ML082980701
| ML082980701 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/24/2008 |
| From: | Russ Bywater Region 4 Engineering Branch 1 |
| To: | Ridenoure R Southern California Edison Co |
| References | |
| FOIA/PA-2011-0157 IR-08-010 | |
| Download: ML082980701 (46) | |
See also: IR 05000361/2008010
Text
October 24, 2008
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 COMPONENT
DESIGN BASES INSPECTION REPORT 05000361/2008010 and
Dear Mr. Ridenoure:
On September 3, 2008, the US Nuclear Regulatory Commission (NRC) completed a component
design bases inspection at your San Onofre Nuclear Generating Station Units 2 and 3. The
enclosed report documents our inspection findings. The preliminary findings were discussed on
July 17, 2008, with Mr. Ed Scherer and other members of your staff. After additional in-office
inspection, a final telephonic exit meeting was conducted on September 11, 2008, with you and
other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
The team reviewed selected procedures and records, observed activities, and interviewed
cognizant plant personnel.
Based on the results of this inspection, the NRC has identified six findings that were evaluated
under the risk significance determination process. Violations were associated with all of the
findings. All six of the findings were found to have very low safety significance (Green) and the
violations associated with these findings are being treated as noncited violations, consistent with
Section VI.A.1 of the NRC Enforcement Policy. If you contest any of the noncited violations, or
the significance of the violations you should provide a response within 30 days of the date of
this inspection report, with the basis for your denial, to the US Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional
Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 East Lamar Boulevard,
Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, US Nuclear Regulatory
Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the San Onofre
Nuclear Generating Station.
UNITED STATES
NUCLEAR REGULATORY COMMISSION
R EGIO N I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
Southern California Edison Company
- 2 -
In accordance with Code of Federal Regulations, Title 10, Part 2.390 of the NRC's Rules of
Practice, a copy of this letter and its enclosure will be available electronically for public
inspection in the NRC Public Document Room or from the Publicly Available Records (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web
site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Russell L. Bywater, Chief
Engineering Branch 1
Division of Reactor Safety
Dockets: 50-361;50-362
License: NPF-10
Enclosure:
Inspection Report 05000361/2008010 and 05000362/2008010
w/Attachments:
Attachment 1: Supplemental Information
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
Dr. David Spath, Chief
Division of Drinking Water and
Environmental Management
California Department of Health Services
850 Marina Parkway, Bldg P, 2nd Floor
Richmond, CA 94804
Southern California Edison Company
-3-
Michael J. DeMarco
San Onofre Liaison
San Diego Gas & Electric Company
8315 Century Park Ct. CP21G
San Diego, CA 92123-1548
Director, Radiological Health Branch
State Department of Health Services
P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414
Mayor
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, Esq.
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 92675
A. Edward Scherer
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Mr. Steve Hsu
Department of Health Services
Radiologic Health Branch
MS 7610, P.O. Box 997414
Sacramento, CA 95899-7414
Southern California Edison Company
-4-
Mr. James T. Reilly
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
Chief, Radiological Emergency Preparedness Section
National Preparedness Directorate
Technological Hazards Division
Department of Homeland Security
1111 Broadway, Suite 1200
Oakland, CA 94607-4052
Southern California Edison Company
-5-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@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 (Michael.Hay@nrc.gov)
Senior Project Engineer, DRP/D (Don.Allen@nrc.gov )
SO Site Secretary (Heather.Hutchinson@nrc.gov )
Public Affairs Officer (Victor.Dricks@nrc.gov )
Team Leader, DRP/TSS (Chuck.Paulk@nrc.gov )
RITS Coordinator (Marisa.Herrera@nrc.gov )
Only inspection reports to the following:
Mark Cox, OEDO RIV Coordinator (Mark.Cox@nrc.gov )
ROPreports
SUNSI Review Completed: __KDC_ ADAMS: X Yes
No Initials: _KDC__
X Publicly Available Non-Publicly Available Sensitive
X Non-Sensitive
R:REACTORS/SO 2008010 KDC
EB1
EB1
EB1
EB1
EB1
DRP: D
EB1: C
KClayton; MBloodgood; LEllershaw
PGage
JAdams
MHay;
RBywater
/RA/
/RA/
/RA/
/RA/
/RA/
/RA/
/RA/
10/24/08
10/23/08
10/24/08
10/24/08
10/24/08
10/24/08
10/24/08
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
- 1 -
Enclosure
ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket:
50-361, 50-362
Licenses:
Report Nos.:
05000361/2008010 and 05000362/2008010
Licensee:
Southern California Edison Company (SCE)
Facility:
San Onofre Nuclear Generating Station, Units 2 and 3
Location:
5000 S. Pacific Coast Hwy
San Clemente, California
Dates:
June 23-27, 2008 and July 7-17, 2008 onsite
July 21-Sept 3, 2008 in office inspection
Team Leader:
K. Clayton, Senior Reactor Inspector, Engineering Branch 1
Team:
L. Ellershaw, PE, Senior Reactor Inspector
P. Gage, Senior Operations Engineer
Dr. J. Adams, Reactor Inspector
Accompanying
Personnel:
G. Skinner, Electrical Contractor (Beckman)
C. Baron, Mechanical Contractor (Beckman)
M. Bloodgood, Reactor Inspector (in training)
Dr. D. Reinert, Reactor Inspector, NSPDP (in training)
Approved By:
Russ Bywater, Chief
Engineering Branch 1
- 2 -
Enclosure
SUMMARY OF FINDINGS
IR 05000361/2008010 and 050000362/2008010; June 23-27, 2008, and July 7-17, 2008, onsite
with in office inspection the weeks of July 21-September 3, 2008; San Onofre Nuclear
Generating Station: baseline inspection, NRC Inspection Procedure 71111.21, "Component
Design Basis Inspection."
The report covers an announced inspection by a team of four regional inspectors, two
contractors, and two inspectors in training. Six noncited violations (NCVs) were identified. All
six violations were of very low safety significance. The final significance of most findings is
indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter
(IMC) 0609, "Significance Determination Process (SDP)." Findings for which the 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
Cornerstone: Mitigating Systems
Green. The team identified a noncited violation of 10 CFR 50, Appendix B,
Criterion III, "Design Control," for failure to ensure that plant conditions were
consistent with design calculation inputs and assumptions (rate of established
component cooling water heat exchanger tube plugging). Specifically, there were
no procedures to verify that the periodic heat treatments of the intake tunnel and
intake structure were effective and that the population of shells available for plugging
the component cooling water heat exchangers was consistent with the historical data
used to develop the engineering calculation and operating instruction curves. As a
result, the design basis calculation and operating instructions did not ensure the
capability of the heat exchangers to perform their design function during anomalous
conditions. The licensee has entered this into their corrective action program as
Notification NN 200006369.
This finding is more than minor in that the performance of the component cooling
water heat exchangers is essential in protecting the mitigating systems cornerstone
objective (design control and equipment performance attributes) of ensuring the
availability, reliability, and capability of systems needed to mitigate the
consequences of an accident. Specifically, the existing design analyses did not
adequately demonstrate that the component cooling water heat exchangers would
perform adequately in the event of anomalous tube plugging events and plant
procedures did not ensure that these anomalous events would be detected and
mitigated prior to the heat exchangers being plugged. These deficiencies
represented reasonable doubt regarding the operability of the component cooling
water heat exchangers. Using the Inspection Manual Chapter 0609, "Significance
Determination Process," Phase 1 Worksheets, the finding is determined to have very
low safety significance (Green) because the deficiency did not result in a loss of
safety function of component cooling water Train A for greater than the Technical
Specification allowed outage time. Train B was not adversely affected by this event.
This finding was reviewed for cross-cutting aspects and none were identified since
- 3 -
Enclosure
the performance deficiency is long standing and is not indicative of current licensee
performance (Section 1R21.2.11)
Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion
III, "Design Control," for failure to properly analyze voltage drop in 125 Volts Direct
Current control circuits. Specifically, the licensee failed to consider and analyze the
voltage drop that occurs in control circuit elements such as cables, relay contacts,
and fuses that could result in considerably lower voltage at the devices than is
available at the corresponding distribution panels. The licensee has entered this into
their corrective action program as Notifications NN 200051692 and NN 200059581.
This finding is more than minor because it is associated with the mitigating systems
cornerstone objective (design control attribute) of ensuring the availability and
reliability of safety systems, and closely parallels inspection manual chapter 0612,
Appendix E, Example 3.j, in that there was reasonable doubt regarding the capability
of the 125 Volts Direct Current system to perform its intended function pending
reanalysis. Using the Inspection Manual Chapter 0609, "Significance Determination
Process," Phase 1 Worksheets, the finding is determined to have very low safety
significance (Green) because the 125 Volts Direct Current system was determined to
have sufficient voltage margin to accommodate the additional voltage drop in the
circuit elements that had not been considered. This finding was reviewed for cross-
cutting aspects and none were identified (Section 1R21.2.14.1)
Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion
XVI, "Corrective Action," for failure to identify, evaluate, or correct conditions adverse
to quality. Specifically, in 2007 the licensee failed to recognize, evaluate, or write an
action request when the performance test for Station Battery 2B008 was terminated
early due to test equipment issues. The licensee has entered this into their
corrective action program as Notification NN 200060319.
This finding is more than minor because it is associated with the mitigating systems
cornerstone objective (equipment performance attribute) of ensuring the availability
and reliability of safety systems. Specifically, the failure to verify that battery testing
anomalies are recognized, evaluated, and corrected is a condition adverse to quality
with respect to ensuring that the battery would be capable of performing its design
function. Using the Inspection Manual Chapter 0609, "Significance Determination
Process," Phase 1 Worksheets, the finding is determined to have very low safety
significance (Green) because it was not a design issue resulting in loss of function,
did not represent an actual loss of a system safety function, did not result in
exceeding a Technical Specification allowed outage time, and did not affect external
event mitigation. This finding was reviewed for cross-cutting aspects and none were
identified (Section 1R21.2.14.3).
Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion
V, "Instruction, Procedures, and Drawings," for failure to follow procedures while
performing the battery performance tests. Specifically, on four occasions,
performance tests for Battery 2B008 were terminated early instead of continuing the
tests until reaching one of the test termination criteria in the applicable test
procedure. The licensee has entered this into their corrective action program as
Notification NN 200060319.
- 4 -
Enclosure
This finding was more than minor because it was associated with the mitigating
systems cornerstone (equipment performance attribute) and affected the cornerstone
objective of ensuring the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. Using the
Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1
Worksheets, the finding is determined to have very low safety significance (Green)
because it was not a design issue resulting in loss of function, did not represent an
actual loss of a system safety function, did not result in exceeding a Technical
Specification allowed outage time, and did not affect external event mitigation. This
finding has a cross-cutting aspect in the area of human performance (Work Practices
component) because the licensee did not ensure that appropriate error prevention
techniques were used to avoid deviation from the test termination criteria provided in
test procedures [H.4.(a)] (Section 1R21.2.14.4).
Green. The team identified a noncited violation of Technical Specification 5.5.1.a for
inadequate procedures for 480 Volts Alternating Current system grounds.
Specifically, the procedures do not identify the deleterious effects of 480 Volts
Alternating Current system grounds on connected equipment, or the proper sense of
urgency in removing grounds. Due to inadequate procedures for alarm response
and abnormal operations, the licensee was slow in responding to a ground alarm on
Bus 3B04 in March of 2008. It took 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> to identify and remove the ground.
This indicated a routine, rather than a prompt response and may have exposed
connected equipment to overvoltage for an unnecessarily long period of time. The
licensee has entered this into their corrective action system as Notifications NN
200057494 (addresses trending of ground faults) and NN 200057495 (addresses
procedure change).
This finding was more than minor because the procedure deficiency affected the
mitigating system cornerstone objective (procedure quality attribute) of ensuring
availability, reliability, and capability of systems needed to respond to initiating
events to prevent undesired consequences. Using the Inspection Manual
Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the
finding is determined to have very low safety significance (Green) because the
finding was not a design or qualification deficiency, did not result in a loss of safety
function, and did not screen as potentially risk significant due to external events.
This finding was reviewed for cross-cutting aspects and none were identified
(Section 1R21.2.16).
Cornerstone: Initiating Events
Green. The team identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, "Corrective Actions," for the failure of operations management,
operations training, and engineering to ensure that conditions adverse to quality are
promptly identified and corrected. Specifically, multiple reactivity excursions
occurred in the plant over the past two years, where corrective actions have been
ineffective at addressing blended flow evolutions. The licensee has entered this into
their corrective action program as Notifications NN 200062659 (addresses procedure
change) and NN 200006366 (addresses common cause evaluation).
The finding is more than minor because it is associated with the initiating events
cornerstone (human performance attribute) and affects the associated cornerstone
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Enclosure
objective to limit the likelihood of those events that upset plant stability and challenge
the critical safety functions during shutdown as well as power operations. If left
uncorrected, the conditions would continue to contribute to additional operator errors
or significantly impact the operators ability to perform blended flow evolutions.
Using the Inspection Manual Chapter 0609, "Significance Determination Process,"
Phase 1 Worksheets, the finding is determined to have very low safety significance
(Green) because it did not contribute to both the likelihood of a reactor trip and the
likelihood that mitigating equipment or functions will not be available. This finding
has a crosscutting aspect in the area of problem identification and resolution
associated with the corrective action program because the licensee did not
thoroughly evaluate problems such that resolutions address causes and extent of
condition P.1(c) (Section 4OA2).
B.
Licensee-Identified Violations.
No findings of significance were identified.
- 6 -
Enclosure
REPORT DETAILS
1
REACTOR SAFETY
Inspection of component design bases verifies the initial design and subsequent
modifications and provides monitoring of the capability of the selected components and
operator actions to perform their design bases functions. As plants age, their design
bases may be difficult to determine and important design features may be altered or
disabled during modifications. The plant risk assessment model assumes the capability
of safety systems and components to perform their intended safety function successfully.
This inspectable area verifies aspects of the initiating events, mitigating systems, and
barrier integrity cornerstones for which there are no indicators to measure performance.
1R21 Component Design Bases Inspection (71111.21)
The team selected risk-significant components and operator actions for review using
information contained in the licensees probabilistic risk assessment. In general, this
included components and operator actions that had a risk achievement worth factor
greater than two or a Birnbaum value greater than 1E-6.
a.
Inspection Scope
To verify that the selected components would function as required, the team reviewed
design basis assumptions, calculations, and procedures. In some instances, the team
performed calculations to independently verify the licensee's conclusions. The team
also verified that the condition of the components was consistent with the design bases
and that the tested capabilities met the required criteria.
The team reviewed maintenance work records, corrective action documents, and
industry operating experience records to verify that licensee personnel considered
degraded conditions and their impact on the components. For the review of operator
actions, the team observed operators during simulator scenarios, as well as during
simulated actions in the plant.
The team performed a margin assessment and detailed review of the selected risk-
significant components to verify that the design bases have been correctly implemented
and maintained. This design margin assessment considered original design issues,
margin reductions because of modifications, and margin reductions identified as a result
of material condition issues. Equipment reliability issues were also considered in the
selection of components for detailed review. These included items such as failed
performance test results; significant corrective actions; repeated maintenance; 10 CFR
50.65(a)1 status; operable, but degraded, conditions; NRC resident inspector input of
problem equipment; system health reports; industry operating experience; and licensee
problem equipment lists. Consideration was also given to the uniqueness and
complexity of the design, operating experience, and the available defense in-depth
margins.
The inspection procedure requires a review of 20-30 risk-significant and low design
margin samples in the following categories: components, operator actions, and operating
experience. The sample selection for this inspection was 16 components, five operator
actions, and four operating experience items.
- 7 -
Enclosure
The components selected for review are listed below with a brief description of the
attributes reviewed for that component. The four operating experience samples and
the five operator actions selected for review are located in section 1R21.3 and
section 1R21.4 of this report, respectively.
This report covers a transitional period for San Onofre Nuclear Generating Stations
corrective action document tracking system. During the first portion of the inspection
all corrective action documents were called Action Requests (or ARs). The week of
July 1, 2008, the entire site (all units) transitioned to a new corrective action document
system that utilizes documents called notifications (or NNs). Therefore, both record
types are referenced in this report.
During the week of July 4, 2008, the resident inspectors reviewed documents pertaining
to equipment on the sample selection list provided to them from the Component Design
Basis Inspection (CDBI) team leader and discovered a loose electrical connection Action
Request document with the 2D2 bus (vital 125 volts direct current bus) and its
associated 2B008 battery. An event occurred in March 2008 involving the discharge of
the 2B008 vital battery below its Technical Specification limit that required entry into a 2-
hour Technical Specification Action Statement. While the licensee was preparing to shut
down Unit 2 as part of the Action Statement, the problem was found in the battery's
output breaker to the 2D2 bus. Several of the fasteners that connected the battery bus
to the battery output breaker for this bus were several turns loose. Further review and
questioning by the resident inspector staff at San Onofre during the July 4th week led to
discussions with NRC senior management about the previous Operability Assessment of
this battery and its associated vital bus. The NRC senior regional management staff,
along with headquarters management, decided that the CDBI team could continue the
review of the 2D2 bus and corresponding 2B008 battery as part of the sample in lieu of a
special inspection team because this appeared to be an isolated issue and did not have
generic implications for other equipment. The CDBI team would provide a more detailed
focus on loose connections on these components as well as the remaining 15
components selected for inspection. The CDBI team found approximately 13 AR
documents with loose connection issues, with a majority of these electrical connections
found stripped from being over-tightened or found several turns loose. Several of these
components that had a history of electrical connection issues were extremely risk-
significant, including the Unit 3 Train B Emergency Diesel Generator 3G002, the Unit 2
battery 2B008 and its associated 2D2 bus, and an Emergency Chiller. The CDBI team
called NRC Region 4 to discuss these issues with NRC senior management and the
decision was made to conduct a special inspection at San Onofre for the loose electrical
connection issues. The CDBI team turned over all potential findings regarding loose
electrical connections for the 2D2 bus and 2B008 battery as well as the 13 AR's to the
NRC Region 4 Special Inspection Team on July 30, 2008. This included all potential
findings in training, procedures, and maintenance practices as they related to the 2D2
bus/2B008 battery and the 3G002 Emergency Diesel Generator.
.2
Results of Detailed Reviews for Components:
.2.1 Component Cooling Water (CCW) Pump 2P025:
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Enclosure
a.
Inspection Scope
The team reviewed corrective action documents (listed as Action Requests in the back
of the report as ARs) for this component to look for repeat problems and adequate
corrective actions to repair; verified motive energy source is safety-grade; reviewed
several Inservice Test (IST) results - including procedures, measurement uncertainties
and vendor pump curves - to verify the pump was not degraded and that the actual
pump performance exceeded the operability curve by an amount that exceeds
measurement uncertainties; verified that loss of non-safety grade instrument air would
not compromise the ability of the pump to fulfill its safety function during accident
conditions; verified the pump will meet its startup time requirement, even in the event of
a loss of voltage accident; reviewed permanent plant modification documents to ensure
they were performed in accordance with 10 CFR 50.59 and to look for loose bolt
conditions; conducted a walk-down to verify general equipment conditions and to
examine two permanent plant modifications on this pump.
b.
Findings
No findings of significance were identified.
.2.2
Safety Injection Tank 2SIT-10:
a.
Inspection Scope
The team reviewed AR documents to determine whether issues involving inadvertent
pressure fluctuations are adequately addressed; reviewed pressure and level
instrumentation responses and calibrations and pressure/level surveillance procedures
to ensure the tank will adequately perform its safety function in the case of a Loss-of-
Coolant Accident (LOCA) - compared against assumptions in the LOCA analysis - to
ensure important-to-safety parameters are periodically verified and that measurement
uncertainties are adequately accounted for; reviewed calculations to verify tank sizing
(minimum and maximum liquid volumes and minimum and maximum pressures) and
level transducer uncertainties.
b.
Findings
No findings of significance were identified.
.2.3 Main Feedwater Isolation Valve 2HV4048:
a.
Inspection Scope
The team reviewed AR documents to determine whether issues are being adequately
addressed and corrected; reviewed calculations relating IST test acceptance criteria with
Design Basis Accident assumptions and compared Design Basis Documents, Updated
Safety Analysis Report, and Technical Specification values with those used in the
calculations; reviewed procedures and results from IST testing to verify that this
component is not degrading; reviewed the Design Basis Document to understand the
safety function for this component; reviewed the permanent plant modification
documents to ensure that they were performed in accordance with 10 CFR 50.59 and to
look for loose bolt conditions.
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Enclosure
b.
Findings
No findings of significance were identified.
.2.4 Atmospheric Dump Valve 2HV8419:
a.
Inspection Scope
The team reviewed AR documents to determine whether issues are being adequately
addressed and corrected; reviewed IST procedures and test results to ensure this valve
is not degrading and is able to fulfill its safety function; reviewed the accumulator sizing
calculations to ensure the accumulator has sufficient capacity and is maintained
adequately to provide the motive force for the 8-hour safety function requirement.
b.
Findings
No findings of significance were identified.
.2.5
Main Steam Isolation Valve 2HV8204:
a.
Inspection Scope
The team reviewed Technical Specifications, Updated Final Safety Analysis Report,
Design Basis Documents, calculations, design drawings, and plant procedures to verify
the appropriateness of design assumptions, boundary conditions, and models. This
review was also conducted to verify that the licensees analytical methods were
appropriate. The team verified that design assumptions and limitations were translated
to operational and testing procedures. IST data (i.e., stroke test closed, fail safe test
closed, power operated valve non-timed stroke exercise, and position indicator test) was
reviewed. Plant personnel were interviewed and a component walk down was
conducted to verify that potential degradation was being monitored or prevented. The
walk down also verified that the observable material condition would support the design
operation, component configuration was being maintained consistent with design
assumptions, and the equipment was adequately protected from external events.
The team also reviewed operating experience history, maintenance history, and
corrective action history to verify that potential degradation was being monitored or
prevented and that component replacement was consistent with qualification life.
b.
Findings
No findings of significance were identified.
.2.6
High Pressure Safety Injection (HPSI) Pump S21204MP019:
a.
Inspection Scope
The team reviewed HPSI Pump S21204MP019, Unit 2, Train B, system hydraulic, Net
Positive Suction Head (NPSH), and minimum performance calculations. The team
reviewed surveillance test acceptance criteria bases and test results to verify that the
pumps had sufficient capacity at the minimum acceptable performance. The team
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Enclosure
verified that the pump had adequate protection for potential minimum flow and run-out
conditions. The team reviewed corrective action documents and maintenance
associated with the equipment to verify that degraded conditions were appropriately
addressed. The team reviewed operating experience associated with this component to
verify that the information was adequately addressed. The team reviewed component
modifications, engineering design changes, and field change notices to verify that the
performance capability of the equipment was not degraded due to component
alterations. The team reviewed operating procedures and control logic associated with
the pump and associated equipment to verify that the equipment was capable of
performing the design function.
b.
Findings
No findings of significance were identified.
.2.7
Emergency Chiller SA1513ME335:
a.
Inspection Scope
The team reviewed Emergency Chiller SA1513ME335, Loop B, heat capacity,
performance, and set-point calculations to verify that the equipment was capable of
performing the design function. The team reviewed surveillance test acceptance criteria
bases and test results to verify that the chiller had sufficient cooling capacity. The team
reviewed corrective action documents and maintenance associated with the equipment
to verify that degraded conditions were appropriately addressed. The team reviewed
drawings and control schematics to verify that the component control logic is consistent
with design bases. The team reviewed operating procedures and control logic to verify
that the equipment was capable of performing the design function. The team reviewed
component modifications, engineering design changes, and field change notices to verify
that the performance capability of the equipment was not degraded due to component
alterations. The team performed a walk down of the emergency chiller and associated
components to verify that the material condition and configuration of the equipment was
consistent with design requirements.
b.
Findings
No findings of significance were identified.
.2.8
Refueling Water Tanks, T005 & T006:
a.
Inspection Scope
The team reviewed the Updated Safety Analysis Report, Design Basis Documents,
selected drawings, calculations, maintenance records, and operating procedures to
verify the capability of the tanks to perform their intended function during design basis
events. The team reviewed various calculations to evaluate the inventory, instrument
uncertainty, and transfer set-point of the tanks. The team reviewed the vortex limit and
NPSH calculations for the pumps related to the tanks to verify adequate water level prior
to transfer to the containment sumps and that adequate water would be transferred to
the containment sump. The team also reviewed operating procedures related to the
tanks to ensure they were consistent with the design basis.
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Enclosure
b.
Findings
No findings of significance were identified.
.2.9
Saltwater Cooling (SWC) Pump, 3P307:
a.
Inspection Scope
The team reviewed the Updated Safety Analysis Report, Design Basis Documents,
selected drawings, calculations, maintenance records, and operating procedures to
verify the capability of the pump to perform its intended function during design basis
events. The team reviewed SWC system flow calculations and system test acceptance
criteria and results to evaluate the capability of the pump to provide the required flow to
the CCW heat exchanger under the most limiting accident conditions. The team
reviewed the calculations and procedures related to the periodic backwash and heat
treatment of the SWC to verify adequate SWC would be available whenever the system
was considered operable. The team also reviewed operating procedures related to the
pump to ensure that they were consistent with the design basis calculations and the
licensing basis. The team also reviewed alternating current flow and voltage
calculations to determine whether adequate motive power was available to start and run
the pump during worst case degraded voltage and service conditions. The team
reviewed maintenance and corrective action documents to determine if the equipment
has exhibited adverse performance trends.
b.
Findings
No findings of significance were identified.
.2.10 Containment Emergency Sump Motor Operated Valve (MOV), 3HV9303:
a.
Inspection Scope
The team reviewed the Updated Safety Analysis Report, Design Basis Documents,
selected drawings, calculations, maintenance records, and operating procedures to
verify the capability of the MOV to perform its intended function during design basis
events. The team reviewed Generic Letter 89-10 calculations to evaluate the capability
of the valve to change position as required under the most limiting accident conditions.
The team reviewed the calculations to verify that the most limiting system operating
conditions were considered in the calculations, including the potential to pressurize the
pipe between the normally closed sump isolation valves. The team reviewed electrical
calculations to verify the appropriate voltage values were included in the valve
calculations. The team also reviewed operating procedures related to the valve to
ensure they were consistent with the design basis calculations and the licensing basis.
The team reviewed alternating current flow and voltage calculations to determine if
adequate motive power was available during worst case degraded voltage and service
conditions. The team reviewed motor control center control circuit voltage drop
calculations to determine whether MOV contactors had adequate voltage to pick up
when required. The team reviewed elementary wiring diagrams to determine whether
control logic was in conformance with the design bases.
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Enclosure
b.
Findings
No findings of significance were identified.
.2.11 Component Cooling Water (CCW) Heat Exchanger, 2E-002:
a.
Inspection Scope
The team reviewed the Updated Safety Analysis Report, Design Basis Documents,
selected drawings, calculations, maintenance records, and operating procedures to
verify the capability of the heat exchanger to perform its intended function during design
basis events. The team reviewed CCW thermal performance calculations and heat
exchanger test acceptance criteria and results to evaluate the capability of the heat
exchanger to maintain the required CCW system supply temperature under the most
limiting accident conditions. The team reviewed the calculations and procedures related
to the periodic backwash and heat treatment of the heat exchanger tubes to verify
adequate SWC flow and heat transfer capability would be available whenever the
system was considered operable. The team also reviewed operating procedures related
to the heat exchanger to ensure they were consistent with the design basis calculations
and the licensing basis.
b.
Findings
Introduction: The team identified a Green noncited violation (NCV) of 10 CFR 50,
Appendix B, Criterion III, "Design Control," for failure to ensure that plant conditions were
consistent with design calculation inputs and assumptions (rate of established CCW heat
exchanger tube plugging). Specifically, there were no established procedures to verify
that the periodic heat treatments of the intake tunnel and intake structure were effective
and that the population of shells available for plugging the CCW heat exchangers was
consistent with the historical data used to develop the engineering calculation and
operating instruction curves. As a result, the design basis calculation and operating
instructions did not ensure the capability of the heat exchangers to perform their design
function during anomalous conditions.
Description: The SWC system was designed to provide cooling water to the CCW heat
exchangers under both normal and post-accident conditions. During normal operation,
the operability of the SWC system was verified, in part, based on a set of curves
included in operating instructions SO23-2-8, "Saltwater Cooling System Operation,"
revision 29. These curves provided the operators with the minimum required SWC flow
as a function of SWC temperature and CCW heat exchanger differential pressure during
normal operation, and with SWC flow as a function of SWC temperature during reverse
(backwash) flow operation. During normal operation, the operators verified that the
required SWC operating conditions were met and normally initiated backwash of the
CCW heat exchangers when the operational limits were approached. The SWC
operating instructions also allowed the SWC system to be operable during reverse
(backwash) flow operation if the required SWC operating conditions (temperature and
flow) were met. To initiate backwash flow, the operators were directed to stop the
applicable SWC pump, change manual valve positions, and restart the SWC pump.
After operating with backwash flow for some period of time, the process was reversed to
restore the normal system configuration. In addition, the plant design did not include
instrumentation to monitor CCW heat exchanger differential pressure during backwash
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Enclosure
operation and the operating procedures did not include any specific time limits for the
SWC system to be operable under reverse (backwash) flow operation.
The operating instruction curves were developed by engineering calculation
M-0027-023, "CCW/SWC Heat Exchanger Operability," revision 0 (including calculation
change notices through CCN-10, February 22, 2008). This calculation determined the
minimum SWC flows that would be required to maintain the maximum CCW system
supply temperature limit under design basis accident conditions. The calculation
considered the performance of the CCW heat exchanger under degraded conditions,
assuming that various percentages of the heat exchanger tubes were plugged with
debris from the SWC system. The percentage of plugged tubes was correlated to the
pressure differential across the tube side of the heat exchanger at various SWC flows.
The calculation assumed that an accident could occur when the SWC operating limits
had been reached and that the CCW heat exchanger differential pressure would
continue to increase during the accident due to additional tube plugging. The assumed
rate of tube plugging was based on a review of historical CCW heat exchanger
differential pressure data, the maximum rate of tube plugging was extracted from
approximately six years of actual plant data and applied to the limiting accident
conditions. The resulting curves were intended to represent the limiting SWC "starting
points" that provide acceptable performance for a design basis accident with a
postulated single failure.
The inspection team questioned if these operating instruction curves would ensure the
capability of the SWC system to perform its design function under the most limiting
conditions; specifically, the team was concerned that the predicted rate of tube plugging
used in calculation M-0027-023 might not bound any anomalous conditions that were not
representative of the historical plant data. Anomalous conditions would include an
unusually high rate of tube plugging, such as experienced on June 3, 2008, when
unexpected rapid fouling of unit 2, train A CCW heat exchanger occurred. Historically,
the primary cause of CCW heat exchanger tube plugging was clam shells that were
ingested by the SWC pumps and were of sufficient size to plug the 3/4-inch CCW heat
exchanger tubes. The plant's strategy to minimize tube plugging included periodic heat
treatment of the intake tunnel and intake structure to kill the majority of the clams while
they were still less than approximately 3/8-inch long. It was expected that some amount
of these shells would carry over the traveling screens and be ingested by the SWC
pumps. However, the majority of these small shells were expected to pass through the
CCW heat exchangers without causing significant plugging. The team's concern was
that, in the event of an ineffective heat treatment, a significant population of larger clams
(3/8-inch to 1-inch long) might be allowed to grow prior to the next scheduled heat
treatment. In that case, the subsequent heat treatment could result in more rapid
plugging of the heat exchangers due to the larger shells. As discussed in Apparent
Cause Evaluation (ACE) 080600076-1, a heat treatment of this intake structure was
performed on March 30, 2008 and a subsequent heat treatment was performed on
May 31, 2008. On June 3, 2008 and following days, this heat exchanger experienced
rapid plugging. Subsequent inspections identified a population of larger shells in the
intake structure, inferring that the March 30, 2008 heat treatment was not fully effective.
The inspection team determined that there were no specific procedures to verify that the
periodic heat treatments of the intake tunnel and intake structure were effective and that
the population of shells available for plugging the CCW heat exchangers was consistent
with the historical data used to develop the engineering calculation and operating
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Enclosure
instruction curves. As a result, there was a potential for anomalous conditions that were
not bounded by the subject calculation and operating instruction.
Analysis: The failure to assure that plant conditions were consistent with design
calculation inputs and assumptions (rate of established CCW heat exchanger tube
plugging) is a performance deficiency. This finding is more than minor in that the
performance of the CCW heat exchangers is essential in protecting the mitigating
systems cornerstone objective (design control and equipment performance attributes) of
ensuring the availability, reliability, and capability of systems needed to mitigate the
consequences of an accident. Specifically, the existing design analyses did not
adequately demonstrate that the CCW heat exchangers would perform adequately in the
event of anomalous tube plugging events and plant procedures did not ensure that these
anomalous events would be detected and mitigated prior to the heat exchangers being
plugged. These deficiencies represented reasonable doubt regarding the operability of
the CCW heat exchangers. Using the Inspection Manual Chapter (IMC) 0609,
"Significance Determination Process," Phase 1 Worksheets, the finding is determined to
have very low safety significance (Green) because the deficiency did not result in a loss
of safety function of CCW Train A for greater than the Technical Specification allowed
outage time. Train B was not adversely affected by this event. This finding was
reviewed for cross-cutting aspects and none were identified since the performance
deficiency is long standing and is not indicative of current licensee performance.
Enforcement: Title 10 CFR 50, Appendix B, Criterion III, "Design Control," states, in
part, that measures be established to assure that applicable regulatory requirements and
the design basis, as defined in Section 50.2, are correctly translated into procedures and
instructions. Contrary to the above, the licensee failed to ensure that plant conditions
were consistent with design basis analyses. Specifically, the licensee failed to ensure
that the CCW heat exchangers would perform adequately in the event of anomalous
tube plugging events because plant procedures did not ensure that these anomalous
events would be detected and mitigated prior to the heat exchangers being plugged.
Because this finding is of very low safety significance and was entered into the
licensees corrective action program as Notification 200006369 (AR 080600076), this
violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy: NCV 050000361, 050000362/2008010-01, "Inadequate Design
Control for Design Basis of CCW Heat Exchangers."
.2.12 Turbine Driven Auxiliary Feedwater Pump Steam Inlet MOV 3HV4716:
a.
Inspection Scope
The team reviewed the USAR, Design Basis Documents, selected drawings,
calculations, maintenance records, and operating procedures to verify the capability of
the motor operated valve to perform its intended function during design basis events.
The team reviewed generic letter 89-10 calculations to evaluate the capability of the
valve to change position as required under the most limiting accident conditions. The
team reviewed the calculations to verify that the most limiting system operating
conditions were considered in the calculations. The team reviewed electrical
calculations to verify the appropriate voltage values were included in the valve
calculations. The team also reviewed operating procedures related to the valve to
ensure they were consistent with the design basis calculations and the licensing basis.
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Enclosure
b.
Findings
No findings of significance were identified.
.2.13 Emergency Diesel Generator 3G002:
a.
Inspection Scope
The team reviewed static loading calculations to determine whether the maximum
automatic and manual load expected during worst case accident conditions was within
the specified ratings of the diesel generators. The team reviewed load sequencing logic
and dynamic loading calculations to determine whether the transient loading expected
during worst case conditions was within the capability of the diesel generators. The
team reviewed Emergency Diesel Generator testing procedures and results to determine
whether they were consistent with licensing basis requirements, and whether they
demonstrate adequate performance. The team reviewed permissible frequency
variations to determine whether they have been properly accounted for in pump
performance and diesel loading calculations. The team reviewed maintenance and
corrective action documents to determine whether the equipment has exhibited adverse
performance trends. The team performed a visual inspection of the Emergency Diesel
Generators to assess materiel condition and the presence of hazards.
b.
Findings
No findings of significance were identified.
.2.14 125VDC Battery 2B008 and 125VDC Distribution Panel 2D2:
a.
Inspection Scope
The team reviewed battery sizing and voltage drop calculations to determine whether the
battery would have sufficient capacity and capability to supply its design loads during
accident and SBO scenarios. The team reviewed voltage drop calculations to determine
whether loads had sufficient voltage to operate when required. The team reviewed
battery surveillance test procedures and completed surveillances to determine whether
tests were being performed in accordance with Technical Specifications and applicable
IEEE standards, and whether the acceptance criteria was consistent with design
calculations. The team reviewed vendor manuals, maintenance procedures, completed
Maintenance Orders (MO) to determine whether maintenance was performed in
accordance with vendor recommendations. The team reviewed system health,
maintenance, and corrective action documents to determine whether the equipment has
exhibited adverse performance trends. The team performed a visual inspection of
batteries, the distribution panels, and their environs to assess material condition, and the
presence of hazards.
b.
Findings
1. Failure to Correctly Analyze 125VDC Control Circuit Voltage Drop
Introduction: The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion III,
"Design Control," for failure to properly analyze voltage drop in 125VDC control circuits.
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Enclosure
Specifically, the licensee failed to consider and analyze the voltage drop that occurs in
control circuit elements such as cables, relay contacts, and fuses that could result in
considerably lower voltage at the devices than is available at the corresponding
distribution panels.
Description: Calculation E4C-017 included voltage acceptance criteria of 100 VDC at
certain 125 VDC control panels based on the ratings of relays and other devices in the
panels. The calculation did not consider or analyze the voltage drop in control circuit
elements, including cables, relay contacts and fuses downstream of the panels that
could result in considerably lower voltage at the devices than is available at the panel.
In some cases the minimum voltage required by devices was actually 100 VDC, so the
acceptance criterion was inadequate to assure their operability. In response to the
inspectors concerns, the licensee performed preliminary calculations to show that all
devices would have adequate voltage based on actual minimum expected panel
voltages and estimated circuit lengths.
Analysis: The licensees failure to consider the voltage drop control circuit elements was
a performance deficiency. This finding is more than minor because it is associated with
the mitigating systems cornerstone objective (equipment performance attribute) of
ensuring the availability and reliability of safety systems, and closely parallels IMC 0612,
Appendix E, Example 3.j, in that there was reasonable doubt regarding the capability of
the 125 VDC system to perform its intended function pending reanalysis. Using the
Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1
Worksheets, the finding is determined to have very low safety significance (Green)
because the 125 VDC system was determined to have sufficient voltage margin to
accommodate the additional voltage drop in the circuit elements that had not been
considered. This finding was reviewed for cross-cutting aspects and none were
identified.
Enforcement: Title 10 CFR 50, Appendix B, Criterion III, "Design Control," states, in
part, that design control measures be established and implemented to assure that
applicable regulatory requirements and the design basis for structures, systems, and
components are correctly translated into specifications, drawings, procedures, and
instructions. Contrary to the above, the licensee failed to implement applicable design
bases for the 125 VDC control circuitry. Specifically, the licensee failed to consider or
analyze the voltage drop in control circuit elements, including cables, relay contacts and
fuses downstream of 125 VDC distribution panels. Because this finding is of very low
safety significance and was entered into the licensees corrective action program as
Notifications NN 200051692 and NN 200059581, this violation is being treated as a
NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000361,05000362/2008010-02 "Inadequate Design Control for 125 VDC Control Circuits."
2. Omission of Station Black-Out (SBO) Profile during Battery Service Tests
Introduction: The team identified an unresolved item (URI) associated with Technical
Specification Surveillance 3.8.4.7 for omission of the SBO profile (and corresponding
test duration of 240 minutes) during the battery service test. The calculation for battery
size and minimum battery voltage clearly indicates that the SBO condition is more
limiting than the Loss of Voltage Signal/Safety Injection Actuation Signal (LOVS/SIAS)
condition.
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Enclosure
Description: Service test Procedure SO123-I-2.5 used to satisfy Technical Specification
Surveillance 3.8.4.7 is based on the LOVS/SIAS profile instead of the more limiting SBO
profile. The USAR, section 8.3.2.1.2.1, states that the blackout duty cycles for Batteries
A and B can be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per Battery Sizing Calculation E4C-017. Technical
Specification 3.8.4.7 requires a battery service test to verify battery capacity is adequate
to supply and maintain in operable status the required emergency loads for the design
duty cycle. The Institute of Electrical and Electronic Engineers (IEEE) 450-1980 requires
the discharge rate and test length for the service test to correspond as closely as
possible to the battery duty cycle. Calculation E4C-017 analyzed a 90 minute duty cycle
for the LOV/SIAS scenario and a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> duty cycle for the SBO scenario. The load
profile required by procedure SO123-I-2.5 requires a discharge rate of 471 Amps for the
first minute, 173 Amps for minutes 2 through 89, and 252 Amps for the last minute. The
Battery 2B008 duty cycle determined in Calculation E4C-017 for the SBO scenario was
341 Amps for the first minute, 193 Amps for minutes 2 through 29 minutes, 234 Amps for
the 30th minute, 155 Amps for minutes 31 through 239, and 238.95 Amps for the last
minute. Although the profiles are not directly comparable, Calculation E4C-017
demonstrated that the SBO profile was the more limiting profile for battery sizing,
requiring an uncorrected size of 5.01 positive plates per cell vs. 4.60 for the LOV/SIAS
profile. In addition, the calculation showed that the SBO profile was more limiting with
respect to minimum battery voltage with an expected minimum voltage of 106.72V for
the SBO profile vs. 108.81V for the LOV/SIAS profile. The licensee initiated Notification
NN 200061041 to address this issue. This issue was opened to determine if the
SONGS current licensing basis requires the performance of a service test to
demonstrate the capability of the batteries to complete all design duty cycles (including
SBO) defined in the USAR. Pending completion of this determination by the NRC, this
issue is identified as URI 05000361 and 050000362/2008010-03, "Omission of Station
Blackout Profile During Battery Service Tests."
3. Inadequate Corrective Actions for Battery Performance Test Issues
Introduction: The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI,
"Corrective Action," for failure to identify, evaluate, or correct conditions adverse to
quality. Specifically, in 2007 the licensee failed to recognize, evaluate, or write an AR for
issues that prevented completion of the performance test for Station Battery 2B008 due
to test equipment issues.
Description: San Onofre Nuclear Generating Station's (SONGS) battery performance
tests were required to be performed in accordance with Maintenance Procedure SO123-
1-2.6. The procedure provided three test termination criteria in step 6.3.8, which
included battery overall shut down voltage reached, battery cell(s) temperature
exceeding 110°F, or any battery intercell connection showing evidence of excessive
heating. The team noted that the performance test performed on the Unit 2 station
battery 2B008 on January 23, 2007, was terminated because of "load bank not
maintaining load." This was not one of the test termination criteria listed in the
procedure but the test was marked "Sat" (satisfactory). The licensee failed to recognize,
evaluate, or write an AR at that time to document the apparent test equipment failure, or
to assess whether the battery performance test needed to be repeated. The team noted
that the actual test results showed that the discharge rate and duration of the test were
sufficient to establish a battery capacity of at least 100 percent of its ratings. Because
the battery is considered operable if its capacity is over 80 percent, operability criteria
did not appear to have been violated.
- 18 -
Enclosure
Analysis: The team determined that the licensees failure to identify, evaluate, or correct
conditions adverse to quality was a performance deficiency that was reasonably within
their ability to foresee and prevent. Specifically, in 2007, the licensee failed to
recognize, evaluate, or take any action when the performance test for Station Battery
2B008 was terminated early due to test equipment issues. This finding was more than
minor because it was associated with the mitigating systems cornerstone (equipment
performance attribute) and affected the cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. Using the Inspection Manual Chapter 0609, "Significance
Determination Process," Phase 1 Worksheets, the finding is determined to have very low
safety significance (Green) because it was not a design issue resulting in loss of
function, did not represent an actual loss of a system safety function, did not result in
exceeding a TS allowed outage time, and did not affect external event mitigation. This
finding has a cross-cutting aspect in the area of human performance (Work Practices
component) because the licensee did not ensure that appropriate error prevention
techniques were used to avoid deviation from the test termination criteria provided in test
procedures H.4.a].
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
requires, in part, that conditions adverse to quality are promptly identified and corrected.
Contrary to this requirement, as of January 23, 2007, the licensee failed to identify,
evaluate, or correct conditions adverse to quality involving a test equipment failure that
resulted in the early termination of a required battery performance test. Because this
violation is of very low safety significance and has been entered into the licensees
corrective action program as Notification NN 200060319, this violation is being treated
as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy:
NCV 05000361/2008010-04, "Inadequate Corrective Actions for Battery Performance
Tests Issues."
4. Failure to Follow Procedures During the Battery Performance Tests
Introduction: The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion V,
"Instruction, Procedures, and Drawings," for failure to follow procedures while performing
the battery performance tests. Specifically, on four occasions, performance tests for the
Unit 2 station battery 2B008 were terminated early, instead of continuing the tests until
reaching one of the test termination criteria in the applicable test procedure.
Description: SONGS battery performance tests were required to be performed in
accordance with Maintenance Procedure SO123-1-2.6. The procedure provided three
test termination criteria in step 6.3.8, which included 1) battery overall shut down voltage
reached, 2) battery cell temperature exceeding 110°F, or 3) any battery intercell
connection showing evidence of excessive heating. The team noted that performance
tests performed on the Unit 2 station battery 2B008 in 2002, 2006, 2007, and 2008, were
terminated before the minimum battery voltage was reached and without meeting either
of the other two termination criteria. The Maintenance Orders for tests performed in
2002 and 2006 stated that they were terminated early because cell #14 approached
reversal voltage. The 2007 test was terminated early due to "load bank not maintaining
load." The 2008 test was terminated at four hours without further explanation.
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Enclosure
The team noted that none of the conditions encountered during the 2002, 2006, and
2007 tests necessitated early termination since IEEE Standard 450-1980 and step 6.3.1
of the procedure permit temporary interruptions of the tests, during which the conditions
noted could have been addressed in order to complete the tests.
Technical Specification 3.8.4.8 requires the battery performance test to be conducted
every 60 months until the battery shows evidence of deterioration, or has reached 85
percent of the expected life, at which time the interval becomes 12 months. The team
noted that the 85 percent service life point for Battery 2B008 was reached in October
2006, and that annual performance tests have been performed since then. Technical
Specifications Bases and IEEE Standard 450-1980, state that degradation is indicated
when the battery capacity drops by more than 10 percent relative to its capacity on the
previous performance test, or when it is below 90 percent of the manufacturers rating.
The 10 percent criteria is based on measured battery capacity at its minimum voltage
(fully discharged), compared with the prior performance test. By ending the capacity
tests prior to reaching battery minimum voltage, it was not possible to perform
quantitative measurement of battery degradation in accordance with IEEE-450-1980.
Therefore, since at least 2002, the licensee has been unable to quantitatively evaluate
the technical specification testing frequency based on the 10 percent degradation
criteria. The team noted that sufficient data was available in the existing test reports to
determine that battery capacity was at least 100 percent, so the increased test frequency
does not appear to have been improperly delayed. Since the battery is considered
operable if its capacity is over 80 percent and measured capacity for the tests in
question were at least 100 percent, operability criteria have not been violated.
Analysis: The team determined that the licensees failure to perform station battery
capacity testing in accordance with station procedures was a performance deficiency
that was reasonably within their ability to foresee and prevent. Specifically, the licensee
terminated battery capacity tests before reaching the battery minimum average voltage
per cell, as specified by IEEE Standard 450-1980, or encountering any of the other test
termination criteria. This finding was more than minor because it was associated with
the mitigating systems cornerstone (equipment performance attribute) and affected the
cornerstone objective of ensuring the availability, reliability, and capability of systems
that respond to initiating events to prevent undesirable consequences. Traditional
enforcement does not apply because the finding did not have any actual safety
consequences or potential for impacting the NRC's regulatory function, and was not the
result of any willful violation of NRC requirements. Using the IMC 0609, "Significance
Determination Process," Phase 1 Worksheets, the finding is determined to have very low
safety significance (Green) because it was not a design issue resulting in loss of
function, did not represent an actual loss of a system safety function, did not result in
exceeding a TS allowed outage time, and did not affect external event mitigation. This
finding has a cross-cutting aspect in the area of human performance (Work Practices
component) because the licensee did not ensure that appropriate error prevention
techniques were used to avoid deviation from the test termination criteria provided in test
procedures H.4.a].
Enforcement: Title 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and
Drawings," states, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures, and drawings of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions and
procedures. SONGS Procedure SO123-1-2.6 provided termination criteria for
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Enclosure
performance tests to enable a consistent method of measuring battery degradation.
Contrary to the above, on June 10, 2002, on February 18, 2006, on January 23, 2007,
and on February 20, 2008, battery performance testing was not accomplished in
accordance with the required procedure in that the testing was terminated prior to
reaching battery minimum voltage, or other allowable termination criteria specified in the
procedure. Because this violation is of very low safety significance and has been
entered into the licensees corrective action program as Notification NN 200060319, this
violation is being treated as a NCV consistent with Section VI.A.1 of the NRC
Enforcement Policy: NCV 05000361/2008010-05, "Failure to Follow Procedures during
the Battery Performance Tests."
2.15 Reserve Auxiliary Transformer 2XR2:
a.
Inspection Scope
The team reviewed alternating current load flow calculations to determine whether the
transformer had sufficient capacity to support its required loads under worst case
accident loading and grid voltage conditions. The team reviewed maintenance
procedures and records to determine whether maintenance was adequate to assure
operability of automatic functions during accident conditions. The team reviewed system
health and corrective action documents to assess any adverse equipment operating or
maintenance trends.
b.
Findings
No findings of significance were identified.
.2.16 480 Volt Alternating Current (VAC) Load Center 3BO4:
a.
Inspection Scope
The team reviewed alternating current load flow calculations to determine whether the
bus was loaded within its ratings under accident and degraded voltage conditions.
The team reviewed undervoltage protection calculations and logic to determine whether
connected loads were adequately protected. The team reviewed alarm response
procedures to determine whether they were adequate to address abnormal conditions.
The team reviewed vendor technical manuals and maintenance procedures and records
to determine whether maintenance was performed in accordance with vendor
recommendations. The team reviewed system health and corrective action documents
to assess any adverse equipment operating or maintenance trends.
b.
Findings
Inadequate Alarm Response Procedures for 480 VAC Grounds
Introduction: The team identified a Green NCV of Technical Specification 5.5.1.a for
inadequate procedures for 480 VAC system grounds. Specifically, the procedures do
not identify the deleterious effects of 480 VAC system grounds on connected equipment,
or the proper sense of urgency in removing grounds. Due to inadequate procedures for
alarm response and abnormal operations, the licensee was slow in responding to a
ground alarm on Bus 3B04 in March, 2008. It took 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> to identify and remove the
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Enclosure
ground. This indicated a routine, rather than a prompt response and may have exposed
connected equipment to overvoltage for an unnecessarily long period of time.
Description: The 480 VAC electrical distribution system at SONGS is ungrounded.
Ungrounded three phase electrical systems are capable of providing continuity of service
in the presence of a single line to ground fault on the system. However, in the case of a
solid single phase to ground fault, the line to ground voltage on the unfaulted phases will
increase by a factor of 1.73. In the case of an intermittent or arcing ground fault, line to
ground voltage could increase to several times normal values. In either case, the
insulation systems of connected equipment such as motors will be subjected to
increased stresses and possible failure.
The SONGS 480 VAC electrical distribution system is equipped with ground detection
relays that provide annunciation in the control room in case a ground occurs.
The procedures for responding to 480 VAC system ground alarms include Alarm
Response Instruction SO-15-63.B, and Operating Instruction SO23-6-33, Ground
Isolation. Neither of these procedures identifies that a ground alarm indicates the
presence of an overvoltage condition on the affected 480 VAC system, or alerts
operators to the increased possibility of secondary faults. The compensatory actions
stated in Procedure SO-15-63.B simply require monitoring ground volts once an hour.
This conveys a lack of urgency in removing grounds. In addition, the procedure does
not identify prudent measures that could be taken to prevent secondary grounds.
Because of train separation, a ground will only affect one 480 VAC train at a time. This
enables compensatory actions such as starting redundant loads on the unaffected train
and securing loads on the affected train to isolate them from the overvoltage. Although,
Procedure SO23-6-33 provides for evaluating starting of redundant equipment, this step
is provided to avoid loss of functions when de-energizing power supplies, rather than a
proactive measure to transfer functions to the unaffected train to protect equipment.
Entries in AR 080300460 for response to a ground alarm on Bus 3B04 in March, 2008,
showed that it took 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> to identify and remove the ground. This indicated a routine,
rather than a prompt response and may have exposed connected equipment to
overvoltage for an unnecessarily long period of time.
Analysis: The failure to provide an adequate alarm response procedure was a
performance deficiency as demonstrated by the event in March of 2008. This finding
was more than minor because the procedure deficiency affected the mitigating system
cornerstone objective (procedure quality attribute) of ensuring availability, reliability, and
capability of systems needed to respond to initiating events to prevent undesired
consequences. Using the IMC 0609, "Significance Determination Process," Phase 1
Worksheets, the finding is determined to have very low safety significance (Green)
because it was not a design or qualification deficiency, did not result in a loss of safety
function, and did not screen as potentially risk significant due to external events. This
finding was reviewed for cross-cutting aspects and none were identified.
Enforcement: SONGS, Units 2 and 3, Technical Specifications 5.5.1.a, requires, in part,
procedures recommended by Regulatory Guide 1.33, Appendix A. Section 5 of
Appendix A recommends procedures for abnormal, off-normal, or alarm conditions and
states that these procedures identify the meaning of the annunciator, the immediate
operator actions and the long-range actions. Contrary to the above, July 18, 2008,
Alarm Response Instruction SO3-15-63.B and Operating Instruction SO26-6-33 were
inadequate, in that they failed to adequately identify the meaning of the alarm, and
- 22 -
Enclosure
provide appropriate immediate and long-range operator actions. Specifically, the
instructions did not identify the potential for the presence of harmful overvoltage or
provide appropriate actions for responding to the condition. Since this finding was of
very low safety significance and has been entered into the licensees corrective action
program as Notifications NN 200057494 (addresses ground fault trending) and NN
200057495 (addresses procedure change), this violation is being treated as a NCV,
consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000361,05000362/2008010-06, "Inadequate Procedures for 480 VAC Bus Grounds."
.3
Results of Reviews for Operating Experience:
.3.1
Inspection of Information Notice (IN) 2006-15, Vibration-Induced Degradation and
Failure of Safety-Related Valves.
a.
Inspection Scope
The team reviewed this IN, which documented vibration-induced degradation of safety-
related valves manufactured by Fisher Controls, Henry Pratt Company, and Flowserve
Corporation. The team reviewed the licensee response to this IN by reviewing the
procedures associated with monitoring and control of vibration-induced damage to
various types of valves, including those manufactured by these companies. The team
requested and reviewed a list of valves installed at San Onofre Units 2 and 3 - of which
there were approximately 150 - and searched the AR database using a sample of
specific valve numbers (approximately 15 percent sample) to identify whether any of
these valves had experienced vibration-induced degradation.
b.
Findings
No findings of significance were identified.
.3.2
Inspection of IN 2006-21, Operating Experience Regarding Entrainment of Air into
Emergency Core Cooling and Containment Spray Systems.
a.
Inspection Scope
The team reviewed the licensees evaluation of IN 2006-21, which documented
operating experience regarding possible air entrainment into emergency core cooling
and containment spray systems under post accident conditions. The licensee performed
this evaluation under AR 061001406-01, which included a reference to calculation M-
0012-036, Postulated Transient Recirculation Flow from Refueling Water Storage Tanks,
Revision 2. The team reviewed calculation M-0012-036, which evaluated the potential of
air entrainment from the refueling water storage tank and the potential of air reaching the
Emergency Core Cooling System pumps. In addition, the team interviewed engineering
and operations personnel regarding the post accident operation of these systems from
both the refueling water storage tank and containment sump to verify that system
operation was consistent with the analyzed condition.
b.
Findings
No findings of significance were identified.
- 23 -
Enclosure
.3.3
Inspection of IN 1997-40, Potential Nitrogen Accumulation Resulting from Back-Leakage
from Safety Injection Tanks.
a.
Inspection Scope
The team reviewed the licensee's response to IN 1997-40. The licensee conducted an
evaluation on December 31, 1997, which did not reveal any occurrences of nitrogen
accumulation due to back-leakage from the Safety Injection Tanks into the Emergency
Core Cooling System. The team reviewed operating and maintenance procedures for
the Emergency Core Cooling System to verify that proper guidance was provided to
operators to ensure that gasses, including nitrogen, do not accumulate in the system.
The team reviewed the isometric drawing of the Emergency Core Cooling System to
verify that the vent paths encompassed the high points of the system. The licensee is
currently performing actions in response to Generic Letter 08-02 for which the required
action date has not yet been reached.
b.
Findings
No findings of significance were identified.
.3.4
Inspection of IN 2006-26, Failure Of Magnesium Rotors In MOV Actuators.
a.
Inspection Scope
In response to IN 2006-26, the licensee initiated AR 061101243 dated November 22,
2006, to address the identified concerns. The team reviewed the AR and all documents
referenced therein, to assure that the licensees actions were appropriate and fully
addressed those concerns. The team also verified, by review of daily logs, that the
environmental conditions (i.e., temperature and humidity) used by the licensee to
support their conclusions with respect to the potentially degrading conditions identified in
the IN, were appropriate. Further, the team was able to verify through review of the
licensees Generic Letter 89-10 program that the population of valves identified by the
licensee as being subject to the conditions of the Information Notice was complete.
b.
Findings
No findings of significance were identified.
.4
Results of Reviews for Operator Actions:
The team selected risk-significant components and operator actions for review using
information contained in the licensees probabilistic risk assessment. This included
components and operator actions that had a risk achievement worth factor greater than
two or Birnbaum value greater than 1E-6.
a.
Inspection Scope
For the review of operator actions, the team observed operators during simulator
scenarios associated with the selected components as well as observing simulated
actions in the plant.
- 24 -
Enclosure
Inspection procedure 71111.21 requires a review of three to five relatively high-risk
operator actions. The sample selection for this inspection was five operator actions.
The selected operator actions were:
Reactor Coolant Pump Seal Heat Exchanger tube leak into the Component
Coolant Water system with a failure of the main turbine to trip during subsequent
Reactivity management during blended flow evolutions such as delithiation
Loss of Offsite Power with Emergency Diesel Generator malfunctions
Steam Generator Tube Rupture with failed Main Steam Isolation Valve and Main
Feed Isolation valve on the affected steam generator
Loss of a class 1E 125 VDC bus D2 with a failed open steam generator safety
relief valve
b.
Findings
No findings of significance were identified.
4
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (71152)
.1
Routine Reviews of Identification and Resolution of Problems
a.
Inspection Scope
The team chose one issue for a more in-depth review to verify that the licensee
personnel had taken corrective actions commensurate with the significance of the issue.
The team noted that several reactivity management issues were identified within the
licensees Corrective Action Process. The team reviewed the corrective actions
associated with a sample of these conditions focusing on Action Requests which
addressed blended flow evolutions and the Chemical and Volume Control System.
When evaluating the effectiveness of the licensees corrective actions, the following
attributes were considered:
Timeliness of corrective actions and/or repairs to components
Repetitive reactivity excursions from blended flow evolutions, indicating possible
ineffective corrective actions
Functionality/operability of components which affect reactivity
Documents reviewed are listed in the attachment.
b.
Findings
Introduction: The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion
XVI, "Corrective Action," for the failure of operations management, operations training,
- 25 -
Enclosure
and engineering to ensure that conditions adverse to quality are promptly identified and
corrected. Specifically, multiple reactivity excursions occurred in the plant over the past
two years, where corrective actions have been ineffective at addressing blended flow
evolutions.
Description: The team noted that several reactivity management issues were identified
within the licensees corrective action process. The team reviewed the corrective
actions associated with a sample of these conditions focusing on ARs which addressed
blended flow evolutions and the chemical and volume control system. The team noted
the occurrence of a significant number of reactivity events during the past two and one-
half years (January 2006 to June 2008). Associated with the specified time interval, the
team determined that an average occurrence of more than one reactivity condition per
week was identified, of which at least one every two months involved blended flow
operations. Based on this observation, the team performed a review of ARs associated
with blended flow evolutions. The team noted the following:
Since the licensees inception of the reactivity management program, two
assessments have been performed at approximately two-year intervals. The
initial assessment, documented in AR 050600107, was accomplished in June
2005 by a combination of in-house and peer evaluators. The team noted that
this effort was conducted only six months after the programs origination and
focused on the programs structure and compliance effectiveness as it pertains to
industry guidance. The assessment documented that training was not provided
for all work groups directly impacted.
The second assessment, documented in AR 070900159, was a self assessment
performed in 2007 to evaluate the implementation and effectiveness of the
reactivity management program as it pertains to the same industry guidance as
before. The self-assessment once again identified that not all stakeholders
received initial or continuing training on their associated program responsibilities.
Although some equipment reliability concerns were discussed in the self-
assessment, the sampled ARs reviewed by the team supported the equipment
response as indicated in the field support analysis sections. Recommended
remedial actions included procedure changes and training on blended makeup
operations. However, the team noted that several Action Requests have
documented blended flow issues since these actions were implemented. The
team acknowledged that some training was provided to operating crews as part
of their normal requalification program during 2008. The team concluded that
these remedial measures have not been effective.
The team noted three conditions which addressed procedures associated with
reactivity control aspects. AR 071000317 identified the need to incorporate a
procedure change to SO23-3-2.2 "Makeup Operations," Revision 21, to ensure
proper blend settings on the borate and dilute flow controllers before returning
the makeup mode selector switch to automatic, or leave the switch in manual if
blend settings are not verified. Although AR 071000317 documented the
specified procedure changes as been completed in Revision 22, the team
observed that the latest procedure change (Revision 23) had included flow
setting steps after having placed the mode selector switch in automatic.
- 26 -
Enclosure
In addition, Procedure SO23-3-2.2, step 6.6.19, allows adjusting flow controller
settings for blended flow evolutions; however, it also required any such changes
be annotated in the Nuclear Control Operator (NCO) log. Procedure SO123-0-
A1, "Conduct of Operations," Revision 14, section 6.4.9 identifies the NCO log as
an official site document providing an overall plant record of significant operating
events." As documented in AR080600116, multiple changes in flow settings
occurred during blended flow operations on June 1, 2008. However, the team
verified that no entries were made in the NCO logs as required by procedure step
6.6.19.
The team observed that Procedure SO23-3-2.2 was revised in January of 2008
(Revision 22 reference AR 071001452-3) to raise the minimum flow setting
criteria for boric acid makeup flow to greater than 2 gallons per minute, since low
flow controller settings contribute to control system inaccuracy. The team noted
that section L&S 3.1 of the procedure was changed to identify the inaccuracy for
low boric acid flow conditions and prevent such occurrences for blended flows of
raising pure water vice lowering boric acid flow. Contrary to the procedure, on
June 1, 2008, boric acid flow was dropped to 1.6 gallons per minute and
remained less than 2.9 gallons per minute over the makeup evolution or to the
power response.
Licensee staff indicated that ACE's are utilized for level three or higher
classifications of reactivity events. The team noted that procedure SO123-XV-
91, "Reactivity Management Implementation," Revision 2, section 6.11.2 states
for level three or higher a possible cause evaluation be accomplished on a case
by case basis. Regarding blended flow evolutions, the team found only two
occasions which were categorized at level three. Following one instance,
documented in AR 071100792, a Direct Cause Evaluation (DCE) was performed.
The team found that in the other level three event, as annotated in AR
070700065, no Cause Evaluation was performed and thus no corrective actions
identified. As a result, most AR's annotating blended flow difficulties incurred by
licensed operators were classified as level four or five with narrowly focused
actions, if any, such that each event was treated as an isolated case.
The team observed that even with a DCE as an assigned task, Procedure
SO123-XV-50 "Corrective Action Process," Revision 7, section 6.9.1.2,
considered evaluations of site operating experience, industry operating
experience, and determining corrective actions to prevent recurrence as optional.
Additionally, for an ACE, the same options are stated only if a Common Cause
Evaluation is performed in lieu of an ACE; thus, the evaluation of site operating
experience is not applicable. The team concluded that multiple opportunities
existed during the past two and one-half years to identify corrective actions to
preclude blended flow reactivity events.
During the teams last week onsite a meeting was held to discuss these observations
regarding reactivity oversight with the licensee as well as the Branch Chief for
Engineering Branch 1 from the Division of Reactor Safety of the Region IV office. During
this meeting the licensees Operations Director communicated to the Branch Chief and
team that additional corrective actions, beyond those already implemented or planned as
part of the reactivity oversight group recommendations were not necessary. The team
- 27 -
Enclosure
concluded that the corrective actions taken and planned were ineffective as indicated by
the continuous trend of reactivity management events.
Analysis: The performance deficiency associated with this finding was the failure of
operations management (ineffective management by the reactivity oversight group),
operations training (ineffective training on these events), and engineering personnel
(mechanical issues with the batch controller and inability to complete installation of
equipment to preclude some of these issues) to implement adequate corrective actions
to prevent these reactivity excursions. The finding is more than minor because it is
associated with the initiating events cornerstone (human performance attribute) and
affects the associated cornerstone objective to limit the likelihood of those events that
upset plant stability and challenge the critical safety functions during shutdown as well
as power operations. If left uncorrected, the conditions would continue to contribute to
additional operator errors or significantly impact the operators ability to perform blended
flow evolutions. Using the IMC 0609, "Significance Determination Process," Phase 1
Screening Worksheets, the finding is determined to have very low safety significance
(Green) because it did not contribute to both the likelihood of a reactor trip and the
likelihood that mitigating equipment or functions will not be available. This finding has a
crosscutting aspect in the area of problem identification and resolution associated with
the corrective action program because the licensee did not thoroughly evaluate
problems such that resolutions address causes and extent of condition [P.1.(c)].
Enforcement: The regulations in Title 10 of CFR Part 50, Appendix B, Criterion XVI,
"Corrective Action," state in part that measures shall be established to assure that
conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations,
defective material, and equipment, and nonconformances are promptly identified and
corrected. Contrary to the above, between January 1, 2006, and June 30, 2008, multiple
deficiencies involving reactivity management issues identified by the licensee and
entered into the corrective action program were not corrected and continue to occur.
However, because the finding is of very low safety significance and has been entered
into the licensees new corrective action program as Notifications NN 200062659
(addresses procedure change) and NN 200006366 (addresses common cause
evaluation), this violation is being treated as a NCV, consistent with Section VI.A of the
Enforcement Policy: NCV 05000361,05000362/2008010-07, " Inadequate Corrective
Actions for Reactivity Events."
4OA6 Meetings, Including Exit
On July 17, 2008, the team leader presented the preliminary inspection results to Mr. Ed
Scherer, Director, Nuclear Regulatory Affairs, and other members of the licensees staff.
On September 11, 2008, the Engineering Branch 1 Chief conducted a telephonic exit
meeting with Mr. Ridenoure and other members of the licensee's staff. The licensee
acknowledged the findings during each meeting. While some proprietary information
was reviewed during this inspection, no proprietary information was included in this
report.
4OA7 Licensee Identified Violations
No findings of significance were identified.
- 28 -
Enclosure
Attachment: 1 - Supplemental Information
- 1 -
Attachment
ATTACHMENT 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
R. Ridenoure, CNO, Sr. VP, and Site Manager
E. Scherer, Director, NRA
J. Reilly, VP, Engineering Services
A. Hochevar, Station Manager
K. Johnson, Manager, Design Engineering
T. Yackle, Director, Operations
M. Short, Director, Nuclear Oversight
NRC personnel
D. Loveless, Senior Reactor Analyst, Region IV
G. Warnick, Senior Resident Inspector, SONGS
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000361;
Inadequate Design Control for Design Basis of 05000362/2008010-01
CCW/CWC Heat Exchangers (1R21.2.11)
05000361;
Inadequate Design Control for 125VDC Control 05000362/2008010-02
Circuits (1R21.2.14.1)
05000361;
Omission of Station Blackout Profile During Battery 05000362/2008010-03
Service Tests (1R21.2.14.2)05000361/2008010-04
Inadequate Corrective Actions for Battery
Performance Test Issues (1R21.2.14.3)05000361/2008010-05
Failure to Follow Procedures During
the Battery Performance Tests (1R21.2.14.4)
05000361;
Inadequate Procedures for 480 VAC System 05000362/2008010-06
Grounds (1R21.2.16)
05000361;
Inadequate Corrective Actions for Reactivity 05000362/2008010-07
Events (40A2)
- 2 -
Attachment
LIST OF DOCUMENTS REVIEWED
In addition to the documents called out in the inspection report, the following documents were
selected and reviewed by the team to accomplish the objectives and scope of the
inspection and to support any findings:
Action Requests
930500081
050700549
061001697
070801013
970900408
050701607
061001698
070801046
981101445
050801233
061001789
070801063
990302025
051000729
061100354
070801396
990801013
051001114
061100634
070900159
000100562
051100747
061100714
070900365
000200411
051200478
061101106
070900384
001100920
060100422
061101243
070900547
010901053
060100999
061101250
070900711
011001711
060101152
061101692
071000587
020601551
060101159
070100063
071000901
020701508
060120817
070100371
071001091
020800811
060200377
070100499
071001452
030101629
060200599
070101104
071100792
030102296
060201081
070101383
071200996
030202237
060201528
070101434
071201105
030202237
060300008
070200254
071201294
030300125
060300900
070200495
071201393
030300661
060301020
070300161
080100576
030301185
060500485
070300185
080100597
030600531
060500578
070301055
080100702
031000026
060500834
070400046
080101004
031000264
060600109
070400088
080101072
031100614
060600239
070400389
080101122
031100924
060600564
070400447
080101185
031200173
060600921
070400701
080200288
040300714
060600979
070400993
080200592
040401569
060601355
070500432
080201055
040401649
060700747
070500439
080201394
040500921
060700806
070500593
080201438
040700701
060700878
070500820
080300460
040701362
060701285
070501022
080300491
040800664
060800056
070501169
080300619
040801026
060800601
070501189
080300673
040801061
060800603
070501385
080301117
040801171
060800843
070600347
080400613
040801333
060800980
070600413
080400668
040801372
060801229
070600607
080400813
040900319
060900352
070600862
080500060
- 3 -
Attachment
041101031
060900770
070601000
080500286
041101251
060900839
070700065
080500859
050100457
060900881
070700216
080501171
050300741
060901108
070700459
080600073
050301561
060901110
070700489
080600076
050600107
061000406
070700909
080600116
050601315
061000820
070701028
080600438
050601324
061001150
070701212
GR-0035
050700073
061001371
070800069
050700141
061001379
070800940
050700169
061001406
070800993
Notifications Written from the Inspection
200006366
200054737
200057494
200059581
200047962
200054738
200057495
200060319
200048442
200054739
200057527
200061041
200048884
200056981
200058348
200062659
200051692
200056986
200059004
200054736
200057484
200059017
Calculations
M-0026-011, "CCW Flow/Pressure Distribution Analysis," CCN-10
Misc-PEC-119, "3410 MWT Plant Safety Injection Tank Sizing," 9/7/1972
J-BHA-060, "Scaling Calculation for Safety Injection Tank Level Transmitters," Rev 1
J-BHA-002, "Instrument Uncertainties for SIT Narrow Range Pressure Loops," Rev. 0
M-0056-034, "Dynamic Simulation of MFIV Closure," Rev. 0
M-AOV-SP-2HV8419, "Setpoint Calculation for AOV 2HV8419," Rev. 0
M 41.35, "Sizing of Nitrogen Storage Bottle for Valve Actuation," 3/17/1981
J-BHA-011, "Containment Emergency Sump (Wide Range) Level Loop Uncertainties," Rev. 0
J-BHA-012, "Containment Emergency Sump High Level Setpoint," Rev. 1
M-0027-017, "Backup Nitrogen Supply for the CCW Surge Tanks," Rev. 0
M-0027-023, "CCW/SWC Heat Exchanger Operability," Rev. 0
M-0027-029, "CCW/SWC Heat Exchanger Performance Tests," Rev. 0
M-0027-035, "CCW System Letdown Heat Exchanger Bypass Sizing Calculation," Rev. 0
M-1204-002-04A, "Valve Seat Leakage to a Refueling Water Storage Tank," Rev. 0
M-8910-SP-3HV4716, GL 89-10 Setpoint Calculation: 3HV4716, Rev. 3
M-8910-SP-2HV9302, GL 89-10 Setpoint Calculation: 2HV9302, Rev. 2
M-8910-SP-3HV9302, GL 89-10 Setpoint Calculation: 3HV9302, Rev. 2
M-8910-SP-2HV9303, GL 89-10 Setpoint Calculation: 2HV9303, Rev. 2
M-8910-SP-3HV9303, GL 89-10 Setpoint Calculation: 3HV9303, Rev. 3
M-8910-SP-2HV9304, GL 89-10 Setpoint Calculation: 2HV9304, Rev. 2
M-8910-SP-3HV9304, GL 89-10 Setpoint Calculation: 3HV9304, Rev. 2
- 4 -
Attachment
M-8910-SP-2HV9305, GL 89-10 Setpoint Calculation: 2HV9305, Rev. 3
M-8910-SP-3HV9305, "GL 89-10 Setpoint Calculation: 3HV9305," Rev. 2
M-0012-01D, "NPSH of ESF Pumps," Rev. 2
N-0240-006, "RWST Tech Spec Requirement," Rev. 0
N-4060-030, "Containment Flooding Level," Rev. 1
N-6060-003, "LOCA ESF Leakage, CR & Offsite Doses - AST," Rev. 0
N-6060-004, "LOCA RWST Releases, CR & Offsite Doses - AST," Rev. 0
SO23-452-F, "Salt Water Cooling System Pump Sizing," Rev. 1
J-GJA-055, "Emergency Chiller Low Chilled Water Temperature Setpoint," Rev. 0
J-GJA-075, ICCN C-1, "ECW Oil Heater Temperature Control Switch Setpoint," Rev. 2
M-0073-130, ICCN C-2, "Evaluation of ECW System Surveillance Test Result"
M-0073-88, "Evaluation for Chiller Performance," Rev. 0
M-0073-83, "Plant Emergency Chilled Water System Equipment Sizing Calcs."
M-0073-87, "Pressure Drop Emergency Chilled Water System"
EC-119, "Emergency Chiller Freon Level -Units 2/3," Rev. 0
M-0075-052, "Units 2&3 Trains A and B Emergency Room Cooler Capacity Verification"
J-PEC-24/S-PEC-10, "Sizing of HP & LP Safety Injection Pumps"
M-0073-034, "Aux. Bldg. Control Area 9'-0" El. Chiller Room Emer. Heat Load Calc."
M-0041-096, "Maximum Differential Pressure Across MSIV," Rev. 0
N-4080-027, with CCN N-6, "Containment P-T Analysis For Design Basis MSLB," Revision 1
E4C-017, "125V Battery & DC System Sizing," CCN-94, Rev. 19
E4C-017.1, "Class 1E 125VDC System Data/Loading," Rev. 3
E4C-082, "System Dynamic Voltages During DBA," Rev. 3
E4C-084, "Unit 2 MCC Control Circuit Voltage Analysis," ECN No. A44808, Rev. 0
E4C-085, "Unit 3 MCC Control Circuit Voltage Analysis," ECN No. A44809, Rev. 0
E4C-088, "Emergency Diesel Generator Loading," Rev. 4
E4C-090, "Auxiliary System Voltage Regulation," Rev. 5
E4C-102, "GL 89-10 MOV Voltages During Design Basis Accident," Rev. 3
E4C-109, "CLASS 1E 125V DC System Protection Calculation," Rev. 4
E4C-123, "Voltage Requirements for 120VAC Vital Buses," Rev. 1
J-BHA-082, "Indicator TLU, Alarm Set-points, and Strapping Data for Safety Injection Tank
Level Loops," Rev. 0
J-BHB-021, "RWST 2(3) T005 & T006 Level Loop Uncertainties and Minimum Volume Required
During Modes 5&6," Rev. 0
J-EPA-002, "TLU for Saltwater Flow to CCW Heat Exchangers 2(3)E001A & 2(3)E002B, Rev. 1
M-0012-036, Postulated Transient Recirculation Flow from Refueling Water Storage Tanks,"
Rev. 2
M-42750, "Safety Injection System Recirculation Realignment Function Failure Modes and
Effects Analysis," Rev. 0
M-8910-1301-OB-001, GL 89-10 "Operational Basis Calculation for the AFW Pump Turbine
Stop Valve," Rev. 0
N-4060-015, "Sump Level vs. Volume: Normal, Emergency, and Wide Range Containment
Area," Rev. 0
M-0073-041, ICCN 1, "Aux. Bldg Ctrl Area El. 30'-0", Heat Load & Equip Sizing Normal &
Emergency"
M-0012-033, ICCN-1, "HPSI Pump Tech. Spec. Minimum Performance Requirements," Rev. 2
E4C-130, "TLU Calc for Undervoltage Relay Circuit at Class 1E 4KV Switchgear," ECN, Rev. 1
- 5 -
Attachment
No. A4780
E4C-131, "125V DC Control Circuit Analysis for Class 1E 4kV and 480V Circuit Breaker
Operation," Rev. 1
Completed Inservice Tests
SO23-3-3.60.4, Salt Water Cooling Pump, April 9, 2008, April 10, 2008, April 29, 2008, May 13,
2008, May 19, 2008, June 5, 2008
Design Basis Documents
DBD-SO23-740, "Safety Injection, Containment Spray, and Shutdown Cooling Systems," Rev. 9
DBDSO23-400, Component Cooling Water System, Rev. 11
DBD-SO23-410, Saltwater Cooling System, Rev. 8
DBD-SO23-740, Safety Injection, Containment Spray, and Shutdown Cooling Systems, Rev. 9
DBD-SO23-780, Auxiliary Feedwater System, Rev. 8
DBD-SO23-800, "Auxiliary Building Emergency Chill Water System," Rev. 9
DBD-SO23-740, "Safety Injection, Containment Spray, and Shutdown Cooling Systems," Rev. 9
DBD-SO23-360, "Reactor Coolant System," Rev. 10
DBD-SO23-TR-AA, "Accident Analysis Topical Report"
DBD-SO23-120, "6.9kV 4.16kV and 480V Electrical System," Rev. 6
DBD-SO23-140, "Class 1E 125V DC System," Rev. 6
DBD-SO23-750, "Emergency Diesel Generator," Rev. 3
DBD-SO23-TR-SF, "Single Failure Topical Report," Rev. 5
DBD-SO23-400, "Component Cooling Water System Design Basis Document," Rev. 11
DBD-SO23-365, "Steam Generators and Secondary Side Design Basis Document, Rev. 9
Drawings
40111A, "P&I Diagram: Reactor Coolant System," Rev. 40
40111B, "P&I Diagram: Reactor Coolant System," Rev. 31
40111D, "P&I Diagram: Reactor Coolant System," Rev. 10
40112A, "P&I Diagram Safety Injection System," Rev. 33
40112C, "P&I Diagram Safety Injection System," Rev. 19
40112D, "P&I Diagram Safety Injection System," Rev. 23
40113A, "P&I Diagram Safety Injection System," Rev. 17
40113B, "P&I Diagram Safety Injection System," Rev. 16
40123A, "P&I Diagram: Rector Coolant Chemical and Volume Control System," Rev. 24
40124B, "P&I Diagram: Reactor Coolant Chemical and Volume Control System," Rev. 33
40040, "Tube Plug Map for Component Cooling Water Heat Exchanger S21203WE001," Rev. 5
40041, "Tube Plug Map for Component Cooling Water Heat Exchanger S21203WE002," Rev 10
40042, "Tube Plug Map for Component Cooling Water Heat Exchanger S31203WE001," Rev. 5
40043, "Tube Plug Map for Component Cooling Water Heat Exchanger S31203WE002," Rev. 4
40112A, "P&I Diagram - Safety Injection System," Rev. 33
40112B, "P&I Diagram - Safety Injection System," Rev. 35
40112C, "P&I Diagram - Safety Injection System," Rev. 19
- 6 -
Attachment
40112D, "P&I Diagram - Safety Injection System," Rev. 23
40113A, "P&I Diagram - Safety Injection System," Rev. 17
40113B, "P&I Diagram - Safety Injection System," Rev. 16
40126A, "P&I Diagram - Component Cooling Water System (Salt Water Pumps)," Rev. 28
40126B, "P&I Diagram - Component Cooling Water System (Salt Water Pumps)," Rev. 28
40127A, "P&I Diagram - Component Cooling Water System," Rev. 29
40127B, "P&I Diagram - Component Cooling Water System," Rev. 38
40127C, "P&I Diagram - Component Cooling Water System," Rev. 44
40127D, "P&I Diagram - Component Cooling Water System," Rev. 15
40127E, "P&I Diagram - Component Cooling Water System," Rev. 18
40127F, "P&I Diagram - Component Cooling Water System," Rev. 34
40127G, "P&I Diagram - Component Cooling Water System," Rev. 15
40180A, "Auxiliary Building Emergency Chilled Water System Loop B," Rev.31
40180B, "Auxiliary Building Emergency Chilled Water System," Rev. 9
40180C, "Auxiliary Building Emergency Chilled Water System Loop B," Rev. 13
40180D, "Aux Bldg Emergency Chilled Water System - Water Chiller E335," Rev.15
40112A, "P&I Diagram - Safety Injection System," Rev. 33
40112B, "P&I Diagram - Safety Injection System," Rev. 35
40112C, "P&I Diagram - Safety Injection System," Rev.18
40112D, "P&I Diagram - Safety Injection System," Rev. 23
30644, "Elementary Diagram Reactor High Pressure Safety Injection Pump P019," Rev. 15
32116, "One Line Diagram 480V Loadcenter," Rev. 13
32118, "One Line Diagram 480V Loadcenter," Rev. 19
32122, "One Line Diagram 480V Loadcenter," Rev. 13
48778, "Pressurizer Heater Map," Rev. 3
32171, "One Line Diagram Pressurizer Heaters Distribution Panels," Rev. 12
40141C, "Main Steam System, Electro-Hydraulic Valve 2HV-8204"
40141G, "Main Steam System"
30142, "One Line Diagram 480V Motor Control Center 2BJ (ESF)," Rev. 29
30172, "One Line Diagram Class 1E 124V DC and 125VAC Power System," Rev. 16
30174, "One Line Diagram 125V DC Distribution Switchboard 2D2," Rev. 21
30263, "Elementary Diagram Electrical Aux - 40V Bus 2B04 & 3B04 Metering," Rev. 21
32017, "One Line Diagram 4160V Switchgear Bur 3A04 (ESF)," Rev. 16
32113, "One Line Diagram Diesel Generator Protection," Rev. 7
32164, "One Line Diagram 480V Motor Control Center 3BZ (ESF)," Rev. 35
32328, "Elementary Diagram Elect. Aux. 4.16kV Bus 3A04 DG 3G002 Breaker," Rev. 25
32329 Sht. 1, "Elementary Diagram Diesel Generator 3G002 Protection AC System," Rev. 12
32329 Sht. 2, "Elementary Diagram Diesel Generator 3G002 Protection AC System," Rev. 8
32330, "Elementary Diagram Diesel Generator 3G002 Protection DC System," Rev. 12
32342 Sht. 1, "Elementary Diagram Diesel Generator 3G002 Control DC System," Rev. 12
32342 Sht. 2, "Elementary Diagram Diesel Generator 3G002 Control DC System," Rev. 10
32342 Sht. 3, "Elementary Diagram Diesel Generator 3G002 Control DC System," Rev. 9
32342 Sht. 4, "Elementary Diagram Diesel Generator 3G002 Control DC System," Rev. 10
32342 Sht. 5, "Elementary Diagram Diesel Generator 3G002 Control DC System," Rev. 12
5105074, "One Line for Operation Position 1 thru 6," Rev. 7
5105075, "One Line for Operation Position 7 thru 14," Rev. 9
5105076, "One Line for Operation Position 15 thru 19," Rev. 7
- 7 -
Attachment
Maintenance Work Orders
MO 07040293000, "IST Loop Flow Inst. Cal. Pump 1st Req. Maint." 2/28/2008
MO 05110847000, "IST LIST Pressure Gauge Calibration," 9/20/2006
MO 06080564000, "IST Loop Flow Inst. Cal. Pump 1st Req. Maint.", 6/6/2007
MO 04121873000, "Loop Level Instrumentation Calibration," 1/26/2006
MO 04121881000, "Loop Level Instrumentation Calibration," 1/25/2006
MO 04121880000, "Loop Level Instrumentation Calibration," 1/25/2006
MO 04121894000, "Loop Pressure Instrumentation Calibration," 1/25/2006
MO 04121895000, "Loop Pressure Instrumentation Calibration," 1/20/2006
MO 04121896000, "Loop Pressure Instrumentation Calibration," 1/20/2006
MO 04121894000, "Pressure Instrumentation Calibration," 1/23/2006
MO 10016014001, "Replace Hydraulic Dump Valves"
CM 08060875000, "no title"
PM 08052498000, "no title"
PM 08051945000, "no title"
CM 08011406000, "no title"
CM 08010792000, "no title"
MO 01111509000, "no title"
MO 03022903000, "no title"
MO 05011306000, "no title"
MO 05062182000, "no title"
MO 05080446000, "no title"
MO 06031767000, "no title"
MO 06070902000, "no title"
MO 06070902000, "no title"
MO 06081267000, "no title"
MO 06091358000, "no title"
MO 06100894000, "no title"
MO 07020581000, "no title"
MO 07090743000, "no title"
MO 07101712000, "no title"
MO 07031974000 for Valve 2HV9336, "no title"
MO 07031975000 for Valve 3HV9336, "no title"
MO 07011556000 for Valve 2HV9337, "no title"
MO 07011557000 for Valve 3HV9337, "no title"
MO 07011558000 for Valve 2HV9339, "no title"
MO 07011559000 for Valve 3HV9339, "no title"
MO 06100088000, "Snoop for nitrogen leakage at all assessable valves, tubing and transmitter
connections and man-way on S21204MT010." 6/29/2007
MO 06031230000, "Snoop for nitrogen leakage at all assessable vales, tubing and transmitter
connections and man-way on S21204MT010." 3/16/2006
MO 03031674000, "Replace Section of Tubing and Fittings Between 2LT0341 and
S21204MR056," 3/20/2004
MO 03031113000, "Investigate for nitrogen leaks and repair and or generate AR for future
repair," 3/20/2003
- 8 -
Attachment
MO 03030234000, "Investigate for nitrogen leaks and repair and or generate AR for future
repair," 3/12/2003
MO 06121233000, "Filling and Venting Instrument Xmtrs / Sensing Lines and Position
Verification / Seal Wiring of Instrument Xmtrs Sensing Line Valves," 1/12/2008
MO 04101832000, "Support of Train A LOVs Procedure Load Sequence for 2A0406,
Component Cooling Pump 2P025," 10/17/2005
MO 06110701000, "Support of Train A LOVs Procedure Load Sequence for 2A0406,
Component Cooling Pump 2P025," 5/2/2007
Modifications/Design Change Packages
ECP 02040125-13, "no title", Rev. 0
ECP 030501114-25, "no title", Rev. 0
ECP 070600872-5, "no title", Rev. 0
ECP 031101156-10, "no title", Rev. 0
ECP 031101156-9, "no title", Rev. 0
ECP 071200620-7, "no title," Rev 0
ECP 030600531-4, "no title," Rev 0
ECP 010101512-6, "no title," Rev 0
ECP 061001379-78, "Add CCW By-Pass around Letdown Heat Exchanger in Unit 2," Rev. 0
ECP 061001379-84, "Install CCW By-Pass around the Unit 3 Letdown Heat Exchanger," Rev. 0
ECP 060201528-7, "Update Tube Coating Information," Rev. 0
ECP 030201099-2, "Removal of internals for solenoid valve 2/3HYJ888A," Rev. 0
ECP 020301046-5, "Update HPSI pump Inservice testing curves," Rev. 0
ECP 020401425, "Installation of oil sample ports," Rev. 0
ECP 060102006-5, "Replacement of Hydraulic Dump Valves Solenoid Valves," Rev. 0
050101273-4, Drill and Tap the Top Edge of the Unit 2 (MSIV) Limit Switch Tree for the
Installation of a Shouldered Eyebolt to Facilitate Rigging and Lifting of the Tree Assembly
During Maintenance Activities," Rev. 0
030202237-8, "MSIV Inservice Testing is Being Performed Under Conditions Which do not
Reflect the Conditions Under Which the Valves Would Be Required to Perform Their Safety
Function," dated 5/1/03
070600074-37, "U2 MSIV, MFIV and MFBV Hydraulic Actuator Skid Dissimilar Weld Removal,"
Rev. 0
DCP-3-6204, "Modification of the Saltwater Side of the CCW Heat Exchanger to Install a
Backflush System," Rev. 0
ECP 060700806-5, "Replace Existing Level Transmitter 3LT0301 with new Model per SEE
060049," Rev. 0
ECP 070500468-9, "Add Pipe Liquid Level Indicator to Line S2-1203-ML-100 between MR-095
and MR-039," Rev. 0
ECP 051100747-13, "Update DBD-SO23-400 to Clarify Performance Requirements for CCW
HXs," Rev. 0
ECP 0801046-5, "Addition of jumper to prevent spurious trips due to electrical noise in
Emergency Chiller ME335," Rev. 0
ECP 070801295-17, "Replacement of bearing oil return and compressor discharge temperature
sensors for SA1513ME335," Rev. 0
- 9 -
Attachment
ECP 020701592-2, "Addition of resistors to Emergency Chiller SA1513ME335 bearing
temperature and discharge temperature circuit," Rev. 0
ECP 030102083, "Addition of inspection windows on terminal boxes for obtaining thermographic
images," Rev. 0
Minor Modification Package 2/3-6794.05E, "Motor Temperature Module Removal from Chiller
Units," Rev. 0
Procedures
SO123-II-9.14, "Electronic Differential Pressure and Pressure Transmitter Calibration," Rev. 2
SO23-3-3.30.4, "Main Steam System Online Valve Test," Rev. 8
SO23-3-3.31.4, "Main Steam Valve Testing - Offline," Rev. 8
SO123-V-5.22.1, "Motor Operated Valve Program," Rev. 1
SO23-V-5.22.2, "Administration of the Air Operated Valve Program," Rev. 5
SO23-3-2.7.1, "Safety Injection Tank Operation," Rev. 16
SO23-3-3.25, "Once a Shift Surveillance (Modes 1 - 4)," Rev. 28
SO23-3-3.8, "Safety Injection Monthly Tests," Rev. 20
SO23-3-3.31.6, "Main and Aux. Feedwater Valve Testing - Offline or Long Interval," Rev 8
SO23-2-8, "Saltwater Cooling System Operation," Rev. 29
SO23-2-8.1, "Train A Emergency Discharge Line Operations," Rev. 7
SO23-2-17, "Venting CCW for Nitrogen Removal," Rev. 26
SO23-3-2.7.2, "Safety Injection System Removal/Return to Service Operation," Rev. 16
SO23-3-3.60.4, "SWC Pump Response Time and Vent Valve Inservice Testing," Rev. 10
SO23-5-1.1, "Heat Treating the Circulating Water System," Rev. 21
SO23-12-11, Emergency Operating Instructions, Rev. 6
SO23-15-64.A, Annunciator Response Instruction - "CCW Hx Train A Outlet Temp Hi," Rev. 13
SO23-V-2.8, "Saltwater Cooling Piping Internal Inspection," Rev. 2
SO23-XVII-8, "Outside Containment Leakage Reduction Program," Rev. 3
SO23-3-2.7, "Safety Injection System Operation," Rev.21
SO23-3-2.7.1, "Safety Injection Tank Operation," Rev.14
SO23-3-3.60.1, "Surveillance Operating Instruction," Rev.8
SO23-933-68, "Ingersoll - Rand HPSI Pump Manual," Rev.6
SO23-3-2.7.2, "Safety Injection System Removal / Return to Service Operation," Rev.13
SO23-3-2.6,"Shutdown Cooling System Operation," Rev.22
SO23-1-3.1, "Emergency Chilled Water System Operation," Rev.21
SO23-1-8.116, "HVAC - Carrier Chiller Inspections and Testing," Rev.5
SO23-3-3.20, "Control Room Emergency Air Cleanup System Test - Train B," Rev.20
SO23-3-2.10, "Main Steam Isolation Valve Operation," Rev.17
SO23-V-3.5, "Inservice Testing of Valves Program," Rev.29
SO23-3-3.31.4, "Main Steam Valve Testing - Offline," Rev.8
SO123-RX-1, "Reactivity Management Program," Rev.03
SO123-XV-50, "Corrective Action Process," Rev.07
SO123-XV-91, "Reactivity Management Implementation," Rev.02
SO123-O-A1, "Conduct of Operations," Rev.14
SO23-5-1, "Power Operations," Rev.27
SO23-6-17, "Transferring Vital Bus from Inverter to Alternate Source," Rev.13
SO23-6-33, "Ground Isolation," Rev.04
- 10 -
Attachment
SO23-9-8, "Main Feedwater Isolation and Block Valves," Rev.16
SO23-12-1, "Standard Post Trip Actions," Rev.20
SO23-12-2, "Reactor Trip Recovery," Rev.17
SO23-12-3, "Loss of Coolant Accident," Rev.19
SO23-12-4, "Steam Generator Tube Rupture," Rev.20
SO23-12-5, "Excess Steam Demand Event," Rev.20
SO23-12-7, "Loss of Forced Circulation / Loss of Offsite Power," Rev.19
SO23-12-8, "Station Blackout," Rev.19
SO23-12-9, "Functional Recovery," Rev.24
SO23-12-11, "EOI Supporting Attachment 14 - RAS Operation," Rev.14
SO23-13-4, "Operation During Major System Disturbances," Rev.09
SO23-13-6, "Reactor Coolant Pump Seal Failure," Rev.4
SO23-13-7, "Loss of Component Cooling Water / Saltwater Cooling," Rev. 8
SO23-13-14, "Reactor Coolant Leak," Rev. 10
SO23-13-22, "Loss of Control Room Annunciators," Rev. 2
SO23-13-26, "Loss of Power to an AC Bus," Rev. 3
SO123-I-9.26, "Miscellaneous Low Voltage Bus Panel Inspection, Cleaning and Testing," Rev. 2
SO123-I-9.12, "Motor Control Center Cleaning, Inspection and Megger Testing," Rev. 9
SO123-I-9.13, "480 VAC Linestarter Inspection, Coil and Power Contact Replacement," Rev. 9
SO123-I-4.59.4, "4kV/6.9kV Power Cable Termination & Repair Guide," Rev. 0
SO123-I-4.59.6, "600V Power Cable Termination & Repair Guide," Rev. 0
SO23-V-2.14, "Thermal Inspection of Plant Components," Rev. 8
SO123-1-2.5, "Battery Service Test and Rapid Recharge," Rev. 10
SO123-1-2.6, "Battery Performance Test and Rapid Recharge," Rev. 8
SO23-3-3.23, "Diesel Generator Monthly and Semi-Annual Testing," Rev. 33
SO23-3-3.23.1, "Diesel Generator Refueling Interval Tests," Rev. 27
SO23-6-33, "Ground Isolation," Rev. 4
SO-15-63.A, "63A53 2D2 Charger Trouble," Rev. 9
SO3-15-63.B, "63B36 3B04 480V Ground," Rev. 10
SO23-955-20, "20 KVA Uninterruptible Power Supply SN 9609," Rev. 4
SO23-302-4-2-357, "Model 4 MCC Installation & Maintenance Instruction Manual," Rev. 3
SO23-302-4-1-619, "Instruction Manual for Unitrol Motor Control Centers," Rev. 0
SO23-302-4-1-93-2, "Instruction Manual for Unitrol Motor Control Centers," Rev. 2
SO123-II-9.10, "Dietrich Model 1151DP and Rosemount Differential/Absolute/Gage Pressure
Transmitter Models 1151, 1152, 1153, 1154, and 3051 Calibration," Rev. 6
SO123-XXIV-10.1, "Preparation, Review, Approval, Issuance, Implementation, and Closure of
Engineering Change Packages (ECPs) and Engineering Change Notices (ECN)," Rev. 17
SO23-410-7-164-2, "Operating Instructions for Carrier Centrifugal Refrigeration Machines Using
Refrigerant Number 12"
Completed Surveillance Packages
SO123-1-2.5, "Battery Service Test and Rapid Recharge," performed 03/06/04
SO123-1-2.5, "Battery Service Test and Rapid Recharge," performed 01/02/08
SO123-1-2.6, "Battery Performance Test and Rapid Recharge," performed 02/16/08
SO123-1-2.6, "Battery Performance Test and Rapid Recharge," performed 01/25/07
SO123-1-2.6, "Battery Performance Test and Rapid Recharge," performed 02/18/06
- 11 -
Attachment
SO123-1-2.6, "Battery Performance Test and Rapid Recharge," performed 06/08/02
SO23-3-3.12, "Integrated ESF System Refueling," performed 10/12/06
SO23-3-3.23, "Diesel Generator G002 Semi-Annual Surveillance," performed 3/17/08
SO23-3-3.23, "Diesel Generator Monthly Surveillance," performed 04/14/08
SO23-3-3.23.1, "Diesel Generator G002 Refueling Interval Tests," performed 10/31/06
SO23-3-3.23.1, "Diesel Generator G002 Refueling Interval Tests," performed 9/3/06
Operating Experience
OE21243, "SONGS - Unit 2 Shutdown to Repair Leaking Solenoid Dump Valve on a Feedwater
Isolation Valve," 8/18/2005
OE25725, "Multiple Feedwater and Main Steam Isolation Valves Disabled by Solenoid Control
Valve Problems (SONGS), 10/29/2007
PVNGS Failure Number 920, "ADV-184 Failed the N2 Drop Test per 73ST-9SG05," 5/8/2006
IN 2006-15, "Vibration-Induced Degradation and Failure of Safety-Related Valves," 7/27/06
IN 2006-26, "Failure of Magnesium rotors In Motor-Operated Valve Actuators," 11/20/06
IN 1997-40, "Potential Nitrogen Accumulation Resulting from back-leakage from Safety Injection
Tanks," 6/26/97
IN 2006-21, "Operating Experience Regarding Entrainment of Air into Emergency Core Cooling
and Containment Spray Systems," 9/21/06
Miscellaneous Documents
LER 2007-004, "Docket Nos. 50-361 and 50-362; Licensee Event Report No. 2007-004; San
Onofre Nuclear Generating Station, Units 2 and 3," 12/19/07
Westinghouse Electric Company, LLC, Calculation Note Number CN-OA-03-45, Figure 1,
"Comparison of ABB-CE-W and EPRI WKM Closure Models for SONGS MSLB Re-analyses,"
Rev. 0
"Reload Ground Rules Figure IV-3 (MSIV Closure Pattern)" and Table "MSIV Closure Pattern,"
from SONGS Units 2/3 Cycle 15 Reload Ground Rules, Rev. 2
Accident Analysis Topical Report, DBD-SO23-TR-AA, Section 4.4.5, "Containment Peak
Pressure Analysis," Rev. 10
Operating Experience Report OE 25761, "Update to OE 25725 - Multiple Feedwater and MSIVs
Disabled By Solenoid Control Valve problems (SONGS)"
Licensee Event Report 2007-004, "Tech Spec Violation Caused By Moisture Contamination in
Hydraulic Dump Valve Solenoids," dated 12/19/2007
Performance monitoring Data System (Inservice Test Data for 2(3) HV8204 and HV8205) from
January 2006 to January 2008
Summary of Diagnostic traces over the last three cycles for Valves 2/3 HV9336, 2/3 HV9337,
and 2/3 HV9339
Summary data of temperature and humidity conditions inside containment at the 35 level
(location of valves 2/3 HV 9337 and 2/3 HV 9339)
Memorandum to File dated September 10, 1999, with subject of "Input Data for Containment
Flood Level Calculation"
SD-SO23-740, "Safety Injection, Containment Spray and Shutdown Cooling System," Rev. 15
Reactivity Oversight Group (ROG) Meeting Minutes and Agenda from June 23, 2008
FSAR 15.6.3.2, Steam Generator Tube Rupture, May 2007
- 12 -
Attachment
Quarterly MOV Program Health Reports from 1st Quarter 2007 through 1st Quarter 2008
Generic Letter 89-10 Program data Base
Substitution Equivalency Evaluation Report SEE 070058, dated 8/02/07
SD-SO23-160, "Main and Reheat Steam System," Rev. 18
1814-AA018-M0002, Vendor Manual - Salt Water Cooling Pumps, Rev. 0
SO23-404-4-113, Vendor Manual - CCW Heat Exchangers, Rev. 1
Memorandum, Input Data for Containment Flood Level Calculation, dated 9/10/99
SONGS Letter to Division of NRR, USNRC, dated 10/29/90
SONGS Letter to Division of NRR, USNRC, dated 6/28/91
SONGS System Health Report, SWC System, 1st Quarter 2008
SONGS System Health Report, CCW System, 1st Quarter 2008
SONGS System Health Report - Main Steam: 1st Quarter 2008
Refrigerant Tracking Log for various Safety Chillers (including E335)
Field Change Notice F23273M, "Auxiliary Building Emergency Chilled Water System," Rev. 6
Field Change Notice F23271M, "Auxiliary Building Emergency Chilled Water System," Rev. 6
Communication from Michael Jones dated 2/6/92, "Emergency Chiller Oil Temp Limits"
SD-SO23-120, "6.9 kV, 4.16 kV and 480 V Electrical Distribution System," Rev. 19
SD-SO23-130, "120 VAC Class 1E Electrical Distribution System," Rev. 12
SD-SO23-140, "1E and Non-1E 125 and 250 VDC Systems," Rev.15
SD-SO23-160, "Main and Reheat Steam System," Rev. 18
SD-SO23-250, "Main Feedwater System," Rev. 15
SD-SO23-360, "Reactor Coolant System," Rev. 16
SD-SO23-390, "Chemical and Volume Control System," Rev. 17
SD-SO23-400, "Component Cooling Water System," Rev. 18
SD-SO23-410, "Saltwater Cooling System," Rev. 7
SD-SO23-618, "Control Room Annunciators," Rev. 0
SD-SO23-720, "Engineered Safety Features Actuation System," Rev. 8
SD-SO23-780, "Auxiliary Feedwater System," Rev. 10
Email dated 07/10/2208 Dale Wickam to Paul Blake et. al., 3BD 21 EOC data summary (ref
20047962 notification)
LER 2005-001, "Loose electrical connections affecting Unit 3 Diesel Generators," 8/23/08
IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement
of Large Lead Storage Batteries for Generating Stations and Substations," Rev. 1980
- 13 -
Attachment
List of Abbreviations
action request
apparent cause evaluation
Agencywide Documents Access and Management System
corrective action program
component cooling water
component design basis inspection
core damage frequency
CFR
Code of Federal Regulations
control room supervisor
design basis document
DCE
direct cause evaluation
emergency operating procedure
Final Safety Analysis Report
high pressure safety injection
IEEE
Institute of Electrical and Electronic Engineers
IMC
Inspection Manual Chapter
Inservice Tests
IP
Inspection Procedure
LOVS
Loss of Voltage Signal
low pressure safety injection
MO
maintenance orders
motor-operated valve
net positive suction head
NRC
U.S. Nuclear Regulatory Commission
NN
Notification
NSPDP
nuclear safety professional development program
operating experience
PC
performance criteria
permanent plant modifications
problem identification and resolution
preventative maintenance
station blackout
steam generator tube rupture
Safety Injection Actuation Signal
structures, systems, and components
Saltwater Cooling
Updated Safety Analysis Report
VAC
Volts Alternating Current
VDC
Volts Direct Current