss of component cooling water. The
inspectors questioned why operations personnel for Unit 3 failed to enter Technical
Specification 3.7.10, Emergency Chilled Water, in response to this event. Specifically,
the inspectors noted that the units technical specifications defined operability as:
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Enclosure
A system, subsystem, train, component or device shall be OPERABLE or have
OPERABILITY when it is capable of performing its specified function(s). Implicit
in this definition shall be the assumption that all necessary attendant
instrumentation, controls, normal and emergency electrical power sources,
cooling or seal water, lubrication or other auxiliary equipment that are required for
the system, subsystem, train, component or device to perform its function(s) are
also capable of performing their related support function(s).
As such, the inspectors determined that the loss of the only operable train of salt water
cooling which resulted in the loss of component cooling water represented the loss of
required support systems for the emergency chiller, which were required for the chiller to
be considered operable.
The inspectors informed operations personnel of their concern. Operations personnel
subsequently informed the inspectors that they had 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to transfer the emergency
chiller before it had to be considered inoperable, and referred the inspectors to the
bases of Technical Specification 3.7.10, which stated, in part:
An emergency chiller is considered OPERABLE when it is or can be aligned to
either Unit's operating or standby OPERABLE Component Cooling Water (CCW)
critical loop, provided that the OPERABLE CCW critical loop can be placed in
operation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a design basis event is detected in the Control
Room. Thus, an emergency chiller, under normal circumstances, remains
OPERABLE during a transfer operation between OPERABLE CCW critical loops
completed in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The inspectors questioned whether this language constituted a change to the intent of
the technical specification. The licensee initiated Nuclear Notification NN 200747320 to
evaluate the inspectors concern.
The inspectors determined that the licensee had changed the bases for Technical
Specification 3.7.10 to add the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowance in 1997 under bases change B96-001.
The inspectors reviewed this bases change package and determined that the
10 CFR 50.59 review that licensing personnel performed had not appropriately
evaluated this allowance. Furthermore, the inspectors determined that the only
documentation the licensee had to support the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowance was a memorandum
from Engineering to Operations, V. Barone to T. Vogt, dated December 22, 1994,
Component Cooling Water System/Emergency Chilled Water System Interaction,
SONGS, Units 2 and 3, which the inspectors determined was not adequate to support
the bases change.
Following consultation with the NRC Technical Specification Branch regarding the intent
of Technical Specification 3.7.10, the inspectors determined that the intent of the
specification was that the emergency chiller could not be considered operable if a
required support system was inoperable. Consequently, the inspectors determined that
the licensees bases change had, in effect, changed the intent of Technical
Specification 3.7.10, and this had been done without a license amendment. As such,
the inspectors determined that on December 23, 2009, operations personnel failed to
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enter Limiting Condition of Operation 3.7.10 when a required support system for the
emergency chillers was inoperable, which rendered emergency chiller ME-336
The inspectors informed the licensee of their determination. The licensee initiated
Nuclear Notification NN 200758329 to address this issue. Subsequently, the licensee
determined that the bases change did constitute a change to the technical specifications.
Analysis. The failure to adequately implement the requirements of 10 CFR 50.59 for a
change made to the bases of Technical Specification 3.7.10, which changed the intent of
the specification, was a performance deficiency. The finding is greater than minor
because the failure to follow the requirements of 10 CFR 50.59 and receive prior NRC
approval for changes in licensed actions impacted the NRCs regulatory ability. The
inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and
the NRC Enforcement Manual. Through this review, the inspectors determined that
traditional enforcement was applicable to this issue because the NRC's regulatory ability
was affected. The inspectors determined that this finding was not suitable for evaluation
using the significance determination process, and as such, was evaluated in accordance
with the NRC Enforcement Policy. The finding was reviewed by NRC management and
because the violation was determined to be of very low safety significance, was not
repetitive or willful, and was entered into the corrective action program, this violation is
being treated as a Severity Level IV noncited violation consistent with the NRC
Enforcement Policy. Since the bases change was made in 1996, the inspectors
determined that this was not reflective of current licensee performance and therefore did
not have a crosscutting aspect associated with it.
Enforcement. Title 10 CFR 50.59 (c)(1)(i) states, in part, that a licensee may make
changes in the facility as described in the final safety analysis report (as updated)
without obtaining a license amendment pursuant to 10 CFR 50.90 only if a change to the
technical specifications incorporated in the license is not required. Contrary to the
above, in 1997, licensing personnel implemented a technical specification bases change
for Limiting Condition for Operation 3.7.10, Emergency Chilled Water, which changed
the intent and application of the technical specification. Specifically, licensing personnel
added wording which allowed a period of time for required support systems to be
inoperable without declaring the emergency chillers inoperable. This finding was
determined to be applicable to traditional enforcement because the failure to follow the
requirements of 10 CFR 50.59 and receive prior NRC approval for changes in licensed
actions impacted the NRCs regulatory ability. The finding was evaluated in accordance
with the NRC's Enforcement Policy. The finding was reviewed by NRC management
and because the violation was of very low safety significance, was not repetitive or
willful, and was entered into the corrective action program as Nuclear Notifications NNs
200747320 and 200758329, this violation is being treated as a Severity Level IV
noncited violation, consistent with the NRC Enforcement Policy: NCV 05000361;05000362/2010002-11, Failure to Obtain a License Amendment for a Technical
Specification Bases Change.
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Enclosure
4. Threshold for Problem Identification
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of
licensee personnel to follow procedures to enter conditions adverse to quality into the
corrective action program.
Description. The inspectors reviewed Nuclear Notification NN 200794912 which had
been initiated following operations personnel attempted start of reactor coolant pump
motor M004, following work being performed on its control panel under engineering
change package 800074306. The attempted start resulted in the main transformer
breakers and the unit auxiliary transformer breakers tripping. The inspectors noted that
the licensee had determined that maintenance personnel had encountered an issue with
the installation of new components causing interference with existing terminal boards in
the panels. This resulted in the maintenance personnel deviating from the approved
engineering change package 800074306 and relocating a terminal block within the
panel. The inspectors determined that this deviation was inappropriate because it
resulted in a change in the scope of the work, and as such, should have required a
revision to the engineering change package.
Subsequently, the inspectors attended the human performance error review board which
reviewed the sequence of events and relevant facts associated with this issue. During
this review, licensee personnel confirmed that maintenance personnel had deviated from
the engineering change package when relocating the terminal blocks. They also pointed
out that this had been done under verbal approval from station engineering in response
to Nuclear Notification NN 200247324, Task 31.
At the completion of the review board, the inspectors expressed concerns to the licensee
about how this work had been accomplished and the fact that a nuclear notification had
not been written to capture this issue in the corrective action program. The licensee
informed the inspectors that this work had been done using the modification problem
reporting process detailed in procedure SO123-XXIX-2.16, Modification Problem
Reports, Revision 7, and that another nuclear notification was not necessary since their
process had been followed.
The inspectors reviewed the modification problem reporting process and noted that for
systems that were out of service with modifications being performed, maintenance
personnel were directed to generate a principle notification, and then add tasks to this
nuclear notification as issues were encountered. The inspectors questioned this process
since it appeared to conflict with corrective action program procedure
SO123-XV-50.CAP-1, Writing Nuclear Notifications for Problem Identification and
Resolution, Revision 2. Specifically, Section 6.1.3 required that, All SONGS
employees and supplemental personnel are responsible for promptly identifying,
reporting and documenting problems by writing a nuclear notification.
During subsequent review, the inspectors determined that the modification problem
reporting process was being used for modification activities on safety-related equipment
as well. Specifically, Nuclear Notifications NN 200457233 and 200718733 had been
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Enclosure
initiated as principle notifications for issues discovered while performing modifications to
the turbine of the steam driven auxiliary feedwater pump and the train B emergency
diesel generator. As such, the inspectors determined that this represented a program
operating outside of the corrective action program. The licensee initiated Nuclear
Notification NN 200770377 to capture the inspectors concern. Subsequently, the
licensee determined that this program was being implemented in a manner inconsistent
with the corrective action program.
As the inspectors continued to monitor the licensees activities during the refueling
outage they became aware that contractor personnel were being allowed to implement
their own problem identification process, field change requests, instead of entering all
conditions adverse to quality into the licensees corrective action program as required.
The inspectors determined that this contractor process was being used for issues that
were identified with safety-related and non-safety-related plant equipment. The
inspectors questioned this program because it appeared to be another example of a
program operating outside of the corrective action program.
The inspectors informed the licensee of their concern. The licensee informed the
inspectors that they had opted to allow the contractor to use their process during the
refueling outage, and that licensee staff was reviewing all field change requests to
determine if they warranted generation of a nuclear notification. The licensee informed
the inspectors that this contractor process was being implemented in accordance with
procedure 25221-000-GPP-GCP-00018, Field Change Request/Notices, Revision 0.
When the inspectors asked about the procedure controlling the licensees staff reviews
of the field change requests they were informed that there was none.
The inspector reviewed procedure GPP-GCP-00018 and noted that its purpose was for
systems that were out of service with modifications being performed under engineering
change packages. It directed contractor personnel to initiate a field change request
when issues were identified, which would be reviewed by contractor personnel for
disposition using contractor procedures. The inspectors concluded that this was an
additional process that did not meet the requirements of procedure
SO123-XV-50.CAP-1, Section 6.1.3. The licensee initiated Nuclear Notification
NN 200827841 to document the inspectors concern. Subsequently, the licensee
determined that this program was being implemented in a manner inconsistent with the
corrective action program.
The inspectors concluded that these examples of licensee personnels failure to enter
conditions adverse to quality into the licensees corrective action program, individually
and collectively, did not impact the licensees overall ability to monitor the condition of
station equipment. However, multiple departments, which included supervisors, were
responsible for not entering conditions adverse to quality into the corrective action
program even when these issues clearly resulted in degraded and nonconforming
conditions. Therefore, these instances were indicative of a systemic programmatic issue
with proper implementation of the corrective action program, with respect to
communicating and reinforcing the requirements for nuclear notification initiation.
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Enclosure
Analysis. The failure to follow procedures for entering conditions adverse to quality into
the corrective action program was a performance deficiency. The finding is greater than
minor because it was similar to more than minor example 3.j in NRC Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that programmatic deficiencies were
identified associated with this issue that would have the potential to lead to more
significant safety concerns if left uncorrected. Specifically, contractor and licensee
personnels failure to enter conditions adverse to quality into the station corrective action
program could result in the licensees failure to recognize that risk significant equipment
is in a degraded or nonconforming condition, and as such, may not be able to perform its
specified safety function. This finding is associated with the Mitigating Systems
Cornerstone. Using the Manual Chapter 0609, Significance Determination Process,
Phase 1 Worksheets, the finding is determined to have very low safety significance
because the finding: (1) is not a design or qualification issue confirmed not to result in a
loss of operability or functionality; (2) did not represent an actual loss of safety function
of the system or train; (3) did not result in the loss of one or more trains of nontechnical
specification equipment; and (4) did not screen as potentially risk significant due to a
seismic, flooding, or severe weather initiating event. This finding has a crosscutting
aspect in the area of problem identification and resolution associated with the corrective
action program because the licensee failed to implement a corrective action program
with a low threshold for identifying issues. This also includes identifying such issues
completely, accurately, and in a timely manner commensurate with their safety
significance P.1(a).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
procedures, or drawings. Procedure SO123-XV-50.CAP-1, Writing Nuclear
Notifications for Problem Identification and Resolution, Revision 2, required, in part, All
SONGS employees and supplemental personnel are responsible for promptly
identifying, reporting and documenting problems by writing a Nuclear Notification.
Contrary to the above, between January 4 and March 14, 2010, the inspectors identified
multiple examples where licensee and contractor personnel failed to appropriately enter
identified conditions adverse to quality into the corrective action program, without being
prompted by the inspectors. Because this finding is of very low safety significance and
has been entered into the licensees corrective action program as Nuclear Notifications
NNs 200778816 and 200780926, this violation is being treated as a noncited violation,
consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000361;05000362/2010002-12, Failure to Enter Conditions Adverse to Quality into the
Corrective Action Program.
5. Oversight of Switchyard Work Activities
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of
maintenance personnel to follow Work Order 800195196 and provide appropriate
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Enclosure
oversight to transmission and distribution personnel while performing work in the
electrical switchyard.
Description. In accordance with work order 800195196 and procedure SO123-XV-15.3,
Temporary System Alteration and Restoration, Revision 17, maintenance personnel
were required to provide transmission and distribution personnel with a calibrated torque
wrench, followed by oversight and concurrent verification, to complete steps associated
with torquing bolts on the reserve auxiliary transformer circuit breakers, since these bolts
were designated as critical components. Further, the work order also required
maintenance personnel to perform independent torque verifications on the bolted
connections of the reserve auxiliary transformer circuit breakers.
On February 22, 2010, maintenance personnel were preparing to implement Work Order 800195196 steps for performing the independent torque verification on the reserve
auxiliary transformer circuit breakers. During their preparation, maintenance personnel
determined that transmission and distribution personnel had not been provided with a
calibrated torque wrench, and there had not been oversight and concurrent verification
of the bolt torquing on the reserve auxiliary transformer circuit breakers as required by
the work order. Maintenance personnel subsequently generated Nuclear Notification
NN 200803364 to request engineering input for performing the torque verifications, and
to identify the possibility of rework.
The inspectors reviewed Nuclear Notification NN 200803364 and Work Order 800195196. During their review the inspectors questioned the wording of the nuclear
notification, in that it stated that the work order had not been followed, however, no
actions were identified to correct this condition. Also, the section of the work order that
directed the bolt torquing did not allow the independent verification to be performed
without the concurrent verification having already been performed.
The inspectors questioned licensee personnel as to the purpose of the nuclear
notification, and learned that it had been written to have engineering personnel provide
acceptable torque values since it was possible that the bolts had been torqued to values
that exceeded the values specified in the work order. During these discussions, the
inspectors determined that the licensee intended to continue to use this work order to
perform the independent verification. The inspectors determined that this was
inappropriate since the work order could no longer be performed as written, and as it
was intended.
The inspectors informed the licensee of their concerns, and the licensee entered this
issue into their corrective action program as Nuclear Notification NN 200811993.
Subsequently, the licensee determined that nine of the bolted connections had been
torqued to values that exceeded the values specified in the work order. The licensee
corrected the over torqued bolt condition.
Analysis. The failure to follow work order instructions and provide proper oversight and
concurrent verification to transmission and distribution personnel performing work in the
switchyard was a performance deficiency. The finding is greater than minor because
circumventing procedural requirements, if left uncorrected, would have the potential to
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Enclosure
lead to a more significant safety concern, in that, more risk significant equipment could
be rendered inoperable without the knowledge and approval of appropriate management
or control room personnel. This finding is associated with the Mitigating Systems
Cornerstone. Using the Manual Chapter 0609, Significance Determination Process,
Phase 1 Worksheets, the finding is determined to have a very low safety significance
because the finding: (1) is not a design or qualification issue confirmed not to result in a
loss of operability or functionality; (2) did not represent an actual loss of safety function
of the system or train; (3) did not result in the loss of one or more trains of nontechnical
specification equipment; and (4) did not screen as potentially risk significant due to a
seismic, flooding, or severe weather initiating event. This finding has a crosscutting
aspect in the area of human performance associated with work practices because
maintenance personnel failed to ensure supervisory and management oversight of work
activities, including contractors, such that nuclear safety was supported H.4(c).
Enforcement. Title 10 of the CFR, Part 50, Appendix B, Criterion V, Instructions,
Procedures, and Drawings, requires, in part, that activities affecting quality shall be
prescribed by documented instructions, procedures or drawings, of a type appropriate to
the circumstances and shall be accomplished in accordance with these instructions,
procedures, or drawings. Work Order 800195196, and procedure SO123-XV-15.3,
Temporary System Alteration and Restoration, Revision 17, provided instructions for
performing maintenance on critical components associated with the reserve auxiliary
transformers. Contrary to the above, on February 26, 2010, maintenance personnel
failed to follow work order 800195196, and procedure SO123-XV-15.3, to provide
appropriate oversight of transmission and distribution personnel who were performing
work in the plant switchyard, which resulted in the over torquing of nine bolts on the
reserve auxiliary transformer circuit breakers. Because this finding is of very low safety
significance and has been entered into the licensees corrective action program as
Nuclear Notifications NNs 200803364 and 200811993, this violation is being treated as a
noncited violation consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000361/2010002-13, Failure to Adequately Implement Station Work Order.
4OA3 Event Follow-up (71153)
.1
Event Follow Up
a.
Inspection Scope
The inspectors reviewed the below listed events for plant status and mitigating actions
to: (1) provide input in determining the appropriate agency response in accordance with
Management Directive 8.3, NRC Incident Investigation Program; (2) evaluate
performance of mitigating systems and licensee actions; and (3) confirm that the
licensee properly classified the event in accordance with emergency action level
procedures and made timely notifications to NRC and state/governments, as required.
January 27, 2010, Unit 2, component cooling water surge tank drain down
evolution
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Enclosure
March 17, 2010, Units 2 and 3, review extent of condition inspections for
identification of leaks in schedule 10 piping
Documents reviewed by the inspectors are listed in the attachment.
These activities constitute completion of two inspection samples as defined in Inspection
Procedure 71153-05.
b.
Findings
Introduction. A self-revealing Green noncited violation of Technical Specification 5.5.1.1
was identified for the failure of operations personnel to follow procedures for operating
the component cooling water system.
Description. Prior to the event, both component cooling water surge tank levels were
rising due to intersystem leakage. The problem with intersystem leakage was being
investigated and was eventually discovered to be from a cross tie valve which was not
adequately closed. On January 27, 2010, operations personnel planned to use
procedure SO23-2-17, Component Cooling Water System Operation, Revision 31, to
drain down the component cooling water surge tank. Prior to the drain down evolution,
operations personnel failed to perform an adequate pre-job brief or properly review of
the procedure regarding maintaining pressure since the surge tank draining had become
a routine evolution to compensate for the intersystem leakage. Furthermore, operations
personnel performing the evolution failed to use the proper human error prevention
techniques regarding the change in plant conditions and proceeded with the evolution
without asking for help. Due to time pressures and complacency, operations personnel
proceeded with the assumption that the nitrogen supply valves would maintain the
nitrogen pressure within the required limits during the drain down evolution.
Procedure SO23-2-17 required operations personnel to perform the following steps:
.1 THROTTLE OPEN S2(3)1203MU117, CCW Train A HX E001 CCW (Shell
Side) Drain Valve.
.2 While maintaining CCW surge tank pressure 33-40 psig, LOWER CCW
Surge Tank to the desired level, then CLOSE S2(3)1203MU117, CCW Train A
HX E001 CCW (Shell Side) Drain Valve.
An equipment operator commenced the drain down evolution and opened the
appropriate component cooling water heat exchanger shell drain valves and observed
levels dropped to 60 percent in the surge tanks. The plant was in a refueling outage and
changes to radiological control boundaries prevented the operator from having
immediate access to the surge tank pressure gauge, which was in the next room. The
equipment operator rationalized that the pressure regulator would properly function to
maintain the required pressure band, and decided to continue with the rest of his rounds
before checking pressure. About two hours later, the equipment operator observed
component cooling water train A surge tank pressure was at 30 psig, which was below
the minimum pressure for operability per procedure SO23-2-17. Control room personnel
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Enclosure
were notified and declared the component cooling water train A and associated
shutdown cooling loop inoperable. This required an unplanned entry into Technical
Specification 3.9.5.A, and immediate actions to restore the shutdown cooling loop.
Operations personnel raised the level in the component cooling water train A surge tank
to 65 percent, which increased the surge tank pressure to 34 psig, which was within the
acceptable range. Operations personnel also initiated an immediate investigation and
discovered the nitrogen pressure regulator was not maintaining the proper pressure in
component cooling water surge tank train A.
Analysis. The failure to follow procedures for operating plant equipment was a
performance deficiency. The finding is greater than minor because the continued failure
to follow procedures when operating safety-related plant equipment, if left uncorrected,
would have the potential to lead to a more significant safety concern. The finding is
associated with the Mitigating Systems Cornerstone. Using the Manual Chapter 0609,
Appendix G, Shutdown Operations Significance Determination Process, Phase 1
guidance, the finding is determined to have very low safety significance because the
finding did not result in an increase in the likelihood of a loss of reactor coolant system
inventory, degrade the ability to add reactor coolant system inventory, or degrade the
ability to recover decay heat removal. This finding has a crosscutting aspect in the area
of human performance associated with work practices because operations personnel
failed to use proper human error prevention techniques and proceeded in the face of
unexpected circumstances when operating the component cooling water system
Enforcement. Technical Specification 5.5.1.1 requires, in part, that procedures be
established, implemented, and maintained covering the activities specified in Appendix
A, Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors, of
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operations), Dated
February 1978. Appendix A, Item 3.e, requires procedures for operating the component
cooling water system. Procedure SO23-2-17, Component Cooling Water System
Operation, Revision 31, provided instructions for operating the component cooling water
system. Contrary to the above, on January 27, 2010, operations personnel failed to
follow the requirements of procedure SO123-2-17, while performing a planned drain
down of the component cooling water surge tanks. Specifically, operations personnel,
while draining the component cooling water surge tank, failed to maintain the surge tank
pressure, in accordance with procedure SO23-2-17, such that, component cooling water
surge tank pressure was permitted to go low out of the expected operating range. As a
result of this low surge tank pressure, operators declared the component cooling water
and shutdown cooling train A systems inoperable. Because this finding is of very low
safety significance and has been entered into the licensees corrective action program
as Nuclear Notification NN 200771367, this violation is being treated as a noncited
violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000361/2010002-14, Failure to Follow Operations Procedure to Monitor Component
Cooling Water Surge Tank Pressure.
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Enclosure
4OA6 Meetings
Exit Meeting Summary
On January 6, 2010, the inspector conducted a telephonic exit meeting to present the results of
the in-office inspection of changes to the licensees emergency plan to Mr. B. Ashbrook,
Manager, Onsite Emergency Preparedness. The licensee acknowledged the issues presented.
On March 23, 2010, the inspectors presented the results of the resident inspections to Mr. R.
Ridenoure, Senior Vice President and Chief Nuclear Officer, and other members of the licensee
staff. The licensee acknowledged the issues presented.
The inspectors asked the licensee whether any materials examined during the inspections
should be considered proprietary or sensitive. The inspectors returned or destroyed all
proprietary information reviewed during the inspections and all identified sensitive information
has been returned to the appropriate licensee custodian.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as noncited violations.
.1
Title 10 CFR 50.65(a)(4), states in part, that before performing maintenance activities
(including but not limited to surveillance, post maintenance testing, and corrective and
preventive maintenance), the licensee shall assess and manage the increase in risk that
may result from the proposed maintenance activities. Contrary to the above, on
February 17, 2010, the licensee failed to adequately assess and manage the increase in
risk associated with maintenance activities in the electrical switchyard. Specifically, the
licensee determined that the station had failed to perform an adequate risk assessment
for proposed crane activities in the switchyard with regard to Unit 3, which was operating
at full power. Before allowing the activities to commence the licensee performed the
required risk assessment, and classified the work as a high risk activity in the switchyard
for Unit 3, and commenced the crane activity. This was licensee identified because the
failure to perform a risk assessment was identified by licensee personnel during an
additional final review prior to commencing work. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance
Determination Process flowchart 1, Assessment of Risk Deficit, the finding is
determined to be of very low safety significance because it only involved risk
management actions. The issue was entered into the licensee's corrective action
program as Nuclear Notification NN 200767351.
.2
Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
requires, in part, that activities affecting quality shall be prescribed by documented
instructions, procedures or drawings, of a type appropriate to the circumstances and
shall be accomplished in accordance with these instructions, procedures, or drawings.
Contrary to the above, on December 20, 2009, licensee personnel failed to follow
procedure SO123-XV-50.CAP-1, Writing Nuclear Notifications for Problem Identification
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Enclosure
and Resolution, Revision 2, and enter conditions adverse to quality into the corrective
action program. Specifically, when engineering inspections identified what appeared to
be indications on emergency core cooling system suction piping train A on Unit 3,
operations personnel were not informed, and an operability assessment was not
performed. Subsequently, on January 13, 2010, while performing inspections on the
Unit 3 emergency core cooling system suction piping, engineering personnel again
identified indications and informed operations personnel, which resulted in the piping
being declared inoperable until the ASME code case evaluations could be performed.
This was licensee identified because licensee personnel identified the failure to follow
procedures during follow up investigations. Using the Manual Chapter 0609,
Significance Determination Process, Phase 1 Worksheets, this finding is determined to
have a very low safety significance because the finding: (1) is not a design or
qualification issue confirmed not to result in a loss of operability or functionality; (2) did
not represent an actual loss of safety function of the system or train; (3) did not result in
the loss of one or more trains of non-technical specification equipment; and (4) did not
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event. The issue was entered into the licensee's corrective action program as
Nuclear Notification NN 200756139.
.3
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part,
measures to be established to assure that applicable regulatory requirements and the
design basis, as defined in 10 CFR 50.2 and as specified in the license application, for
those components to which this appendix applies are correctly translated into
specifications, drawings, procedures, and instructions. Contrary to the above, on
February 3, 2010, the licensee failed to appropriately classify a section of emergency
core cooling system mini-flow piping as ASME code class II as specified in the Updated
Final Safety Analysis Report. This was licensee identified because licensee personnel
identified this issue during their reviews. Using the Manual Chapter 0609, Significance
Determination Process, Phase 1 Worksheets, this finding is determined to have a very
low safety significance because the finding: (1) is not a design or qualification issue
confirmed not to result in a loss of operability or functionality; (2) did not represent an
actual loss of safety function of the system or train; (3) did not result in the loss of one or
more trains of non-technical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
The issue was entered into the licensee's corrective action program as Nuclear
Notification NN 200778570.
ATTACHMENT: SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
T. Adler, Manager, Maintenance/Systems Engineering
B. Arbour, Operator Continuing Training Supervisor
J. Armas, Supervisor, Maintenance Engineering Fluid Process
B. Ashbrook, Manager, Emergency Preparedness
D. Axline, Technical Specialist, Nuclear Regulatory Affairs
D. Bauder, Plant Manager
B. Corbett, Manger, Performance Improvement
G. Cook, Manager, Compliance, Nuclear Regulatory Affairs
R. Elsasser, Manger, Training
J. Fee, Manager, Site Emergency Preparedness
S. Gardner, Electrical/System Engineering Manager
M. Graham, Manager, Plant Operations
A. Hochevar, Station Manager, Plant Operations
E. Hubley, Director, Maintenance/Construction
G. Johnson, Jr., Senior Nuclear Engineer, Maintenance/Systems Engineering
K. Johnson, Manager, Design Engineering
L. Kelly, Engineer, Nuclear Regulatory Affairs
D. Spires, Director, Work Control
J. Madigan, Manager, Health Physics
A. Meichler, Mechanical/System Engineering Supervisor
B. MacKissock, Director, Plant Operations
N. Quigley, Manager, Maintenance/System Engineering
R. Richter, Engineering Supervisor, Fire Protection
C. Ryan, Manager, Maintenance & Construction Services
R. St. Onge, Director Nuclear Regulatory Affairs
J. Todd, Manager, Security
D. Wilcockson, Manager of Operations Training
NRC Personnel
D. Loveless, Senior Reactor Analyst
M. Runyan, Senior Reactor Analyst
A-1
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed 05000361/2010002-01
Failure to Implement Fire Protection Plan Requirements
Related to Hot Work Activities (Section 1R05)05000361/2010002-02
Failure to Appropriately Scope Auxiliary Feedwater Pump
Trench Eductors in the Maintenance Rule Monitoring Program
(Section 1R12)05000361/2010002-03
Failure to Enter Operating Experience into Corrective Action
Program for Timely Evaluation (Section 1R13)05000361/2010002-04
Failure to Assess and Manage Risk for Electrical Switchyard
Impacting Maintenance (Section 1R13)05000362/2010002-05
Failure to Follow Procedure Results in an Inadequate
Operability Determination (Section 1R15)05000361/2010002-06
Failure to Perform an Adequate Postmaintenance Test
(Section 1R19)05000361/2010002-07
Failure to Adequately Implement Foreign Material Exclusion
Controls (Section 1R20)05000361/2010002-08
Failure to Appropriately Control Access to a Locked High
Radiation Area (Section 1R20)05000361/2010002-09
Failure to Notify the NRC Within Eight Hours of a
Nonemergency Event (Section 4OA2)05000361/2010002-10
Failure to Report a Safety System Functional Failure (Section
4OA2)05000361/2010002-11
Failure to Obtain a License Amendment for a Technical
Specification Basis Change (Section 4OA2)
A-2
Attachment 05000361/2010002-12
Failure to Enter Conditions Adverse to Quality into the
Corrective Action Program (Section 4OA2)05000361/2010002-13
Failure to Adequately Implement Station Work Order (Section
4OA2)05000361/2010002-14
Failure to Follow Operations Procedure to Monitor
Component Cooling Water Surge Tank Pressure (Section
4OA3)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
PROCEDURES
NUMBER
TITLE
REVISION
SO23-13-8
Severe Weather
7
NUCLEAR NOTIFICATIONS
NUMBER
200498067
200755444
Section 1R04: Equipment Alignment
PROCEDURES
NUMBER
TITLE
REVISION
SO23-3-2.7.2
Safety Injection System Removal/Return to Service
Operation
22
SO2-V-3.12
Attachment 5; Containment Integrated Leakage Rate Test
8
SO23-2-4
Auxiliary Feedwater System Operation
27
A-3
Attachment
SO23-2-13.1
Diesel Generator Alignment
6
SO23-2-8.1
Saltwater cooling System Return to Service Evolution
9
NUCLEAR NOTIFICATIONS
NUMBER
200806892
MAINTENANCE ORDERS
NUMBER
800466402
DRAWINGS
NUMBER
TITLE
REVISION
40112 A and C
P&I Diagram Safety Injection System
23
MISCELLANEOUS
NUMBER
WCD 30005922
Section 1R05: Fire Protection
PROCEDURES
NUMBER
TITLE
REVISION
SO123-XV-1.41
Control of Ignition Sources
14
SO23-XV-4.13
Control of Work and Storage Areas Within the Protected
Area
5
SO123-XIII-
4.600
Fire Protection Impairment
10
A-4
Attachment
SO123-XV-1.41
Control of Ignition Sources
14
NUCLEAR NOTIFICATIONS
NUMBER
200729747
200746059
DRAWINGS
NUMBER
TITLE
REVISION
2-006
SONGS pre-fire plans
6
Section 1R06: Flood Protection Measures
NUCLEAR NOTIFICATIONS
NUMBER
200758566
200409164
200765185
200001761
200760572
200318922
200758652
CALCULATIONS
NUMBER
TITLE
REVISION
M-0120-015
Plant Flood Analysis Review
8
N-4090-009
Units 2&3 Auxiliary Feedwater Pump Room and Doghouse
Pressure Temperature Analysis
0
Section 1R11: Licensed Operator Requalification Program
PROCEDURES
NUMBER
TITLE
REVISION
SO23-15-56
Alarm Response Instruction 56A
8
A-5
Attachment
SO23-13-18
Reactor Protection System Failure
30
SO23-12.1
Standard Post Trip Actions
22
SO23-12-10
Safety Function Status Checks
4
SO123-VIII-10
Emergency Coordinator Duties
26
SO123-VIII-1
Loss of RCS Inventory
29
Section 1R12: Maintenance Effectiveness
PROCEDURES
NUMBER
TITLE
REVISION
SO123-XV-5.3
11
NUCLEAR NOTIFICATIONS
NUMBER
200815548
200409164
200760572
200765185
200318922
200758652
200758566
200001761
200819522
200804181
200815848
MAINTENANCE ORDERS
NUMBER
800078277
MISCELLANEOUS
NUMBER
TITLE
REVISION /
DATE
SONGS System Health Report AFWS 4th Quarter-2009
DBD-SO23-780
Auxiliary Feedwater System
9
A-6
Attachment
STS-SO123-
2001
Maintenance Rule Scoping Matrix
February
23, 2000
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
PROCEDURES
NUMBER
TITLE
REVISION
SO23-XX-8
Integrated Risk Management
3
SO123-I-1.37
Diver Safety During Intake and Forebay Structure Diving
Operations
4
SO23-XX-8
Integrated Risk Management
4
SO23-5-1.8.1
Shutdown Nuclear Safety
23
SO23-12-11
EOI Supporting Attachments
7
SO23-12-8
Station Blackout
21
NUCLEAR NOTIFICATIONS
NUMBER
200155657
200741690
200755444
200789579
200787617
200810952
200818599
200819462
200797351
200402733
200805635
200801929
MAINTENANCE ORDERS
NUMBER
800436397
800074316
A-7
Attachment
CALCULATIONS
NUMBER
TITLE
REVISION
PRACP-10-
0001
PRA Change Package
0
Emergency Diesel Generator Loading
2
125V Battery & DC System Sizing
20
MISCELLANEOUS
NUMBER
TITLE
REVISION /
DATE
NRC
Administrative
Letter 89-10
Dispositioning of Technical Specifications that are
Insufficient to assure Plant Safety
December
28,1998
IPE-HC-075
Operator Action Summary Data Sheet Post-Initiator Human
Error Probability Calculation Worksheet
August 28,
2006
DCP-2&3-
7048.00SE
10 CFR 50.54(x) Unit to Unit Diesel Generator Crosstie
0
Section 1R15: Operability Evaluations
PROCEDURES
NUMBER
TITLE
REVISION
SO23-3-3.31.3
Component Cooling Water Valve Testing - Offline
15
SO123-XV-52
Functionality Assessments and Operability Determinations
14
NUCLEAR NOTIFICATIONS
NUMBER
200791845
200792682
200769743
200745284
200744216
A-8
Attachment
200714391
200744216
200743712
200760570
MAINTENANCE ORDERS
NUMBER
800451952
CALCULATIONS
NUMBER
M-DSC-443
M-DSC-441
MISCELLANEOUS
NUMBER
TITLE
DATE
AR 000201278
February
25, 2000
Section 1R19: Postmaintenance Testing
PROCEDURES
NUMBER
TITLE
REVISION
SO123-XX-5
Work Clearance Application/Work Clearance
Document/Work Authorization Record
28
SO123-II-9.174
Resistance Temperature Detector or thermistor functional
Verification
1
SO2-XXVI-
9.8001.62890.1
Unit 2 boration dilution control system preoperational test
2
SO123-XXVI-
2.5
Preparation, Revision and Approval of Preoperational,
Acceptance and Special Test Procedures
4
SO23-II-20
Ovation Distributed Control System (DCS)
2
A-9
Attachment
NUCLEAR NOTIFICATIONS
NUMBER
NMO800449052
200766430
NMO 800356395
800250944
20683701
200681431
200651946
200651922
200806892
DRAWINGS
NUMBER
TITLE
REVISION
35149
Area 2C6 conduit and tray 30-45 foot elevation
25
MAINTENANCE ORDERS
NUMBER
ECP 800162890
ECP 800390458
MISCELLANEOUS
NUMBER
TITLE
REVISION /
DATE
M37629
Environment qualification Data Sheets
0
N14856B4
Data Sheet 2TE0921X2
January 28,
2009
Section 1R20: Refueling and Other Outage Activities
PROCEDURES
NUMBER
TITLE
REVISION
SO23-XV-2
Troubleshooting Plant Equipment and Systems
5
NUCLEAR NOTIFICATIONS
NUMBER
NMO800448825
200765286
200766808
200796087
200769743
A-10
Attachment
200709732
200765286
200800403
200791630
MISCELLANEOUS
NUMBER
TITLE
REVISION /
DATE
07050054-01
Fire Protection Impairment Form
May 17,
2007
Bechtel QA
Policy No. Q-12
Codes, Standards, and Regulatory Requirements
3
Sample Id
129939
Release of Liquid, Sludge, Slurry, or Sand
February
23, 2010
Section 1R22: Surveillance Testing
PROCEDURES
NUMBER
TITLE
REVISION
SO23-3-3.31.9
RCS Pressure Isolation Valve Testing Hydro Pump Method-
offline
13
SO23-3-3.31.2
ECCS Valve Testing - Offline
11
SO23-XVII-
8.1.1
Visual Inspection of High Pressure Safety Injection System
5
SO23-3-3.60.4
Saltwater Cooling Pump and Valve Testing
11
SO23-3-3.2
Excore Nuclear Instrumentation Calibration
15
SO23-3-3.25
Once a Shift Surveillance Modes 1-4
31
SO23-3-3.30
Inservice Valve Testing Program
20
SO23-5-1.5
Plant Shutdown for Hot Standby to Cold Shut Down
31
A-11
Attachment
SO23-2-13
Diesel Generator Operation
43
SO23-3-3.60.2
LPSI Surveillance Operating Instruction
9
SO23-3-3.60.7
Containment Spray Pump and Valve Testing
12
NUCLEAR NOTIFICATIONS
NUMBER
200791243
200794544
200823123
200827929
200581670
200829333
200835386
200835812
DRAWINGS
NUMBER
TITLE
REVISION
SO23-507-2-1-
623-X2
8 inch Type 9211 Valve Assembly
1
MISCELLANEOUS
NUMBER
TITLE
DATE
Fisher Anomaly
Notice
FAN 88-2
October 11,
1988
S21204MP016
Inservice Pump Test Record
March 21,
2010
Section 4OA3: Event Follow-Up
PROCEDURES
NUMBER
TITLE
REVISION
SY-SO023-G-2
Systems Engineering guideline
3
A-12
Attachment
DRAWINGS
NUMBER
TITLE
REVISION
S2-1204-ML-
001
From Refueling Water Tank T-005 to Line 108 @ VA. 001
10
S2-1204-ML-
002
From Control Valve 2HV-9301 to Line 109
9
S2-1204-ML-
003
Containment Spray Pump P-013 Suction from Containment
Emergency Sump
20
S2-1204-ML-
004
Containment Spray Pump P-013 Suction from Containment
Emergency Sump
20
S2-1204-ML-
008
From Line 004 Containment Emergency Sump to High
Pressure Safety injection Pump P-019
20
S2-1204-ML-
032
From Line 003 Refuel water tank T-006 to Low Pressure
Safety injection Pump P-015
24
S2-1204-ML-
080
From Line 079 Valve 046 to Refueling Water Tank T-005
8
S2-1204-ML-
151
From 2HV-9306 on Line 052 to Line 080 to Refuel Tank T-
006
2
S2-1219-ML-
068
From Refuel water tank T-005 to Refueling Water Tank T-
006
1
S2-1219-ML-
072
From Refuel water tank T-006 to Drain
0
S2-1219-ML-
073
From Refuel water tank T-005 to Drain
0
S2-1219-ML-
107
From Line 080 Safety Injection to Refuel water tank T-005
8
S2-1204-ML-
033
From Line 031 Refuel Water Tank T006 Sys 1204 to LP
Safety Injection Pump
20
S2-1204-ML-
007
HPSI Pump P-017 Suction from Refueling Tank T-005
15
A-13
Attachment
A-14
Attachment
S2-1204-ML-
009
HPSI Pump P-018 Suction from Refueling Tank T-005
16