ML080440436
| ML080440436 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/13/2008 |
| From: | Clark J NRC/RGN-IV/DRP/RPB-E |
| To: | Rosenblum R Southern California Edison Co |
| References | |
| EA-08-051, FOIA/PA-2011-0157 IR-07-005 | |
| Download: ML080440436 (65) | |
See also: IR 05000361/2007005
Text
February 13, 2008
Richard M. Rosenblum
Senior Vice President and
Chief Nuclear Officer
Southern California Edison Company
San Onofre Nuclear Generating Station
P.O. Box 128
San Clemente, CA 92674-0128
SUBJECT: SAN ONOFRE NUCLEAR GENERATING STATION - NRC INTEGRATED
INSPECTION REPORT 05000361/2007005; 05000362/2007005 AND NOTICE OF
VIOLATION
Dear Mr. Rosenblum:
On December 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your San Onofre Nuclear Generating Station, Units 2 and 3 facility. The enclosed
integrated report documents the inspection findings, which were discussed on December 21,
2007, and February 13, 2008, with Mr. R. Ridenoure and other members of your staff.
The inspection examined activities conducted under your licenses as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your
licenses. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel.
One violation is cited in the enclosed Notice of Violation (Notice) and the circumstances
surrounding this violation are described in detail in the enclosed report. The violation involved
your failure to implement effective corrective actions to ensure thermal overloads associated
with safety-related equipment would not fail prematurely (EA-08-051). Although determined to
be of very low safety significance (Green), this violation is being cited because not all the
criteria specified in Section VI.A.1 of the NRC Enforcement Policy for a noncited violation (NCV)
were satisfied. Specifically, Southern California Edison failed to restore compliance within a
reasonable time after the violation was first identified in Inspection
Report 05000361;05000362/2006005. Please note that you are required to respond to this
letter and should follow the instructions specified in the enclosed Notice when preparing your
response. The NRC will use your response, in part, to determine whether further enforcement
action is necessary to ensure compliance with regulatory requirements.
This report also documents three NRC identified and self-revealing findings of very low safety
significance (Green). These findings were determined to involve violations of NRC
requirements. Additionally, one licensee-identified violation which was determined to be of very
low safety significance is listed in this report. However, because of the very low safety
Southern California Edison Company -2-
significance and because they were entered into your corrective action program, the NRC is
treating these findings as NCVs consistent with Section VI.A of the NRC Enforcement Policy. If
you contest these NCVs, you should provide a response within 30 days of the date of this
inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional
Administrator, U.S. Nuclear Regulatory Commission Region IV, 611 Ryan Plaza Drive,
Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear
Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at San
Onofre Nuclear Generating Station, Units 2 and 3, facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be made available electronically for public inspection
in the NRC Public Document Room or from the Publicly Available Records (PARS) component
of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Jeffrey A. Clark, Chief
Project Branch E
Division of Reactor Projects
Dockets: 50-361
50-362
Licenses: NPF-10
Enclosures:
Notice of Violation
NRC Inspection Report 05000361/2007005; 05000362/2007005
w/Attachment: Supplemental Information
cc w/enclosure:
Mr. Ross T. Ridenoure Gary L. Nolff
Vice President and Site Manager Assistant Director-Resources
Southern California Edison Company City of Riverside
San Onofre Nuclear Generating Station 3900 Main Street
P.O. Box 128 Riverside, CA 92522
San Clemente, CA 92674-0128
Mark L. Parsons
Chairman, Board of Supervisors Deputy City Attorney
County of San Diego City of Riverside
1600 Pacific Highway, Room 335 3900 Main Street
San Diego, CA 92101 Riverside, CA 92522
Southern California Edison Company -3-
Dr. David Spath, Chief Mr. James T. Reilly
Division of Drinking Water and Southern California Edison Company
Environmental Management San Onofre Nuclear Generating Station
California Department of Health Services P.O. Box 128
850 Marina Parkway, Bldg P, 2nd Floor San Clemente, CA 92674-0128
Richmond, CA 94804
Chief, Radiological Emergency
Michael J. DeMarco Preparedness Section
San Onofre Liaison National Preparedness Directorate
San Diego Gas & Electric Company Technological Hazards Division
8315 Century Park Ct. CP21G Department of Homeland Security
San Diego, CA 92123-1548 1111 Broadway, Suite 1200
Oakland, CA 94607-4052
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
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-
Electronic distribution by RIV:
ROPreports
Regional Administrator (EEC)
DRP Director (DDC)
DRS Director (RJC1)
DRS Deputy Director (ACC)
Senior Resident Inspector (CCO1)
Branch Chief, DRP/E (JAC)
Senior Project Engineer, DRP/E (GDR)
Senior Project Engineer, DRP/E (GBM)
Team Leader, DRP/TSS (CJP)
RITS Coordinator (MSH3)
V. Dricks, PAO (VLD)
D. Pelton, OEDO RIV Coordinator (DLP1)
SO Site Secretary (vacant)
MVasquez (GMV)
N Hilton, OE
John Wray, OE
Mary Ann Ashley, NRR
SUNSI Review Completed: _GBM__ ADAMS: WYes G No Initials: __GBM_
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R:\_REACTORS\_SO23\2007\SO2007-05RP-CCO.wpd ADAMS ML080440436
RIV:RI:DRP/E SRI:DRP/E SPE:DRP/E C:DRS/PSB C:DRS/OB
GMiller CCOsterholtz GReplogle MPShannon RELantz
/RA/ /RA teleph./ /RA electronic/ /RA/ /RA/
02/13/08 02/13/08 02/13/08 02/12/08 02/12/08
C:DRS/EB C:DRS/PEB SES/ACES C:DRP/E
RLBywater LJSmith GMVasquez JAClark
/RA/ /RA NOKeefe for/ /RA/ /RA GMiller for/
02/13/08 02/11/08 2/12/08 02/13/08
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NOTICE OF VIOLATION
Southern California Edison Co. Docket No. 50-361;362
San Onofre Nuclear Generating Station License No. NPF-10;15
EA 08-051
During an NRC inspection conducted on September 27 through December 31, 2007, a violation
of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the
violation is listed below:
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that
measures shall be established to ensure that for significant conditions adverse to
quality, the cause of the condition is determined and corrective action taken to preclude
repetition.
Contrary to this, from February 6 through August 8, 2007, the licensee failed to take
corrective actions to preclude repetition of the premature tripping of thermal overloads
for safety-related equipment, a significant condition adverse to quality.
This violation is associated with a Green SDP finding.
Pursuant to the provisions of 10 CFR 2.201, Southern California Edison Company is hereby
required to submit a written statement or explanation to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555 with a copy to the
Regional Administrator, Region IV, and a copy to the NRC Resident Inspector at the facility that
is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of
Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation;
EA-08-051" and should include: (1) the reason for the violation, or, if contested, the basis for
disputing the violation or severity level, (2) the corrective steps that have been taken and the
results achieved, (3) the corrective steps that will be taken to avoid further violations, and
(4) the date when full compliance will be achieved. Your response may reference or include
previous docketed correspondence, if the correspondence adequately addresses the required
response. If an adequate reply is not received within the time specified in this Notice, an order
or a Demand for Information may be issued as to why the license should not be modified,
suspended, or revoked, or why such other action as may be proper should not be taken.
Where good cause is shown, consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS), accessible from the
NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should
not include any personal privacy, proprietary, or safeguards information so that it can be made
available to the public without redaction. If personal privacy or proprietary information is
necessary to provide an acceptable response, then please provide a bracketed copy of your
response that identifies the information that should be protected and a redacted copy of your
response that deletes such information. If you request withholding of such material, you must
ENCLOSURE 1
specifically identify the portions of your response that you seek to have withheld and provide in
detail the bases for your claim of withholding (e.g., explain why the disclosure of information will
create an unwarranted invasion of personal privacy or provide the information required by
10 CFR 2.390(b) to support a request for withholding confidential commercial or financial
information). If safeguards information is necessary to provide an acceptable response, please
provide the level of protection described in 10 CFR 73.21.
Dated this 13th day of February, 2008
-2- ENCLOSURE 1
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-361, 50-362
Report No.: 05000361/2007005 and 5000362/2007005
Licensee: Southern California Edison Co. (SCE)
Facility: San Onofre Nuclear Generating Station, Units 2 and 3
Location: 5000 S. Pacific Coast Hwy.
San Clemente, California
Dates: September 27, 2007 through December 31, 2007
Inspectors: C. C. Osterholtz, Senior Resident Inspector, Project Branch E, DRP
M. O. Miller, Senior Resident Inspector, Project Branch E, DRP
M. R. Young, Resident Inspector, Project Branch E, DRP
G. Warnick, Senior Resident Inspector, Project Branch D, DRP
R. A. Kopriva, Senior Reactor Inspector, Engineering Branch 1, DRS
J. H. Nadel, Reactor Inspector, Engineering Branch 1, DRS
G. A. George, Reactor Inspector, Engineering Branch 1, DRS
B. D. Baca, Health Physics Inspector, Plant Support Branch, DRS
L. T. Ricketson, Senior Health Physics Inspector, Plant Support
Branch, DRS
S. T. Makor, Reactor Inspector, Engineering Branch 1, DRS
J. P. Adams, Reactor Inspector, Engineering Branch 1, DRS
L. E. Ellershaw, Senior Reactor Inspector, Engineering Branch 1, DRS
M. T. Baquera, Reactor Inspector, Engineering Branch 1, DRS
K. Clayton, Senior Operations Engineer, Operations Branch, DRS
Approved By: Jeffrey A. Clark, Chief
Project Branch E
Division of Reactor Projects
-1- ENCLOSURE 2
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3-
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -6-
1R02 Evaluations of Changes, Tests, or Experiments . . . . . . . . . . . . . . . . . . . . . . . -6-
1R04 Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -7-
1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -8-
1R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -9-
1R11 Licensed Operator Requalification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -17-
1R12 Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -18-
1R13 Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . -20-
1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -20-
1R17 Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -23-
1R19 Postmaintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -23-
1R20 Refueling and Other Outage Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -24-
1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -25-
1R23 Temporary Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -25-
1EP6 Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -26-
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -27-
2OS1 Access Control To Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . -27-
2OS2 Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -29-
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -30-
4OA1 Performance Indicator (PI) Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . -30-
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . -32-
4OA5 Other . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -36-
4OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -38-
4OA7 Licensee-Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -39-
ATTACHMENT: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-20
-2- ENCLOSURE 2
SUMMARY OF FINDINGS
IR05000361/2007005, 05000362/2007005; 09/27/07 - 12/31/07; San Onofre Nuclear
Generating Station, Units 2 & 3; Integrated Resident and Regional Report; Emergent Work,
Operability Evaluations, Occupational Radiation Safety, Problem Identification and Resolution.
This report covered a 3-month period of inspection by resident inspectors and Regional office
inspectors. The inspection identified four Green findings consisting of one cited violation and
three noncited violations. The significance of most findings is indicated by their color (Green,
White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance Determination
Process." Findings for which the significance determination process does not apply may be
Green or be assigned a severity level after NRC management's review. The NRCs program
for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a Green noncited violation of
10 CFR 50.65(a)(2) associated with the failure to include Units 2 and 3
emergency diesel generator (EDG) automatic voltage regulator (AVR)
deficiencies as functional failures in the maintenance rule program. The
inspectors noted that the voltage regulator deficiencies should have placed the
emergency diesel generators into Maintenance Rule 10 CFR 50.65(a)(1) status
approximately 6 months after the failures occurred. This caused a lapse in the
determination of appropriate system monitoring and goal setting to maintain
system reliability. This issue was entered into the licensee's corrective action
program as Action Request 070300161.
This finding was associated with the mitigating systems cornerstone. This issue
was similar to non-minor Example 7.b of Manual Chapter 0612, Appendix E, in
that the finding was more than minor since violations of 10 CFR 50.65(a)(2)
necessarily involve degraded system performance. This finding is not suitable
for evaluation using the Significance Determination Process because the
performance deficiency did not cause the degraded equipment performance.
This is a Category II finding per Inspection Procedure 71111.12, so it was
determined to have very low safety significance (Green) by management
judgement per Manual Chapter 0609, Appendix M. The cause of the finding has
a crosscutting aspect in the area of problem identification and resolution
associated with the corrective action program (P.1©) because the licensee failed
to thoroughly evaluate the cause and extent of condition of the failed emergency
diesel generator automatic voltage regulator (Section 1R12).
- Green. The inspectors identified a Green noncited violation of Technical
Specification 5.5.1.1 associated with the failure to implement procedural
guidance to ensure the proper application of a submersible pump to prevent
wetting of the steam supply to the Unit 2 turbine-driven auxiliary feedwater pump.
-3- ENCLOSURE 2
If the water level were to wet the steam line insulation, it would cause
condensation in the steam line and render the auxiliary feedwater pump
inoperable due to possible water hammer or turbine overspeed on a pump start.
This issue was entered into the licensees corrective action program as Action
Request 071000309.
The finding was more than minor because it was associated with the design
control attribute of the mitigating systems cornerstone and impacted the
cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events. Using Manual Chapter 0609,
Significance Determination Process, Phase 1 worksheet, the finding was
determined to have very low safety significance (Green) because it did not result
in a loss of safety function and did not affect the risk of external initiators. The
finding had a crosscutting aspect in the area of problem identification and
resolution associated with the corrective action program (P.1©) in that the
licensee did not thoroughly evaluate the problem such that the resolutions
address causes and extent of conditions (Section 1R15).
- Green. A self-revealing Green violation of 10 CFR Part 50, Appendix B,
Criterion XVI, was identified for the failure to prevent recurrence of premature
tripping of Square D thermal overloads used for equipment protection on safety-
related equipment. The licensee failed to scope the thermal overloads
associated with the Unit 3 saltwater cooling pump room because they had
previously determined that it had sufficient margin such that it would not be
susceptible to failure. This resulted in the premature tripping of thermal
overloads for the Unit 3 saltwater cooling pump room intake structure fan on
August 8, 2007. This issue was entered into the licensee's corrective action
program as Action Request 070800454.
The finding was determined to be more than minor because it was associated
with the equipment performance attribute of the mitigating systems cornerstone
and it affected the cornerstone objective by challenging the availability and
capability of safety-related components. The inspectors also noted that this a
repetitive problem in implementing corrective actions. Based on the results of
the Significance Determination Process Phase 1 evaluation, the finding was
determined to have very low safety significance because it did not result in an
actual loss of a system safety function, a loss of a single train of safety
equipment for greater than its Technical Specification allowed outage time, and
did not screen as potentially risk significant due to seismic, flooding, or severe
weather initiating events. This finding also had crosscutting aspects in the area
of problem identification and resolution associated with the corrective action
program (P.1©) because the licensee failed to thoroughly evaluate the extent of
condition of insufficient solder material on safety-related thermal overloads
(Section 4OA2).
-4- ENCLOSURE 2
Cornerstone: Occupational Radiation Safety
- Green. The inspector reviewed a self-revealing noncited violation of Technical
Specification 5.5.1.1 when a worker failed to follow radiation work permit
instructions. On July 14, 2007, after completing a pre-job site review, a worker
proceeded to verify work authorization boundaries in Unit 3, Room 209, without
contacting radiation protection for current radiological conditions and discussing
the work scope and locations as required by the radiation work permit. The
worker approached Valve S31902MU012 and received a dose rate alarm. The
maximum dose rate levels in the area were 30 millirem per hour on contact with
the piping system and 12 millirem per hour at 30 centimeters. The licensees
corrective actions were to coach the worker and to develop and implement a
mechanism to communicate associated boundary walk downs in maintenance
orders.
The failure to follow a radiation work permit instruction is a performance
deficiency. This finding is greater than minor because it is associated with one of
the cornerstone attributes (exposure control) and affected the Occupational
Radiation Safety cornerstone objective, in that workers not following their
radiation work permit does not ensure adequate protection of the worker health
and safety from additional personnel exposure. The finding was determined to
be of very low safety significance because it did not involve: (1) as low as is
reasonably achievable planning and controls, (2) an overexposure, (3) a
substantial potential for overexposure, or (4) an impaired ability to assess dose.
Further, this finding had a human performance crosscutting aspect in the work
practices component because the workers did not use human error prevention
techniques, such as self checking, to ensure the full work scope, locations, and
radiological conditions were discussed with radiation protection personnel as
required by the radiation work permit H4a] (Section 2OS1).
B. Licensee-Identified Violations
Violations of very low safety significance which were identified by the licensee have
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. These violations and
their corrective actions are listed in Section 4OA7 of this report.
-5- ENCLOSURE 2
REPORT DETAILS
Summary of Plant Status
Unit 2 began the inspection period at 99 percent power. On October 20, 2007, Unit 2 was
shutdown to Mode 3 to perform an extent of condition review as a result of Unit 3 main steam
isolation valve, main feedwater isolation valve, and main feedwater block valve solenoid
failures. The surveillance tests for Unit 2 valves that contained the specific solenoids in
question were performed when Unit 2 was in Mode 3. All surveillance tests were completed
satisfactory. Unit 2 was to restart on October 21, 2007, but did not begin restart until
October 25, 2007, due to complications with the Southern California brush fires. Unit 2
returned to power operation on October 26, 2007.
On November 26, 2007, Unit 2 was shutdown and cooled down for a planned refueling outage.
Unit 2 entered Mode 6 and began core alterations on December 7, 2007. Unit 2 was still in the
refueling outage at the end of the inspection period.
Unit 3 began the inspection period at 99.9 percent. On October 9, 2007, the licensee
performed a shutdown of Unit 3 for a planned mid-cycle outage. Unit 3 was returned to power
operation on November 9, 2007, and ended the inspection period at approximately 99.9 percent
reactor power.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments (71111.02)
a. Inspection Scope
The inspectors reviewed the effectiveness of the licensees implementation of changes
to the facility structures, systems, and components (SSC); risk-significant normal and
emergency operating procedures; test programs; and the Updated Final Safety Analysis
Report (UFSA) in accordance with 10 CFR 50.59, Changes, Tests, and Experiments.
The inspectors utilized Inspection Procedure 71111.02, Evaluation of Changes, Tests,
or Experiments, for this inspection.
The inspectors reviewed eight safety evaluations performed by the licensee since the
last NRC inspection of this area at San Onofre Nuclear Generating Station. The
evaluations were reviewed to verify that licensee personnel had appropriately
considered the conditions under which the licensee may make changes to the facility or
procedures or conduct tests or experiments without prior NRC approval. The inspectors
reviewed 33 screenings, in which licensee personnel determined that evaluations were
not required, to ensure that the exclusion of a full evaluation was consistent with the
requirements of 10 CFR 50.59. Evaluations and screenings reviewed are listed in the
attachment to this report.
The inspectors reviewed and evaluated a sample of recent licensee action requests to
determine whether the licensee had identified problems related to 10 CFR Part 50.59
-6- ENCLOSURE 2
evaluations, entered them into the corrective action program (CAP), and resolved
technical concerns and regulatory requirements. The reviewed action requests are
identified in the Attachment.
The inspection procedure specifies that the inspectors review a minimum sample of
six licensee safety evaluations and 12 applicability determinations and screenings
(combined). The inspectors completed a review of eight licensee safety evaluations and
33 screenings.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial System Walkdowns
a. Inspection Scope
The inspectors: (1) walked down portions of the three listed risk important systems and
reviewed plant procedures and documents to verify that critical portions of the selected
systems were correctly aligned; and (2) compared deficiencies identified during the walk
down to the licensee's UFSAR and CAP to ensure problems were being identified and
corrected.
- October 18, 2007, Unit 3, Shutdown Cooling Train B prior to mid-loop operations
- October 29, 2007, Unit 3, Train B containment spray pump (P013) used as
backup to shutdown cooling
- December 18, 2007, Unit 2, electrical alignment to safety Bus 2A06 while 2A04
is out of service
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed three samples.
b. Findings
No findings of significance were identified.
.2 Complete System Walkdown
a. Inspection Scope
The inspectors: (1) reviewed plant procedures, drawings, the UFSAR, Technical
Specifications (TS), and vendor manuals to determine the correct alignment of the
Unit 2 auxiliary feedwater system; (2) reviewed outstanding design issues, operator
workarounds, and UFSAR documents to determine if open issues affected the
-7- ENCLOSURE 2
functionality of the Unit 2 auxiliary feedwater system; and (3) verified that the licensee
was identifying and resolving equipment alignment problems. Documents reviewed by
the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
a. Inspection Scope
Quarterly Inspection
The inspectors walked down the six listed plant areas to assess the material condition of
active and passive fire protection features and their operational lineup and readiness.
The inspectors: (1) verified that transient combustibles and hot work activities were
controlled in accordance with plant procedures; (2) observed the condition of fire
detection devices to verify they remained functional; (3) observed fire suppression
systems to verify they remained functional and that access to manual actuators was
unobstructed; (4) verified that fire extinguishers and hose stations were provided at their
designated locations and that they were in a satisfactory condition; (5) verified that
passive fire protection features (electrical raceway barriers, fire doors, fire dampers,
steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory
material condition; (6) verified that adequate compensatory measures were established
for degraded or inoperable fire protection features and that the compensatory measures
were commensurate with the significance of the deficiency; and (7) reviewed the UFSAR
to determine if the licensee identified and corrected fire protection problems.
C October 2, 2007, Unit 2, emergency diesel Generator (EDG) 2G002 room
C October 2, 2007, Unit 2, EDG 2G003 room
C October 2, 2007, Unit 3, EDG 3G002 room
C October 2, 2007, Unit 3, EDG 3G003 room
- November 14, 2007, Unit 2, emergency core cooling system pump Room 002
- December 5, 2007, Unit 2, containment
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed six samples.
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b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07A)
a. Inspection Scope
The inspectors reviewed licensee programs, verified performance against industry
standards and reviewed critical operating parameters and maintenance records for the
Unit 3 Train B component cooling water heat Exchanger S31203ME002. The inspectors
verified that: (1) performance tests were satisfactorily conducted for heat
exchangers/heat sinks and reviewed for problems or errors; (2) the licensee utilized the
periodic maintenance method outlined in Electric Power Research Institute (EPRI)
NP- 7552, "Heat Exchanger Performance Monitoring Guidelines;" (3) the licensee
properly utilized biofouling controls; (4) the licensees heat exchanger inspections
adequately assessed the state of cleanliness of their tubes, and (5) the heat exchanger
was correctly categorized under the Maintenance Rule. Documents reviewed by the
inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R08 Inservice Inspection Activities (71111.08)
.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water
Reactor Vessel Upper Head Penetration Inspections, Boric Acid Corrosion Control
a. Inspection Scope
The inspection procedure requires review of two or three types of nondestructive
examination (NDE) activities and, if performed, one to three welds on the reactor coolant
system (RCS) pressure boundary.
The inspectors directly observed the following nondestructive examinations:
System Component/Weld ID Exam Type
RCS Surge Nozzle to Safe End Weld, 02-005-031 PT/UT
RCS Shutdown Cooling Piping 10" SCH 140 PT/UT
Pipe-Valve, 02-059-008
RCS Shutdown Cooling Piping 16" SCH 160 PT/UT
Pipe-Elbow, 02-059-002
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RCS Shutdown Cooling piping 16" SCH 160 PT/UT
Pipe-Valve, 02-059-001
The inspectors reviewed the following NDEs through record review:
System Component/Weld ID Exam Type
RCS Y-Stop Valve, 02-021-068 VT3
RCS Y-Stop Valve, 02-021-081 VT3
RCS Guide & Y-Stop Valve, 02-039-058 VT3
Feedwater Guide & Y-Stop Valve, 02-045-037 VT3
RCS 10" SCH 140 Reducer Tee-Pipe, 02-021-038 UT
The inspectors observed the initial Ultrasonic Examination System calibration for the
Panametrics Epoch 4 instrument, S/N 040229207, which was recorded on Ultrasonic
Instrument Calibration Data Record and Certification. The inspectors reviewed Table 1
in Electric Power Research Institute's PDI Protocol PDI-UT-2, Revision 20, dated 25
APR 07, to verify that the transducers to be used for ultrasonic examinations on
stainless steel piping were appropriately qualified.
The inspectors reviewed the NDE personnel qualification records for those contractor
personnel (Lambert MacGill Thomas, Inc. or LMT) performing ASME Code Section XI
inservice inspections. The LMT personnel had been appropriately certified using LMT's
procedure QA-46, "Qualification and Certification of NDE and Visual Examination
Personnel per ASME Section XI," Revision 0. The inspectors verified that the
requirements in QA-46 were consistent with ASNT CP-189-1995, ASNT Standard for
Qualification and Certification of Nondestructive Testing Personnel, 1995 Edition.
The inspection procedure further required verification of one to three welds on Class 1
or 2 pressure boundary piping to ensure that the welding process and welding
examinations were performed in accordance with the ASME code. The inspectors
observed portions of the preemptive structural weld overlay on the ASME code Class 1
pressurizer surge line nozzle-to-safe end dissimilar weld and pipe-to-safe end stainless
steel weld identified as follows:
System Component/Weld Identification
Pressurizer Surge Weld DMW 02-0005-031and Weld 02-016-001 Gas
Line Nozzle-to-Safe Tungsten Arc Welding (machine)
End-to-Pipe
Welding procedures and NDE of the welding repair conformed to ASME code
requirements and licensee commitments.
-10- ENCLOSURE 2
Welder qualification documentation packages and welder maintenance logs were
reviewed for all contract welders (Welding Services, Inc.) performing welding activities
on the pressurizer surge nozzle. The documentation packages and logs were in
accordance with Article III, QW-300 "Welding Performance Qualification" in Section IX
of the ASME code.
Welding Procedure Specifications WPS 08-08-T-001-Butter SS, Revision 0, and
WPS 03-08-T-804-Bottom, Revision 0, were the welding procedures observed being
used during the weld overlay process on the pressurizer surge nozzle. The inspectors
reviewed the welding procedure specifications and their corresponding procedure
qualification records (identified in the Attachment) to verify that ASME Code required
essential variables for the gas tungsten arc welding process had been identified,
recorded in the procedure qualification record, and formed the basis for qualification of
the welding procedure specifications.
Additionally, the inspectors reviewed manual gas tungsten arc welding and shielded
metal arc welding performed on an ASME Code Class 3 component cooling water
by-pass line around the letdown heat exchanger. This welding consisted of carbon steel
pipe-to-pipe and pipe-to-fitting (4" and 8") welding using ER70S-6 and E7018 welding
filler material. The reviewed welds are identified as Weld Records WR2-07-212,
WR2-07-213, and WR2-07-210.
The inspectors verified, by review, that the Welding Procedure Specification (1-GT-SM)
had been properly qualified in accordance with the requirements of Section IX of the
ASME code. The inspectors verified that the essential variables for both the shielded
metal arc welding and the gas tungsten arc welding processes had been identified,
recorded in the procedure qualification record, and formed the bases for qualification of
the welding procedure specification.
The inspectors also observed the liquid penetrant examinations performed on the buffer
(stainless steel) layer and the transition bead (between the buffer layer and the dilution
layer). The buffer layer represents the initial stainless steel layer of the weld overlay
that started at a point on the stainless steel pipe and covered the pipe, pipe-to-safe end
weld, safe end, and ending as close as practical to the dissimilar metal weld fusion line,
without contacting the dissimilar metal weld. These examinations were recorded on
Liquid Penetrant Nondestructive Examination Report 104532-PT-001. The examination
personnel qualification records for the examiner performing the examination were
reviewed to verify that the individual was properly certified. Further, the inspectors
reviewed the liquid penetrant procedure (WSI QAP 9.21, Revision 1) to verify that it was
properly qualified in accordance with ASME code Section V requirements. Additionally,
the inspectors reviewed the Ultrasonic Examination Report of the ultrasonic examination
performed on December 10, 2007, of the weld overlay which was at a nominal thickness
of 0.30 inches at the examination time.
-11- ENCLOSURE 2
The inspectors also verified by observation that welding filler materials were properly
stored and controlled in accordance with Procedure SO 123-I-11.1. Welding Filler
Material Control Records, used to document issuance and return of welding filler
materials, were reviewed for those materials issued on December 13, 2007, to verify
that specified administrative controls regarding welders, materials (quantity and time
limits), and use of portable ovens or caddys were being implemented.
The inspection procedure required inspection of any augmented or industry initiation
examinations. The inspectors determined that the licensee had not performed such
examinations. Consequently, the inspectors did not perform any activities in this area.
b. Findings
No findings of significance were identified.
.2 Vessel Upper Head Penetration (VUHP) Inspection Activities
a. Inspection Scope
The licensee performed NDEs of 100 percent of reactor VUHP. The inspector directly
observed a sample of the examinations performed on the control element drive
mechanism element (CEDM) and incore instrumentation (ICI) as listed below:
System Component/Weld Identification Examination Method
RCS ICI 96 UT/ET
RCS ICI 95 UT/ET
RCS ICI 94 UT/ET
RCS ICI 93 UT/ET
RCS RVUH vent line UT/ET
-12- ENCLOSURE 2
The NDEs were performed in accordance with the requirements of NRC Order
b. Findings
No findings of significance were identified.
.3 Boric Acid Corrosion Control Inspection (BACC) Activities
a. Inspection Scope
Resident inspectors observed a sample of BACC activities and verified that visual
inspections emphasized locations where boric acid leaks can cause degradation of
safety significant components.
The inspector reviewed five instances where boric acid deposits were found on reactor
coolant system piping components during the walkdown. The inspectors reviewed
licensee procedures governing the boric acid corrosion control program and inspector
qualifications, reviewed the extent of boric acid residue on the various components,
verified that the licensee inspectors who performed the walkdown were qualified, and
determined whether components that exhibited leakage during the current outage had
experienced leakage in the past. The following table lists the specific components
reviewed by the inspector, including the component numbers, brief component
descriptions, and the resulting Action Requests.
Component Number Description Action Request
2HV0512 Pressurizer surge line sample 070500261
isolation valve
2HV9203 Charging line insolation valve 071101172
2HV9201 Charging auxiliary spray 071101173
isolation valve
2HV9339 Shutdown cooling isolation 070500262
valve
2HV9326 Shutdown injection tank drain 070500265
valve
No boric acid leakage evaluations were performed for any of the instances where leaks
were identified during walkdowns.
The condition of the components was appropriately entered into the licensee's CAP and
corrective actions taken were consistent with ASME code requirements. No engineering
evaluations were required for any of the instances where leaks were identified during
walkdowns.
-13- ENCLOSURE 2
b. Findings
No findings of significance were identified.
.4 Steam Generator Tube Inspection Activities
a. Inspection Scope
The inspection procedure specified performance of an assessment of in-situ screening
criteria to assure consistency between assumed NDE flaw sizing accuracy and data
from the EPRI examination technique specification sheets. It further specified
assessment of appropriateness of tubes selected for in situ pressure testing,
observation of in situ pressure testing, and review of in situ pressure test results.
At the time of this inspection, no conditions had been identified that warranted in situ
pressure testing. The inspectors did, however, review the licensee's report for Units 2
and 3, Steam Generator Degradation Assessment for the Cycle 15 Refueling Outages
in 2007 and 2008, dated November 29, 2007, and compared the in situ test screening
parameters to the guidelines contained in the EPRI document In Situ Pressure Test
Guidelines, Revision 2, and the Combustion Engineering Owners Group screening
criteria. This review determined that the remaining screening parameters were
consistent with the EPRI and Combustion Engineering Owners Group guidelines.
In addition, the inspectors reviewed both the licensee site-validated and qualified
acquisition and analysis technique sheets used during this refueling outage and the
qualifying EPRI examination technique specification sheets to verify that the essential
variables regarding flaw sizing accuracy, tubing, equipment, technique, and analysis had
been identified and qualified through demonstration. The inspector reviewed acquisition
technique and analysis technique sheets are identified in the attachment.
The inspection procedure specified comparing the estimated size and number of tube
flaws detected during the current outage against the previous outage operational
assessment predictions to assess the licensee's prediction capability. The inspectors
compared the previous outage operational assessment predictions contained in
Report R-3671-00-1, Tube Degradation Predictions for the San Onofre Nuclear
Generating Station Unit 2 Steam Generators - 2006 Update, with the flaws identified
thus far during the current steam generator tube inspection effort. Compared to the
projected damage mechanisms identified by the licensee, the number of identified
indications fell within the range of prediction and were quite consistent with predictions.
No new damage mechanisms had been identified during this inspection.
The inspection procedure specified confirmation that the steam generator tube eddy
current test scope and expansion criteria meet TS requirements, EPRI guidelines, and
commitments made to the NRC. The inspectors evaluated the recommended steam
generator tube eddy current test scope established by TS requirements and the
licensees degradation assessment report. The inspectors compared the recommended
test scope to the actual test scope and found that the licensee had accounted for all
known flaws and had, as a minimum, established a test scope that met TS
-14- ENCLOSURE 2
requirements, EPRI guidelines, and commitments made to the NRC. The scope of the
licensee's eddy current examinations of tubes in both steam generators included:
- Bobbin examination full length of tubing (tube end hot-tube end cold) from both
hot and cold legs, in non-sleeved tubes, rows 4-147
- Bobbin examination of the unsleeved portion of tubing (sleeve top hot-tube end
cold) from the cold leg, in sleeved tubes, rows 4-147
- Bobbin examination of the straight length section of tubing from both hot and
cold legs, rows 1-3
- Rotating plug point coil examination of hot leg Tubsheet TSH +4", -13",
100 percent of all tubes
- Rotating plug point coil examination of cold leg tubesheet, TSC +2", -13",
100 percent of all tubes. Exception: Steam Generator 89 tubes R141-C63,
R140-C64, R139-C63, and surrounding tubes in 2-tube bounding pattern,
examination extent is TSC +4", -13".
hot), 100 percent of sleeved tubes
- Rotating plug point coil examination of SBF 0.00", -1.25" in Steam Generator 88,
Tube R28-C60 only
- Rotating plug point coil examination of U-bend section of tubing (07H-07C) with
mid/high frequency coil probe, 100 percent of tubes in rows 1-3
- Rotating plug point coil examination of U-bend section of tubing (07H-07C) with
mid-frequency coil probe, 20 percent sample of tubes in rows 4-10 (rows 5-10
sample drawn from tubes not examined with MRPC probe in the 2006
inspection)
- Rotating plug point coil examination of the following bobbin indications: ADR,
DNI, DEI,DSI, DTI, LPI, PLP, NQI, TWD (0-100 percent), DNT >= 2.0 volts, DNG
>= 4.0 volts, TSD, TSM, PDP, and CUD
- Rotating plug point coil examination of PLP indications (with LAR confirmation) in
a 2-tube bounding pattern, location +/- 1-inch of PLP edges
- Rotating plug point coil examination of all sections of tubing which cannot be
examined with the 600UL bobbin probe due to restriction
The inspection procedure specified, if new degradation mechanisms were identified,
verify that the licensee fully enveloped the problem in its analysis of extended conditions
including operating concerns and had taken appropriate corrective actions before plant
startup. To date, the eddy current test results had not identified any new degradation
mechanisms.
-15- ENCLOSURE 2
The inspection procedure requires confirmation that the licensee inspected all areas of
potential degradation, especially areas that were known to represent potential eddy
current test challenges (e.g., top-of-tubesheet, tube support plates, and U-bends). The
inspectors confirmed that all known areas of potential degradation were included in the
scope of inspection and were being inspected.
The inspection procedure further requires verification that repair processes being used
were approved in the TSs. The total number of tubes plugged was 133 tubes in Steam
Generator 88 and 125 tubes in Steam Generator 89. The inspectors verified that the
mechanical expansion plugging process to be used was an NRC-approved repair
process.
The inspection procedure also requires confirmation of adherence to the TS plugging
limit, unless alternate repair criteria have been approved. The inspection procedure
further requires determination whether depth sizing repair criteria were being applied for
indications other than wear or axial primary water stress corrosion cracking in dented
tube support plate intersections. The inspectors determined that the TS plugging limits
were being adhered to (i.e., 40 percent maximum through-wall indication).
If steam generator leakage greater than three gallons per day was identified during
operations or during post shutdown visual inspections of the tubesheet face, the
inspection procedure requires verification that the licensee had identified a reasonable
cause based on inspection results and that corrective actions were taken or planned to
address the cause for the leakage. The inspectors did not conduct any assessment
because this condition did not exist.
The inspection procedure requires confirmation that the eddy current test probes and
equipment were qualified for the expected types of tube degradation and an assessment
of the site-specific qualification of one or more techniques. The inspectors observed
portions of eddy current tests performed on the tubes in Steam Generators 88 and 89.
During these examinations, the inspectors verified that: (1) the probes appropriate for
identifying the expected types of indications were being used, (2) probe position location
verification was performed, (3) calibration requirements were adhered, and (4) probe
travel speed was in accordance with procedural requirements. The inspectors
performed a review of site-specific qualifications of the techniques being used. These
are identified in the attachment.
If loose parts or foreign material on the secondary side were identified, the inspection
procedure specified confirmation that the licensee had taken or planned appropriate
repairs of affected steam generator tubes and that they inspected the secondary side to
either remove the accessible foreign objects or perform an evaluation of the potential
effects of inaccessible object migration and tube fretting damage. At this time of the
inspection, no foreign material had been identified.
Finally, the inspection procedure specified review of one to five samples of eddy current
test data if questions arose regarding the adequacy of eddy current test data analyses.
The inspectors did not identify any results where eddy current test data analyses
adequacy was questionable.
-16- ENCLOSURE 2
b. Findings
No findings of significance were identified.
.5 Identification and Resolution of Problems
a. Inspection Scope
The inspection procedure requires review of a sample of problems associated with
inservice inspections documented by the licensee in the corrective action program for
appropriateness of the corrective actions.
The inspector reviewed corrective action reports which dealt with inservice inspection
activities and found the corrective actions were appropriate. Action requests reviewed
are listed in the documents reviewed section. From this review the inspectors
concluded that the licensee has an appropriate threshold for entering issues into the
corrective action program and has procedures that direct a root cause evaluation when
necessary. The licensee also has an effective program for applying industry operating
experience.
b. Findings
No findings of significance were identified. The inspectors completed one sample by
completing all required inspection activities.
1R11 Licensed Operator Requalification (71111.11)
.1 Quarterly Inspection
a. Inspection Scope
The inspectors observed testing and training of senior reactor operators and reactor
operators to identify deficiencies and discrepancies in the training, to assess operator
performance, and to assess the evaluator's critique. The training scenario on
October 22, 2007, involved just-in-time training for Unit 2 startup. Documents reviewed
by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.2 Annual Inspection
a. Inspection Scope
The inspectors reviewed the annual operating examination test results for 2007. Since
this was the first half of the biennial requalification cycle, the licensee was not required
-17- ENCLOSURE 2
to administer a written examination. These results were assessed to determine if they
were consistent with NUREG 1021, Operator Licensing Examination Standards for
Power Reactors, guidance and Manual Chapter 0609, Appendix I, Operator
Requalification Human Performance Significance Determination Process,
requirements. This review included the test results for a total of 15 crews composed of
87 licensed operators, which included: shift-standing senior operators, staff senior
operators, shift-standing reactor operators, and staff reactor operators. There were no
crew failures and no individual failures on the simulator scenario portion of the test.
There was one individual failure on the job performance measure portion of the test.
This individual was successfully remediated prior to returning to shift.
The inspector completed one sample.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors reviewed the listed maintenance activity to: (1) verify the appropriate
handling of SSC performance or condition problems; (2) verify the appropriate handling
of degraded SSC functional performance; (3) evaluate the role of work practices and
common cause problems; and (4) evaluate the handling of SSC issues reviewed under
the requirements of the maintenance rule, 10 CFR Part 50 Appendix B, and the TSs.
- October 1, 2007, Units 2 and 3, upgraded EDG automatic voltage regulators
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
Introduction. The inspectors identified a Green NCV of 10 CFR 50.65(a)(2) for the
failure to include Units 2 and 3 EDG automatic voltage regulator (AVR) deficiencies as
functional failures in the maintenance rule program. The inspectors noted that the
voltage regulator deficiencies should have placed the EDGs into maintenance rule
10 CFR 50.65(a)(1) status approximately six months after the failures occurred. This
caused a lapse in the determination of appropriate system monitoring and goal setting to
maintain system reliability.
Description. On March 3, 2007, the licensee identified that an AVR for the Unit 3 EDG
was oscillating excessively during a load test. The cause of the oscillation was poor
contact of the R3 potentiometer because of the open type housing of the potentiometers
which made them susceptible to dirt intrusion.
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The licensees analysis of the failed AVR concluded that the R3 potentiometer poor
contact caused the AVR to oscillate the EDG output voltage setting between zero and
3.8 megavolt ampere reactive (MVAR). Operations personnel subsequently declared
the EDG inoperable. All of the susceptible potentiometers on all eight EDGs were
subsequently upgraded to sealed multiturn gold plated potentiometers. The upgraded
installations were completed on August 26, 2007.
The inspectors discovered that the licensee had not evaluated the AVR deficiency in
their maintenance rule program for monitoring or goal setting. The inspectors
determined that the AVR failure impacted the reliability of the EDGs in accordance with
NUMARC 93-01, Nuclear Energy Institute Industry Guideline for Monitoring the
Effectiveness of Maintenance of Nuclear Power Plants, Revision 2. The inspectors
concluded that the AVR failure if correctly counted as a MPFF, would have caused the
EDG to exceed the performance criteria and should have been tracked for monitoring
and goal setting in the licensees maintenance rule program. In response to this finding,
the licensee subsequently placed the EDGs in 10 CFR 50.65(a)(1), and established an
EDG performance goal such that both Unit 2 and 3 EDG AVRs be successfully
surveillance tested four times each, with normal voltage and MVAR control, by the end
of the fourth quarter of 2007. Each EDG contains an AVRs A and B, therefore four
diesels each containing two AVRs would need to be surveillance tested four times to
successfully complete the goal.
Analysis. The failure to recognize the applicability of the maintenance rule for a failure
of the EDG AVR was a performance deficiency. This finding was associated with the
mitigating systems cornerstone. This issue was similar to non-minor Example 7.b of
Manual Chapter 0612, Appendix E, in that the finding was more than minor since
violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance.
This finding is not suitable for evaluation using the Significance Determination Process
because the performance deficiency did not cause the degraded equipment
performance. This is a Category II finding per Inspection Procedure 71111.12, so it was
determined to have very low safety significance (Green) by management judgement per
Manual Chapter 0609, Appendix M. The cause of the finding has a crosscutting aspect
in the area of problem identification and resolution associated with the CAP (P.1(c))
because the licensee failed to thoroughly evaluate the cause and extent of condition of
Enforcement. 10 CFR Part 50.65(a)(1) requires, in part, that holders of an operating
license shall monitor the performance or condition of SSCs within the scope of the rule
against licensee-established goals in a manner sufficient to provide reasonable
assurance that such SSCs are capable of fulfilling their intended safety functions.
10 CFR 50.65(a)(2) requires, in part, that monitoring specified in paragraph (a)(1) is not
required where it has been demonstrated the performance or condition of an SSC is
being effectively controlled through appropriate preventive maintenance, such that the
SSC remains capable of performing its intended function. Contrary to the above, from
March through September, 2007, the licensee failed to demonstrate the performance of
the EDGs was being effectively controlled through appropriate preventive maintenance
and did not establish goals to provide a reasonable assurance that the Units 2 and 3
EDGs were capable of fulfilling their intended function. Because the finding is of very
low safety significance and has been entered into the licensees CAP as AR 070300161,
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this violation is being treated as an NCV consistent with Section VI.A of the Enforcement
Policy: NCV 05000361;05000362/2007005-01, Failure to Properly Implement
Maintenance Rule Requirements for Emergency Diesel Generators.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
.1 Risk Assessment and Management of Risk
a. Inspection Scope
The inspectors reviewed the four below listed assessment activities to verify:
(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and
licensee procedures prior to changes in plant configuration for maintenance activities
and plant operations; (2) the accuracy, adequacy, and completeness of the information
considered in the risk assessment; (3) that the licensee recognizes, and/or enters as
applicable, the appropriate licensee-established risk category according to the risk
assessment results and licensee procedures; and (4) the licensee identified and
corrected problems related to maintenance risk assessments.
- October 4, 2007, Unit 3, risk assessment and management during an unplanned
emergency core cooling system TS 3.0.3 entry
- October 25, 2007, Unit 2, risk assessment and management during a startup
after unplanned shutdown and southern California fires
- October 12, 2007, Unit 3, risk assessment and management during a main
steam isolation valve dual indication
- November 30, 2007, Unit 2, risk assessment and management during the
Devers offsite power out of service - delayed midloop operations
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors: (1) reviewed plants status documents such as operator shift logs,
emergent work documentation, deferred modifications, and standing orders to
determine if an operability evaluation was warranted for degraded components;
(2) referred to the UFSAR and design basis documents to review the technical
adequacy of licensee operability evaluations; (3) evaluated compensatory measures
associated with operability evaluations; (4) determined degraded component impact on
-20- ENCLOSURE 2
any TSs; (5) used the Significance Determination Process to evaluate the risk
significance of degraded or inoperable equipment; and (6) verified that the licensee has
identified and implemented appropriate corrective actions associated with degraded
components.
- October 3, 2007, Units 2 and 3, incorrect calibration probe used for saltwater
cooling flow indicators
- October 4, 2007, Unit 2 turbine-driven auxiliary feedwater pump failed trench
eductor
- October 9, 2007, Unit 3, grounded pressurizer heater
- October 25, 2007, Unit 2 and 3, main feedwater isolation Valve 2HV4048 and
main steam isolation Valve 2HV8204 solenoid failed in-service testing
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples.
b. Findings
Introduction. The inspectors identified a Green NCV of TS 5.5.1.1 associated with the
failure to implement procedural guidance to ensure the proper application of a
submersible pump to prevent wetting of the steam supply to the Unit 2 turbine-driven
auxiliary feedwater pump. If the water level were to wet the steam line insulation, it
would cause condensation in the steam line and render the auxiliary feedwater pump
inoperable due to possible water hammer or turbine overspeed on a pump start.
Description. On October 4, 2007, during a plant walk-down, the inspectors noted that a
submersible pump was in use in a pipe trench in the Unit 2 auxiliary feedwater (AFW)
pump building while steam was discharging into the bottom of the pipe trench. The
pump was a temporary modification installed due to a failure of a permanently installed
eductor. The purpose of the eductor was to ensure water did not accumulate in the
trench such that it could contact the steam piping. If the water level were to wet the
steam line insulation, it would cause condensation in the steam line and render the
turbine-driven AFW pump inoperable due to the possibility of water hammer or
overspeed on turbine start.
The inspectors noted that the atmosphere in the top of the pipe trench felt very hot to
the touch. The inspectors then reviewed the vendor manual for the submersible pump
and hose and found that both had a maximum temperature rating of 140EF. The
inspectors concluded that water in the pipe trench could easily exceed the maximum
temperature rating for the submersible pump and hose rated of 140EF. Since this
temperature would exceed the rating of the pump and hose, the submersible pump
modification could not be relied upon to drain the trench. This could potentially render
the turbine driven AFW pump inoperable.
-21- ENCLOSURE 2
The inspectors interviewed the licensees staff and found that the submersible pump
and discharge hose had been installed per Procedure S023-2-16, Use of Temporary
Sump Pumps, Revision 20. The inspectors noted this procedure did not direct
consideration of the environment in which the pump would be used or the potential
consequences of failure of the pump, as would have been required by
Procedure S0123-XV-5.1, Temporary Modifications Control, Revision 8. Since the
failure of the submersible pump had the potential consequence of rendering safety-
related equipment inoperable, the inspectors concluded the procedure used to install the
modification was inadequate.
Corrective actions taken by the licensee included revising the Use of Temporary Sump
procedure to reflect the guidance found in the Temporary Modifications Control
procedure for consideration of the environmental effects on the submersible pump.
Additionally, the licensee revised Procedure OSM-5, Operator Rounds, Revision 7, and
replaced the submersible pump with one that was adequately temperature rated for the
environment in the AFW trench.
Analysis. The failure to have an adequate procedure resulting in an inadequate
modification with the potential to affect safety-related equipment was a performance
deficiency. The finding was more than minor because it was associated with the design
control attribute of the mitigating systems cornerstone and impacted the cornerstone
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events. Using Manual Chapter 0609, Significance Determination Process,
Phase 1 worksheet, the finding was determined to have very low safety significance
(Green) because it did not result in a loss of safety function and did not affect the risk of
external initiators. The finding had a crosscutting aspect in the area of problem
identification and resolution associated with the CAP (P.1(c)) in that the licensee did not
thoroughly evaluate the problem such that such that the resolutions address causes and
extent of conditions.
Enforcement. TS 5.5.1.1 requires that written procedures be established, implemented,
and maintained for activities specified in Appendix A, Typical Procedures for
Pressurized Water Reactors and Boiling Water Reactors, of Regulatory Guide 1.33,
Quality Assurance Program Requirements (Operations), dated February 1978.
Regulatory Guide 1.33, Appendix A, Section 9.e recommends general procedures for
the control of maintenance and modification work. Contrary to this requirement, on
May 11, 2007, the licensee failed to implement appropriate procedures to control
modification work in the Unit 2 auxiliary feedwater steam supply trench to ensure the
trench would not fill up with water and render the Unit 2 turbine driven auxiliary
feedwater pump inoperable. Because this violation is of very low safety significance and
has been entered into the licensees CAP as AR 071000309, it is being treated as an
NCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000362/2007005-02, Failure to Implement Procedural Requirements for
Modifications in the Auxiliary Feedwater Steam Supply Trench.
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1R17 Permanent Plant Modifications (71111.17B)
a. Inspection Scope
The inspectors reviewed seven permanent plant modification packages and associated
documentation, such as implementation reviews, safety evaluation applicability
determinations, and screenings, to verify that they were performed in accordance with
regulatory requirements and plant procedures. The inspectors also reviewed the
procedures governing plant modifications to evaluate the effectiveness of the program
for implementing modifications to risk-significant SSCs, such that these changes did not
adversely affect the design and licensing basis of the facility.
Procedures and permanent plant modifications reviewed are listed in the attachment to
this report. Further, the inspectors interviewed the cognizant design and system
engineers for the identified modifications as to their understanding of the modification
packages and process.
The inspectors evaluated the effectiveness of the licensees corrective action process to
identify and correct problems concerning the performance of permanent plant
modifications by reviewing a sample of related condition reports. The reviewed
condition reports are identified in the attachment.
The inspection procedure specifies inspectors review a required minimum sample of six
permanent plant modifications. The inspectors completed review of seven permanent
plant modifications.
b. Findings
No findings of significance were identified.
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors selected the six listed postmaintenance test activities of risk significant
systems or components. For each item, the inspectors: (1) reviewed the applicable
licensing basis and/or design-basis documents to determine the safety functions;
(2) evaluated the safety functions that may have been affected by the maintenance
activity; and (3) reviewed the test procedure to ensure it adequately tested the safety
function that may have been affected. The inspectors either witnessed or reviewed test
data to verify that acceptance criteria were met, plant impacts were evaluated, test
equipment was calibrated, procedures were followed, jumpers were properly controlled,
the test data results were complete and accurate, the test equipment was removed, the
system was properly re-aligned, and deficiencies during testing were documented. The
inspectors also reviewed the UFSAR to determine if the licensee identified and
corrected problems related to post maintenance testing.
- October 25, 2007, Unit 2, main steam isolation Valve 2HV8204, Train A & B, fail
safe closure postmaintenance test
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fail safe closure postmaintenance test
- October 29, 2007, Unit 3, Pressurizer Surge Line Nozzle Field Weld OVL-031,
post weld overlay liquid penetrant postmaintenance test
- October 31, 2007, Unit 3, reactor coolant gas vent system postmaintenance test
- November 3, 2007, Unit 3 reactor coolant gas vent system postmaintenance test
following corrective maintenance
- November 8, 2007, Unit 3, saltwater cooling Pump 3P112 postmaintenance test
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed six samples.
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
The inspectors reviewed the following risk significant refueling items or outage activities
to verify defense in depth commensurate with the outage risk control plan, compliance
with the TSs, and adherence to commitments in response to Generic Letter 88-17, Loss
of Decay Heat Removal: (1) the risk control plan; (2) tagging/clearance activities;
(3) reactor coolant system instrumentation; (4) electrical power; (5) decay heat removal;
(6) spent fuel pool cooling; (7) inventory control; (8) reactivity control; (9) containment
closure; (10) reduced inventory or midloop conditions; (11) refueling activities;
(12) heatup and coldown activities; (13) restart activities; and (14) licensee identification
and implementation of appropriate corrective actions associated with refueling and
outage activities. The inspectors' containment inspections included observations of the
containment sump for damage and debris; and observation of supports, braces, and
snubbers for evidence of excessive stress, water hammer, or aging. Documents
reviewed by the inspectors are listed in the attachment. The inspectors reviewed outage
activities for Unit 3 from October 9, 2007 to November 9, 2007. The inspectors also
reviewed outage activities for Unit 2 from November 26, 2007, until the end of the
inspection period.
The inspectors completed two samples.
b. Findings
No findings of significance were identified.
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1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the UFSAR, procedure requirements, and TSs to ensure that
the four listed surveillance activities demonstrated that the SSCs tested were capable of
performing their intended safety functions. The inspectors either witnessed or reviewed
test data to verify that the following significant surveillance test attributes were
adequate: (1) preconditioning; (2) evaluation of testing impact on the plant;
(3) acceptance criteria; (4) test equipment; (5) procedures; (6) jumper/lifted lead
controls; (7) test data; (8) testing frequency and method demonstrated TS operability;
(9) test equipment removal; (10) restoration of plant systems; (11) fulfillment of ASME
Code requirements; (12) updating of performance indicator data; (13) engineering
evaluations, root causes, and bases for returning tested SSCs not meeting the test
acceptance criteria were correct; (14) reference setting data; and (15) annunciators and
alarms setpoints. The inspectors also verified that the licensee identified and
implemented any needed corrective actions associated with the surveillance testing.
- August 1, 2007, Unit 2, 2HV-9900 normal chilled water to containment isolation
Valve 2HV-9900 stroke test
- October 4, 2007, Unit 3, Train A saltwater cooling outlet Valve 3HV6497 partial
manual stroke test
- October 18, 2007, Unit 2, high pressure safety injection Pump 2MP018 response
time testing
- October 18, 2007, Unit 2, component cooling water Pump 2MP024 inservice test
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications (71111.23)
a. Inspection Scope
The inspectors reviewed the UFSAR, plant drawings, procedure requirements, and TSs
to ensure that the below listed temporary modification was properly implemented. The
inspectors: (1) verified that the modifications did not have an affect on system
operability/availability; (2) verified that the installation was consistent with modification
documents; (3) ensured that the post-installation test results were satisfactory and that
the impact of the temporary modifications on permanently installed SSCs were
supported by the test; and (4) verified that appropriate safety evaluations were
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completed. The inspectors verified that licensee identified and implemented any needed
corrective actions associated with temporary modifications.
- October 9, 2007, Unit 3, swap grounded pressurizer Heater ME616 with
Heater E614
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance was identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation (71114.06)
a. Inspection Scope
For the listed drill and simulator-based training evolutions contributing to Drill/Exercise
Performance and Emergency Response Organization Performance Indicators, the
inspectors: (1) observed the training evolution to identify any weaknesses and
deficiencies in classification, notification, and Protective Action Recommendation
development activities; (2) compared the identified weaknesses and deficiencies against
licensee identified findings to determine whether the licensee is properly identifying
failures; and (3) determined whether licensee performance is in accordance with the
guidance of the NEI 99-02, "Voluntary Submission of Performance Indicator Data,"
acceptance criteria.
- October 3, 2007, Units 2 and 3 simulator, control room, technical support center,
operations support center, and emergency operations facility, Unit 3 diesel
Generator 3G003 fuel oil day tank fire, Unit 2 steam generator tube leak and
subsequent tube rupture with potential unfiltered radioactive release pathway
through the steam driven auxiliary feed Pump P-140 turbine exhaust
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
-26- ENCLOSURE 2
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control To Radiologically Significant Areas (71121.01)
a. Inspection Scope
This area was inspected to assess the licensees performance in implementing physical
and administrative controls for airborne radioactivity areas, radiation areas, high
radiation areas, and worker adherence to these controls. The inspector used the
requirements in 10 CFR Part 20, the technical specifications, and the licensees
procedures required by technical specifications as criteria for determining compliance.
During the inspection, the inspector interviewed the radiation protection manager,
radiation protection supervisors, and radiation workers. The inspector performed
independent radiation dose rate measurements and reviewed the following items:
- Performance indicator events and associated documentation packages reported
by the licensee in the Occupational Radiation Safety Cornerstone
- Controls (surveys, posting, and barricades) of radiation, high radiation, or
airborne radioactivity areas in the Auxiliary, Radwaste, Reactor, and
Containment Buildings
- Radiation exposure permits, procedures, engineering controls, and air sampler
locations
- Conformity of electronic personal dosimeter alarm set points with survey
indications and plant policy; workers knowledge of required actions when their
electronic personnel dosimeter noticeably malfunctions or alarms
- Barrier integrity and performance of engineering controls in two potential
airborne radioactivity areas
- Adequacy of the licensees internal dose assessment for any actual internal
exposure greater than 50 millirem committed effective dose equivalent
- Physical and programmatic controls for highly activated or contaminated
materials (non-fuel) stored within spent fuel and other storage pools.
- Self-assessments, audits, licensee event reports, and special reports related to
the access control program since the last inspection
- Corrective action documents related to access controls
- Licensee actions in cases of repetitive deficiencies or significant individual
deficiencies
- Radiation exposure permit briefings and worker instructions
-27- ENCLOSURE 2
- Adequacy of radiological controls, such as required surveys, radiation protection
job coverage, and contamination control during job performance
- Dosimetry placement in high radiation work areas with significant dose rate
gradients
- Changes in licensee procedural controls of high dose rate - high radiation areas
and very high radiation areas
- Controls for special areas that have the potential to become very high radiation
areas during certain plant operations
- Posting and locking of entrances to all accessible high dose rate - high radiation
areas and very high radiation areas
- Radiation worker and radiation protection technician performance with respect to
radiation protection work requirements
The inspector completed 21 of the required 21 samples.
b. Findings
Introduction. The inspector reviewed a self-revealing NCV of TS 5.5.1.1 when a worker
failed to follow radiation work permit instructions.
Description. On July 14, 2007, a worker notified health physics of a pre-job site review
prior to starting work on Valve 3HV7261 in the Post Accident Sampling System Lab. The
worker was informed of the radiological conditions for the work area. However, after
completing the pre-job site review, the worker proceeded to verify the work authorization
boundaries in Unit 3, Room 209. The worker approached Valve S31902MU012 and
received a dose rate alarm. The worker exited the radiologically controlled area and
informed health physics of the alarm. The peak dose rate received by the worker was
11.1 millirem per hour and area around valve S31902MU012 had a maximum dose rate
level of 30 millirem per hour on contact with the piping system and 12 millirem per hour at
30 centimeters. During the licensees investigation of the dose rate alarm, the licensee
determined that the worker did not inform health physics of all areas needing access to
complete the work scope and did not receive a radiological briefing for Unit 3, Room 209.
The licensees corrective actions were to coach the worker and to develop and
implement a mechanism for communicating associated boundary walk downs in
maintenance orders.
Analysis. The failure to follow a radiation work permit instruction is a performance
deficiency. This finding is greater than minor because it is associated with one of the
cornerstone attributes (exposure control) and affected the Occupational Radiation Safety
cornerstone objective, in that workers not following their radiation work permit does not
ensure adequate protection of the worker health and safety from additional personnel
exposure. This occurrence involved a workers unplanned, unintended dose, or potential
for such a dose that could have been significantly greater as a result of a single minor,
-28- ENCLOSURE 2
reasonable alteration of the circumstances, higher dose rate levels. This finding was
determined to be of very low safety significance because it did not involve: (1) as low as
is reasonably achievable (ALARA) planning and controls, (2) an overexposure, (3) a
substantial potential for overexposure, or (4) an impaired ability to assess dose. Further,
this finding has a work practices human performance cross cutting aspect in human error
prevention techniques because the worker failed to self check the work scope and work
locations when briefing with health physics prior to entering the radiological controlled
area H4a].
Enforcement. Technical Specification 5.5.1.1.a requires applicable procedures
recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Section 7(e), of the Appendix, requires procedures for access control and a radiation
work permit system. Procedure SO 123-VII-20, Health Physics Program, Revision 12,
Section 6.10.6.5 states, in part, that individuals entering a radiological controlled area
sign on an appropriate radiation exposure permit acknowledging that they agree to
comply with the radiological controls specified on the radiation exposure permit.
Radiation Exposure Permit 07070562000/200159, states, in part, that workers, prior to
entering the radiologically controlled area, are to inform the Health Physics Control Point
of the job scope and work locations. Contrary to the Radiation Exposure Permit
requirement, on July 14, 2007, the worker did not inform the health physicist at the
control point of the full work scope and work locations prior to entering the radiological
controlled area which resulted in the worker knowing the current radiological conditions of
Room 209. Because this finding is of very low safety significance and was entered into
the licensees corrective action program (Action Request 070700545), this violation is
being treated as a noncited violation in accordance with Section VI.A.1 of the
Enforcement Policy: NCV 05000362/2007005-03, Failure to follow a radiation exposure
permit requirement.
2OS2 Planning and Controls (71121.02)
a. Inspection Scope
The inspector assessed licensee performance with respect to maintaining individual and
collective radiation exposures ALARA. The inspector used the requirements in 10 CFR
Part 20 and the licensees procedures required by technical specifications as criteria for
determining compliance. The inspector interviewed licensee personnel and reviewed:
- Site-specific ALARA procedures
- Interfaces between operations, radiation protection, maintenance, maintenance
planning, scheduling and engineering groups
- Integration of ALARA requirements into work procedure and radiation work permit
(or radiation exposure permit) documents
- Dose rate reduction activities in work planning
- Exposure tracking system
-29- ENCLOSURE 2
- Use of engineering controls to achieve dose reductions and dose reduction
benefits afforded by shielding
- Workers use of the low dose waiting areas
- First-line job supervisors contribution to ensuring work activities are conducted in
a dose efficient manner
- Radiation worker and radiation protection technician performance during work
activities in radiation areas, airborne radioactivity areas, or high radiation areas
- Self-assessments, audits, and special reports related to the ALARA program
since the last inspection
- Resolution through the corrective action process of problems identified through
post-job reviews and post-outage ALARA report critiques
- Corrective action documents related to the ALARA program and follow-up
activities, such as initial problem identification, characterization, and tracking
- Effectiveness of self-assessment activities with respect to identifying and
addressing repetitive deficiencies or significant individual deficiencies
The inspector completed 5 of the required 15 samples and 8 of the optional samples.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification (71151)
a. Inspection Scope
Cornerstone: Mitigating Systems
The inspectors sampled licensee data for the Mitigating System Performance
Index (MSPI) performance indicators (PI) listed below for Units 2 and 3 for the period
from September 26, 2007 through December 31, 2007. The definitions and guidance of
Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator
Guideline," Revision 4, were used to verify the licensees basis for reporting unavailability
and unreliability in order to verify the accuracy of PI data. The inspectors reviewed
operating logs, Limiting Conditions for Operation logs, ARs, and the maintenance rule
database to verify that the licensee properly accounted for planned and unplanned
unavailability as part of the assessment. The inspectors sampled data to verify that the
licensee: (1) accurately documented the actual unavailability hours for the MSPI systems;
and (2) accurately documented the actual unreliability information for each MSPI
-30- ENCLOSURE 2
monitored component. In addition, the inspectors interviewed licensee personnel
associated with PI data collection and evaluation.
- Units 2 and 3, safety system functional failures
The inspectors completed two samples.
Cornerstone: Barrier Integrity
The inspectors sampled licensee submittals for the four performance indicators listed
below for the period September 26, 2007 through December 31, 2007, for Units 2 and 3.
The definitions and guidance of Nuclear Energy Institute 99-02, Regulatory Assessment
Performance Indicator Guideline, Revision 4, were used to verify the licensees basis for
reporting each data element in order to verify the accuracy of PI data reported during the
assessment period. The inspectors: (1) reviewed RCS chemistry sample analyses for
dose equivalent Iodine-131 and compared the results to the TS limit; (2) observed a
chemistry technician obtain and analyze a RCS sample; (3) reviewed operating logs and
surveillance results for measurements of RCS identified leakage; and (4) observed a
surveillance test that determined RCS identified leakage. Licensee performance
indicator data were also reviewed for the following:
C Units 2 and 3, reactor coolant system specific activity
C Units 2 and 3, reactor coolant system leakage
The inspectors completed four samples.
Cornerstone : Occupational Radiation Safety
Occupational Exposure Control Effectiveness
The inspector reviewed licensee documents from January 1 through
September 30, 2007. The review included corrective action documentation that identified
occurrences in locked high radiation areas (as defined in the licensees technical
specifications), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned
personnel exposures (as defined in Nuclear Energy Institute (NEI) 99-02, Regulatory
Assessment Indicator Guideline, Revision 5). Additional records reviewed included
ALARA records and whole body counts of selected individual exposures. The inspector
interviewed licensee personnel that were accountable for collecting and evaluating the
performance indicator data. In addition, the inspector toured plant areas to verify that
high radiation, locked high radiation, and very high radiation areas were properly
controlled. Performance indicator definitions and guidance contained in NEI 99-02,
Revision 5, were used to verify the basis in reporting for each data element.
The inspector completed the required sample (1) in this cornerstone.
Cornerstone: Public Radiation Safety
Radiological Effluent Technical Specification/Offsite Dose Calculation Manual
Radiological Effluent Occurrences
-31- ENCLOSURE 2
The inspector reviewed licensee documents from January 1 through
September 30, 2007. Licensee records reviewed included corrective action
documentation that identified occurrences for liquid or gaseous effluent releases that
exceeded performance indicator thresholds and those reported to the NRC. The
inspector interviewed licensee personnel that were accountable for collecting and
evaluating the performance indicator data. Performance indicator definitions and
guidance contained in NEI 99-02, Revision 5, were used to verify the basis in reporting
for each data element.
The inspector completed the required sample (1) in this cornerstone.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1 Radiological Controls Review
a. Inspection Scope
The inspector evaluated the effectiveness of the licensees problem identification and
resolution process with respect to the following inspection areas:
- Access Control to Radiologically Significant Areas (Section 2OS1)
- ALARA Planning and Controls (Section 2OS2)
b. Findings
No findings of significance were identified.
.2 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
The inspectors performed a daily screening of items entered into the licensee's corrective
action program. This assessment was accomplished by reviewing maintenance orders,
action requests, the management focus list, and attending corrective action review and
work control meetings. The inspectors: (1) verified that equipment, human performance,
and program issues were being identified by the licensee at an appropriate threshold and
that the issues were entered into the corrective action program; (2) verified that
corrective actions were commensurate with the significance of the issue; and
(3) identified conditions that might warrant additional follow-up through other baseline
inspection procedures.
b. Findings
No findings of significance were identified.
-32- ENCLOSURE 2
.3 Selected Issue Follow-up Inspection
a. Inspection Scope
In addition to the routine review, the inspectors selected the two below listed issues for a
more in-depth review. The inspectors considered the following during the review of the
licensee's actions: (1) complete and accurate identification of the problem in a timely
manner; (2) evaluation and disposition of operability/reportability issues; (3) consideration
of extent of condition, generic implications, common cause, and previous occurrences;
(4) classification and prioritization of the resolution of the problem; (5) identification of
root and contributing causes of the problem; (6) identification of corrective actions; and
(7) completion of corrective actions in a timely manner.
C August 7, 2007, Unit 3, saltwater cooling pump room thermal overload trip
- December 18, 2007, Units 2 and 3, comprehensive review of operator
workarounds
Documents reviewed by the inspectors are listed in the attachment.
b. Findings
Introduction. A self revealing Green violation of 10 CFR Part 50, Appendix B,
Criterion XVI, was identified for the failure to prevent recurrence of premature tripping of
Square D thermal overloads used for equipment protection on safety-related equipment.
The licensee failed to scope the thermal overloads associated with the Unit 3 saltwater
cooling pump room because it had erroneously determined that it had sufficient margin
such that it would not be susceptible to failure. This resulted in the premature tripping of
thermal overloads for the Unit 3 saltwater cooling pump room intake structure fan on
August 8, 2007.
Description. The licensee previously had problems with spurious thermal overload trips
and received a noncited violation for untimely corrective actions to resolve the problem
(see NRC Inspection Report 05000361;362/2006-005). On October 17, 2006, the Unit 2
fuel handling building pump room emergency air conditioning Unit 2E441 Phase B
thermal overload tripped for no apparent reason with the fan turned off. The inspectors
noted that six spurious trips of other thermal overloads had occurred since December
2005. These overloads were associated with the Unit 3 fuel handling building post
accident cleanup room emergency air conditioning Unit 3E371, the Unit 2 fuel handling
building pump room emergency air conditioning Units 2E441 and 2E442, and the Unit 2
component cooling water Pump 2P024 room emergency air conditioning Unit 2E453. All
of these thermal overloads were subsequently changed out for larger devices in 2005
because of chronic problems with spurious trips.
The inspectors reviewed the history of spurious thermal overload trips and discovered
that five previous apparent cause assessments (ACEs) had been performed since
January 2001 to identify and correct spurious trips associated with thermal overloads. A
2001 ACE identified equipment aging as the cause, and directed that replacement
thermal overloads be installed. A 2002 ACE identified degraded cabling lugs as the
-33- ENCLOSURE 2
cause, and the lugs were replaced. A 2003 ACE identified the cause as insufficient
margin in the trip settings, which were adjusted. A 2004 ACE attributed a series of
spurious trips to warm weather. Finally, a 2005 ACE identified that the thermal overloads
were undersized, and that new, larger thermal overloads should be installed. The
licensee upgraded 64 thermal overloads to a larger capacity model in December 2005.
However, the inspectors concluded that the ACEs and the associated corrective actions
generated by the licensee had been ineffective in resolving the problem.
The licensee performed a root cause evaluation as part of RCE070901311 initiated in
response to the thermal overload failures. Procedure SO123-XV-50, Corrective Action
Process, Revision 7, directs a root cause evaluation for significant problems and to
prevent recurrence of the consequences of these problems. The inspectors concluded a
root cause evaluation was appropriate since Procedure SO123-XV-50 specifies criteria
for a root cause that include safety equipment failures with generic operability issues and
long-standing problems requiring escalation for resolution. The inspectors determined
these criteria were met based on the generic implications involving failures of safety
related equipment and the numerous apparent causes that had been performed since
January 2001 that had failed to correct the issue. The inspectors therefore concluded
the failure of the thermal overloads represented a significant condition adverse to quality.
The licensee implemented a detailed plan for testing the thermal overloads and X-rayed
the internals to determine if a design defect had previously gone undetected. The
licensee discovered that two mechanisms in concert with each other were causing the
spurious trips. Thermal overloads associated with small motors had a tendency to trip
early due to higher than expected current levels going through the overloads while the
associated line voltage was high in the normal band. Also, the X-ray analysis revealed
that approximately 20 percent of the sample had insufficient melting alloy, contributing to
a thermal overload tripping on lower current.
The licensee established a plan to replace the affected thermal overloads with properly
sized components that would be X-rayed for sufficient melting alloy verification prior to
installation. However, the licensee concluded sufficient margin existed in a group of 75
thermal overloads, including those associated with the Unit 3 saltwater cooling pump
room intake structure fans.
On August 8, 2007, the intake structure fan for the Unit 3 saltwater cooling pump room
tripped. The cause was subsequently determined to be a defective thermal overload on
the Phase C portion due to insufficient solder material in the thermal overload. The
thermal overload was replaced, and temperature in the Unit 3 saltwater cooling pump
never approached its design value of 98°F. The licensee has since replaced all 75
susceptible thermal overloads that were previously scoped out of the corrective action
process.
Analysis. The failure of the licensee to properly scope corrective actions to prevent the
premature tripping of thermal overloads for safety-related equipment was considered a
performance deficiency. The finding was determined to be more than minor because it
was associated with the equipment performance attribute of the mitigating systems
cornerstone and it affected the cornerstone objective by challenging the availability and
capability of safety-related components. Using the Manual Chapter 0609, Significance
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Determination Process, Phase 1 worksheet, the finding was determined to have very low
safety significance (Green) because it did not result in an actual loss of a system safety
function, a loss of a single train of safety equipment for greater than its technical
specification allowed outage time, and did not screen as potentially risk significant due to
seismic, flooding, or severe weather initiating events. The cause of the finding has a
crosscutting aspect in the area of problem identification and resolution associated with
the corrective action program (P.1(c)) because the licensee failed to thoroughly evaluate
the extent of condition of insufficient solder material on safety-related thermal overloads.
Enforcement. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in
part, that measures shall be established to ensure that for significant conditions adverse
to quality, corrective actions are taken to preclude repetition. Contrary to this, from
February 6 through August 8, 2007, the licensee failed to take corrective actions to
preclude repetition of the premature tripping of thermal overloads for safety-related
equipment, a significant condition adverse to quality. This finding has been entered into
the licensee's corrective action program as AR 070800454. Due to the licensees failure
to restore compliance from previous NCV 05000361;05000362/2006005-04, within a
reasonable time after the violation was identified, this violation is being cited as a Notice
of Violation consistent with Section VI.A of the Enforcement Policy: VIO 05000361;05000362/2007005-04, Failure to Prevent Recurrence of Premature Tripping of Square
D Thermal Overloads.
.3 Semiannual Trend Review
a. Inspection Scope
The inspectors completed a semi-annual trend review of repetitive or closely related
issues that were documented to identify trends that might indicate the existence of more
safety significant issues, specifically in the areas of procedural compliance and human
performance. The inspectors review consisted of the six month period from June 25,
2007, through December 31, 2007. When warranted, some of the samples expanded
beyond those dates to fully assess the issue. The inspectors also reviewed corrective
action program items associated with human performance improvement, and met with
representatives from the San Onofre human performance improvement team at regular
intervals. Corrective actions associated with a sample of the issues identified in the
licensee's trend report were reviewed for adequacy. Documents reviewed by the
inspectors are listed in the attachment.
b. Findings
No findings of significance were identified. However, the inspectors noted that the
licensee continued to attempt to implement human performance initiatives to prevent
personnel errors. The licensee indicated that a stand alone performance improvement
plan would be implemented by January 31, 2008.
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4OA5 Other
.1 Temporary Instruction 2515/166, "Pressurized Water Reactor Containment Sump
Blockage," San Onofre Nuclear Generating Station, Unit 2
Temporary Instruction 2515/166 was performed at San Onofre Nuclear Generating
Station, Unit 2. The results of inspection phase of Temporary Instruction 2515/166 for
Unit 2 are subsequently documented in this report. Temporary Instruction 2515/166 for
both Unit 2 and Unit 3 will be closed out after the completion and verification of
modification commitments for Unit 2 containment sumps at the end of Refueling
Outage 15.
Listed below are the commitments and actions taken by the licensee:
1. Design and procurement of replacement sump screens
Actions Taken
Engineering Change Packet ECP#040301974-11 dated Jul 17, 2006, provides for
the design changes of containment sump to address sump blockage concerns.
This engineering change packet has undergone NRC review and supplemental
responses to the NRC are to be received no later than February 29, 2008, per
letter to Nuclear Energy Institute (NEI) from NRC: Supplemental Licensee
Responses to Generic Letter 2004-02, "Potential Impact Of Debris Blockage On
Emergency Recirculation During Design Basis Accidents At Pressurized-Water
Reactors," dated November 30, 2007. Materials for the sump screens have been
procured and are currently being installed during Refueling Outage RF15, with
modifications expected to complete at the end of the outage.
2. Resolution of potential susceptibility of emergency core cooling system and
containment spray system pump mechanical seal to increased leakage due to
debris mix passing through the seals
Actions Taken
The licensee has completed calculations to evaluate seal leakage due to debris
ingestion. This action has undergone NRC review and supplemental responses
to the NRC are to be received no later than February 29, 2008, per letter to NEI
from NRC: Supplemental Licensee Responses to Generic Letter 2004-02,
"Potential Impact Of Debris Blockage On Emergency Recirculation During Design
Basis Accidents At Pressurized-Water Reactors," dated November 30, 2007.
3. Resolution of potential susceptibility of ECCS and CSS pump mechanical seal
cyclone separators to debris blockage
-36- ENCLOSURE 2
Actions Taken
The licensee has completed calculations to evaluate seal leakage due to debris
ingestion. This action has undergone NRC review and supplemental responses to
the NRC are to be received no later than February 29, 2008, per letter to NEI
from NRC: Supplemental Licensee Responses to Generic Letter 2004-02,
"Potential Impact Of Debris Blockage On Emergency Recirculation During Design
Basis Accidents At Pressurized-Water Reactors," dated November 30, 2007.
4. Development of a reduced qualified protective coatings zone of influence (ZOI)
Actions Taken
ALION-CAL-SONGS2933-02, Revision 1 "San Onofre Units 2 and 3 GSI-191
Containment Recirculation Sump Evaluation: Debris Generation Calculation,"
documents the assumptions and methodology that the licensee applied to
determine the ZOI and debris generated for each postulated break. This
evaluation has undergone NRC review and supplemental responses to the NRC
are to be received no later than February 29, 2008, per letter to NEI from NRC:
Supplemental Licensee Responses to Generic Letter 2004-02, "Potential Impact
Of Debris Blockage On Emergency Recirculation During Design Basis Accidents
at Pressurized-Water Reactors," dated November 30, 2007.
5. Validation of the 8 percent head loss margin adjustment factor for chemical
effects (SONGS uses Trisodium Phosphate (TSP) as a post-LOCA pH buffering
agent, and pertinent debris loads are primarily mineral wool fibrous insulation,
making NRC's Integrated Chemical Effects Test (ICET) 2 generally applicable,
but the licensee stated that chemical effects values were subject to follow-on
sump screen vendor testing, and SCE evaluations and walkdowns).
Actions Taken
Chemical effect tests were completed by Alion Science and Technology, and
directly observed by the NRC, in Warrenville, Illinois on August 17 - 18, 2006.
Open items from the NRC review are to be addressed and supplemental
responses to the NRC are to be received no later than February 29, 2008, per
letter to NEI from NRC: Supplemental Licensee Responses to Generic
Letter 2004-02, "Potential Impact Of Debris Blockage On Emergency
Recirculation During Design Basis Accidents At Pressurized-Water Reactors,"
dated November 30, 2007.
6. Containment insulation configuration control to ensure the amounts and types of
insulation remain within acceptable debris loading design margins
Actions Taken
The licensee has removed microtherm insulation on four different piping
segments in containment. This insulation is to be replaced by reflective metal
insulation where appropriate. Mineral wool insulation on the steam generators is
-37- ENCLOSURE 2
to be replaced with RMI during the steam generator replacement activities in
2009. These actions have undergone NRC review and supplemental responses to
the NRC are to be received no later than February 29, 2008, per letter to NEI
from NRC: Supplemental Licensee Responses to Generic Letter 2004-02,
"Potential Impact Of Debris Blockage On Emergency Recirculation During Design
Basis Accidents At Pressurized-Water Reactors" dated November 30, 2007.
7. Replace sump screens at SONGS Unit 2 during Refueling Outage Cycle 15
Actions Taken
Work currently ongoing and expected to be completed by the end of the refueling
outage.
8. Removal of microporous insulation on piping to be completed coincident with
sump screen replacement.
Actions Taken
Work currently ongoing and expected to be completed by the end of the refueling
outage.
9. Modification fo steel grates at the entry to the bioshield to reduce the potential for
debris blockage and resultant hold-up of recirculating water to be completed
coincident with sump screen replacement.
Actions Taken
Work currently ongoing and expected to be completed by the end of the refueling
outage.
4OA6 Meetings, Including Exit
On November 9, 2007, the engineering inspectors presented the results of the
permanent plant modifications inspection and the evaluation of changes, tests, or
experiments inspection to Dr. R. Waldo and others who acknowledged the findings.
On November 30, 2007, the health physics inspectors presented inspection results to
Mr. J. Reilly and others who acknowledged the findings.
On December 3, 2007, the inspector discussed the inspection results of the licensed
operator annual requalification examination with Mr. B. Arbour, Training Supervisor. A
telephone exit was held with Mr. Arbour, on December 3, 2007. The licensee
acknowledged the findings presented in both the briefing and the final exit meeting.
On December 13, 2007, the inspectors presented the results of this inservice inspection
to J.T. Reilly, Vice-President Engineering and Technical Services, and other members of
licensee management. Licensee management acknowledged the inspection findings.
-38- ENCLOSURE 2
On December 21, 2007, and on February 13, 2008, the inspectors presented the
quarterly inspection results to Mr. R. Ridenoure and others who acknowledged the
findings.
The inspectors confirmed that proprietary information was not provided or examined
during the inspection.
4OA7 Licensee-Identified Violations
The following violation of very low significance (Green) was identified by the licensee and
is a violation of NRC requirements which meets the criteria of Section VI of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
- Licensee Technical Specification Section 5.5.1.1.a requires applicable procedures
recommended in Regulatory Guide 1.33. Revision 2, Appendix A, February 1978.
Section 7e of the Appendix requires procedures for access control and a radiation
work permit system. Radiation Exposure Permit A081997001/200117-8 requires
workers to wear radiological protective clothing for entry into contaminated areas,
such as shoe covers and gloves. Contrary to this requirement, there were three
examples of security officers entering contaminated areas without the required
protective clothing. The first example occurred on October 9, 2007, when two
security guards entered a posted contaminated area in Unit 3, Room 411 of the
penetrations building, without the required radiological protective clothing. The
second example occurred on November 12, 2007, when a security guard entered
a posted contaminated area in Unit 2, Room 209 without the required radiological
protective clothing. The third example occurred November 13, 2007, when a
security guard entered a posted contaminated area in Unit 2, Room 209 without
the required radiological protective clothing. In all three examples, the area
postings had changed and with inattention to detail, the officers entered the areas
without the required radiological protective clothing. This issue was entered into
the licensee's corrective action program (Action Requests 071000551,
071100759, and 071100760). This finding is of very low safety significance
because it did not involve: (1) ALARA planning and controls, (2) an overexposure,
(3) a substantial potential for overexposure, or (4) an impaired ability to assess
dose.
ATTACHMENT: SUPPLEMENTAL INFORMATION
-39- ENCLOSURE 2
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Axline, Technical Specialist, Nuclear Regulatory Affairs
D. Breig, Manager, Engineering Standards and Excellence
B. Corbett, Manager, Health Physics
J. Hirsch, Manager, Maintenance
K. Johnson, Manager, Design Engineering
R. Ridenoure, Vice President, Nuclear Generation
L. Kelly, Engineer, Nuclear Regulatory Affairs
C. McAndrews, Manager, Nuclear Oversight and Assessment
N. Quigley, Manager, Mechanical/Nuclear Maintenance Engineering
J. Reilly, Vice President, Engineering and Technical Services
A. Scherer, Manager, Nuclear Regulatory Affairs
R. St. Onge, Manager, Maintenance and Systems Engineering
T. Vogt, Manager, Special Projects
D. Wilcockson, Manager, Plant Operations
C. Williams, Manager, Compliance
T. Yackle, Manager, Operations
O. Flores, Manager, Chemistry
J. Morales, Manager, Projects
M. Cooper, Manager, Maintenance and Systems Engineering
S. Gardner, Nuclear Engineer, Nuclear Regulatory Affairs
A. Mahindrakar, Technical Specialist/Scientist, Maintenance and Systems Engineering
J. Valsvig, Technical Specialist/Scientist, Maintenance and Systems Engineering
M. McDevitt, Senior Nuclear Engineer, Engineering and Technical Services
P. Chang, Nuclear Engineer, Maintenance Engineering
A. Matheney, Senior Nuclear Engineer, Engineering and Technical Services
M. Wade, Westinghouse Representative
M. Short, Director Nuclear Oversight and Assessment
J. Todd, Manager, Nuclear Oversight and Regulatory Affairs
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000361; NOV Failure to Prevent Recurrence of Premature Tripping of
05000362/2007005-04 Square D Thermal Overloads (Section 4OA2.2)
A-1 ATTACHMENT
Opened and Closed
05000361; NCV Failure to Properly Implement Maintenance Rule
05000362/2007005-01 Requirements for Emergency Diesel Generators
(Section 1R12)05000362/2007005-02 NCV Failure to Implement Procedural Requirements for
Modificaitons in the Auxiliary Feedwater Steam Supply
Trench (Section 1R15)05000362/2007005-03 NCV Failure to Follow a Radiation Exposure Permit Requirement
(Section 2OS1)
Closed
None
Discussed
None
LIST OF DOCUMENTS REVIEWED
In addition to the documents called out in the inspection report, the following documents were
selected and reviewed by the inspectors to accomplish the objectives and scope of the
inspection and to support any findings:
Section 1R02: Evaluations of Changes, Tests, or Experiments
10 CFR 50.59 Evaluations
020701289-37 Auxiliary steam system radwaste condensate return Revision 0
line rad monitor flow valve change - Fix position of
Condensate Return Valve 2/3FV-7546 and remove
2/3FIC-7546
050801215-08 Change to the U3C14 Core Fuel Loading Pattern Revision 0
060101335-13 Reduction in the number of Dome Air Circulator Fans Revision 0
Credited for Containment Sprayed and Unsprayed
Region Mixing.
060401009-06 One-time change to the testing frequency for the High Revision 0
Pressure Turbine Stop and Control Valves
A-2 ATTACHMENT
060700747-13 Perform Calculation to evaluate the effects of air pocket Revision 0
on Engineered Safety Feature pump performance.
060700747-18 Perform Calculation to evaluate the effects of air pocket Revision 1
on Engineered Safety Feature pump performance.
060800698-13 Engineering design work by Bechtel to support steam
generator replacement - Remove one Containment Revision 0
Hydrogen Recombiner E146 for one cycle of operation
to facilitate Steam Generator Replacement
060800698-44 Change to UFSAR Section 8.1, paragraph 8.1.4.3.14.B Revision 0
10 CFR 50.59 Screenings
040400696-17 Add ECP vent line at AFW pump motor outboard 09/25/2007
bearing housing to eliminate oil leak
041100092-79 Need to Evaluate U-2 CCW Fisher Butterfly valve
concerning valve taper pin issue
050300070-05 Install Steam Trap in Auxiliary Steam Cross-tie header
050901044-40 Technical specification bases change to allow 11/01/2005
substituting B00X for battery B007 and B008 for
temporary battery outage
050901044-43 Technical specification bases change to allow 11/03/2005
substituting B00X for battery B007 and B008 for
temporary battery outage
050901044-61 Phase I of the Class 1E DC system upgrade 10/27/2005
050901044-61 Technical specification bases change to allow 12/16/2005
substituting B00X for battery B007 and B008 for
temporary battery outage (update)
050901044-82 Technical specification bases change to allow 03/20/2006
substituting B00X for battery B007 and B008 for
temporary battery outage
051000132-06 Update AOV Program Procedure to update valve IST
Procedure.
051200901-07 Installation of a flow orifice downstream of 2PCV4716 07/25/2006
060200607-18 Add DC shunts to batteries 2B007 and 2B009 for 06/08/2006
monitoring current
A-3 ATTACHMENT
060200607-51 Add DC shunts to batteries 2B007 and 2B009 for 08/02/2006
monitoring current - Addition of an 800 Amp, 100 mV
DC shunt at the positive polarity of battery B00X
060400474-04 Modify required actions in procedure SO23-5-1.7 to 04/10/2006
require MODE 3 entry for 1-3 inoperable MSSVs per
060400474-12 Modify required actions in procedure SO23-5-1.7 to 04/14/2006
require MODE 3 entry for 1-3 inoperable MSSVs per
060400474-32 Modify required actions in procedure SO23-5-1.7 to 07/27/2006
require MODE 3 entry for 1-3 inoperable MSSVs per
060400474-41 Modify required actions in procedure SO23-5-1.7 to 10/04/2006
require MODE 3 entry for 1-3 inoperable MSSVs per
060500070-14 ECP# 060500070-10: Replace 3P123 Feeder Breaker 05/052006
060500211-21 Replace vertical air tank S31319MV048 05/18/2006
060500211-38 Replace vertical air tank S31319MV048 06/16/2006
060500211-43 Replace vertical air tank S31319MV048 08/10/2006
060600089-84 Increase Thermal Overload size in breakers 2BY37, 09/18/2006
3BY37, 3BZ33
060800603-02 Replace existing R3, R4 potentiometers with a new
model in AVR for EDG. 01/24/2007
060800603-16 Replace existing R3, R4 potentiometers with a new 01/24/2007
060800603-29 Replace existing R3, R4 potentiometers with a new 03/07/2007
061001071-19 Use of new E4C-109 battery short circuit methodology 03/28/2007
061001842-82 Upsize Thermal Overloads to avoid Spurious Trips 11/15/2006
061100895-11 Material condition of Generator Neutral Grounding
Resistor is poor.
061101272-04 Install Lifting Eye Pad on beam to allow in-line lift
capability when changing out safety valve.
A-4 ATTACHMENT
070200876-05 Code upgrade installation for CENTS computer code 02/26/2007
version 06100
070200876-06 Code upgrade installation for TORCGEOM computer 03/26/2007
code version 1.0.5
070200876-07 Code upgrade installation for REX computer code 09/20/2007
version 2.1.6
070200876-08 Code upgrade installation for CORD computer code 09/20/2007
version 1.3.7
070700512-06 Lower the Set Point of the concerned instruments and
provide Control Room indication of actual pressure.
Calculations
E4C-112, CCN 72 Class 1E 480V MCC Protection Calculation Revision 1
E4C-112, Class 1E 480V MCC Protection Calculation Revision 1
ECN A46476
E4C-112,CCN 55 Class 1E 480V MCC Protection Calculation Revision 1
M-0012-039 ESF Pump Suction with Entrained Air after RAS Revision 0
(Recirculation Actuation Signal)
N-4061-001 Post-Loss Of Coolant Accident Summary of Low Revision 2
Populated Zones and Offsite Doses
N-4061-002 Post-Loss Of Coolant Accident Containment Leakage - Revision 1
Control Room and Offsite Doses
Action Requests
050901044 060200607 060400474 060800603 061001071
Section 1R04: Equipment Alignment
Procedures
SO23-3-2.6 Shutdown Cooling System Operation Revision 24
SD-SO23-780 Auxiliary Feedwater System Revision 10
SD-SO23-120 6.9 kV, 4.16 kV and 480 V Electrical Distribution Systems Revision 16
SO23-5-1.8.1 Shutdown Nuclear Safety Revision 17
A-5 ATTACHMENT
Drawings and Calculations
SD-SO23-740 Shutdown Cooling System Revision 17
40160A Auxiliary Feedwater System - No. 1305" Revision 43
40160B Auxiliary Feedwater Steam Supply System - No. 1301" Revision 21
40160C Auxiliary Feedwater System Hydraulic Valves 2HV-4714 Revision 7
& 4731 Control Fluid System No. 1305"
40160X Auxiliary Feedwater System No. 1305 and Auxiliary Revision 4
Feedwater Steam Supply System No. 1301"
Section 1R05: Fire Protection
Procedures
2-013 Unit 2, diesel generator pre-fire plans Revision 4
3-0345 Unit 3, diesel generator pre-fire plans Revision 4
2-007 Unit 2, Safety Equipment Building (-)15'6" Revision 3
elevation
UFHA 2/3-7.0-2SE Updated Fire Hazard Analysis May 2007
Action Requests
070901019 070901022
Section 1R08: Inservice Inspections
Procedures
Number Title Revision
SO23-XXVII-20.51 Visual Examination Procedure for Operability of Nuclear 2
Components and Supports and Conditions Relating to
Their Functional Adequacy
SO23-XXVII-20.48 Liquid Penetrant Examination 1
SO23-XXVII-30.13 Risk-Informed Ultrasonic Examination of Class 1 0
Austenitic Piping Welds
SO23-XXVII-30.6 Ultrasonic Examination of Austenitic Piping Welds 2
SO23-XXVII-30.9 Ultrasonic Examination of Dissimilar Metal Piping Welds 2
A-6 ATTACHMENT
PDI-UT-10 PDI Generic Procedure for the Ultrasonic Examination of C
9022 Reactor Coolant System Alloy 600 Material Management 5
Program
SO23-XXXIII-8.16 Reactor Coolant System Alloy 600 Inspection 5
SO23-3-2.34 Containment Access Control, Inspections and Airlocks 20
Operation
SO123-XXIV-10.1 Engineering Change Package 15
SO123-0-A4 Configuration Control 9
SO23-1-1.11.1 Plant Maintenance Procedure for Coating Service 6
Level 1 Application
SO23-XV-23.1.1 Containment Cleanliness/Loose Debris Inspection 1
SO23-V-8.17 Containment Coatings Assessment 1
QA-46 Qualification and Certification of NDE and Visual 0
Examination Personnel per ASME Section XI
WSI QAP 9.21 Liquid Penetrant Examination 1
SI-UT-126 Phased Array Ultrasonic Examination 3
T4EN51 Non-RCS Alloy 600 Boric Acid Leakage, Inspection and 1
Evaluation
T4EN52 RCS Alloy 600 Boric Acid Leakage, Inspection and 0
Evaluation
SO23-V-8.15 ISS2 Containment Boric Acid Leak Inspection 2
SO23-V-8.18 Reactor Coolant System (RCS) Leak Monitoring and 0
Investigation Guide
SO23-XV-85 Boric Acid Corrosion Control Program 1
SO23-XXXIII-8.16 Reactor Coolant System Alloy 600 Inspection 5
SO23-XXVII-3.51.9 IntraSpec UT Analysis Guidelines 5
SO23-XXVII-3.51.2 IntraSpec Eddy Current Imaging Procedure for Inspection 5
of Reactor Vessel Head Penetrations
SO23-XXVII-3.51.4 IntraSpec Ultrasonic Procedure for Inspection of Reactor 5
Vessel Head Penetrations, Time-of-Flight Ultrasonic,
Longitudinal Wave & Shear Wave
SO23-XXVII-3.51.3 IntraSpec Eddy Current Analysis Guidelines 6
A-7 ATTACHMENT
SO23-I-2.53 Containment Emergency Sump Inspection Surveillance 7
SO 123-I-11.1 Welding Filler material control 9
Corrective Action Documents
AR 070500261 AR 071101172 AR 071101173 AR 070500262
AR 070500263 AR 070500265 AR 071200384 AR 071200384
AR 060100998 AR 060101057 AR 060100961 AR 071200751
Calculations
Number Title Revision
SONG-10Q-301 Weld Overlay Sizing for Pressurizer Surge Nozzle 2
Drawings
Number Title Revision
SONG-10Q-02 Pressurizer Surge Nozzle Weld Overlay Design and Buffer 1
Layer, Shts 1 and 2
403974 Construction Drawing Surge, SONGS, Unit 2, Shts 1 and 2 0
S2-1203-ML-229 Letdown Heat Exchanger E-602 to Line 100: UA 12
2TV-0223, Sht 1
S2-1203-ML-498 Component Cooling Water, Sht 1 0
Examination Technique Specification Sheets (ETSS)
San Onofre Nuclear Generating Station Qualifying EPRI ETSSs
ETSS
ETSS #1 96004.1, 96005.2, 96008.1, 96012.1,
24013.1, 20511.1
ETSS #9 23514.1, .2, .3
ETSS #3 20510.1, 20511.1, 21409.1, 21410.1,
21998.1, 22401.1, 96703.1
ETSS #4 20510.1, 20511.1, 21409.1, 21410.1,
21998.1, 22401.1, 96703.1
A-8 ATTACHMENT
ETSS #5 96008.1, 96511.2
ETSS #6 96511.2, 99997.1
Welding Procedure Specifications and Corresponding Procedure Qualification Reports
WPS 08-08-T-001-Butter SS, Revision 0: PQRs 08-08-T-009, 08-08-TS-001, 8.8.6-OKG, and
08-08-TS-002
WPS 03-08-T-804-Bottom, Revision 0: PQRs A08202.3-3, 43-43-T-001, 03-03-T-803, and
A843256-52
WPS 1-GT-SM, Manual GTAW and/or SMAW of P-Number 1 CS, Revision 1: PQRs 51, 112,
and 153
Miscellaneous
Number Title Revision
RPA 02-0080 Quantification of Containment Latent Debris 1
ECP#04031974-74 Microtherm Insulation to RMI Change-out ECP; Unit 2
ECP# Microtherm Insulation to RMI Change-out ECP; Unit 3
04031974-58
ECP# Sump Screen Installation and Bioshield Gate
04031974-12 Modification ECP; Unit 2
ECP#04031974-11 Sump Screen Installation and Bioshield Gate
Modification ECP; Unit 3
Letter to NRC from SCE: NRC Generic Letter 2004-02 March 7, 2005
Response To NRC Request For Information San
Onofre Nuclear Generating Station Units 2 and 3
Letter to SCE from NRC: San Onofre Nuclear June 2, 2005
Generating Station Units 2 and 3-Request For
Additional Information (RAI) Related to Generic Letter 2004-02, "Potential Impact Of Debris Blockage On
Emergency Sump Recirculation At Pressurized-Water
Reactors" (TAC NOS. MC4714 and MC4715)
Letter to NRC from SCE: NRC Generic Letter 2004-02 July 5, 2005
Response To NRC Request For Additional Information
Letter to NRC from SCE: NRC Generic Letter 2004-02 September 1,
San Onofre Nuclear Generating Station Units 2 and 3 2005
A-9 ATTACHMENT
Letter to SCE from NRC: San Onofre Nuclear February 9,
Generating Station, Units 2 and 3, Request For 2006
Additional Information RE: Response to Generic Letter 2004-02, "Potential Impact Of Debris Blockage On
Emergency Sump Recirculation At Pressurized-Water
Reactors" (TAC NOS. MC4714 and MC4715)
Letter to PWR Owners Group from NRC: Alternative March 26,
Approach for Responding to the Nuclear Regulatory 2006
Commission Request for Additional Information Letter
RE: Generic Letter 2004-02 (TAC NOS. See
Enclosure)
Letter to PWR Owners Group from NRC: Alternative January 4,
Approach for Responding to the Nuclear Regulatory 2007
Commission Request for Additional Information Letter
RE: Generic Letter 2004-02 (TAC NOS. See
Enclosure)
San Onofre Nuclear Generating Station Units 2 and 3- May 16, 2007
Report on Results of Staff Audit of Corrective Actions
to Address Generic Letter 2004-02 (TAC NOS.
Letter to NEI from NRC: Plant-Specific Requests for November 8,
Extension of Time to Complete One or More 2007
Corrective Actions for Generic Letter 2004-02,
"Potential Impact Of Debris Blockage On Emergency
Recirculation During
Design Basis Accidents At Pressurized-Water
Reactors"
Letter to NEI from NRC: Supplemental Licensee November 30,
Responses to Generic Letter 2004-02, "Potential 2007
Impact Of Debris Blockage On Emergency
Recirculation During Design Basis Accidents At
Pressurized-Water Reactors"
ASNTCP-189-1995, ASNT Standard for Qualification
and Certification of Nondestructive Testing Personnel,
1995 Edition
Request For Relief ISI-3-25, Use of Structural Weld
Overlay and Associated Alternative Repair
Techniques
NRC Safety Evaluation for Request For Relief ISI-3-25 June 12, 2007
Weld Data Sheet, Pressurizer Surge Line Nozzle -
Weld ID DMW 02-005-031
A-10 ATTACHMENT
Welder Bead Logs for ER308L and Alloy 52M
deposition on Unit 2 Pressurizer Surge Nozzle
Steam Generator Degradation Assessment for the November 29,
Cycle 15 Refueling Outages in 2007 and 2008 2007
EA-03-009, Issuance of Order Establishing Interim February 11,
Inspection Requirements for Reactor Pressure Vessel 2003
Heads at Pressurized Water Reactors
EPRI Report 1010087, Materials Reliability Program:
Primary System Piping Butt Weld Inspection and
Evaluation Guidelines (MRP-139) August 2005
Certificate of Compliance dated 5/29/07 for ASME
Code Section II SFA5.9 Class ER 308/308L welding
material used on sacrificial layer on pressurizer surge
nozzle
Certificate of Compliance 06369301 for ASME Code
Section II, Part C SFA-5.14 Inconel 52M welding
material used to deposit weld overlay on pressurizer
surge nozzle
WSI Traveler No. 104532-TR-004 Pressurizer Surge 0
Nozzle Repair Work Steps
San Onofre Nuclear Generating Station Unit 3 Boric
Acid Corrosion Control Program (BACCP) Health
Report for Cycle 13: 12/29/2004 - 12/12/2006 May 8,
2007
Letter from T. G. San Onofre Nuclear Generating Station Units 2 and 3 June 12, 2007
Hiltz (NRC) to R. Re: Third 10-year Inservice Inspection Interval
M. Rosenblum Request ISI-3-25, Use of Structural Weld Overlays
(SCEC) and Associated Alternative Repair Techniques (TAC
NOS MD2579 and MD2580)
Guide 5 System Component Walkdown 1
Generic Letter Boric Acid Corrosion of Carbon Steel Pressure March 17,
88-05 Boundary Components in PWR Plants 1988
Information Notice Degradation of Reactor Coolant System Boundary January 5,86-109, Resulting from Boric Acid Corrosion 1995
Supplement 3
90022 Southern California Edison San Onofre Nuclear 5
Generating Station Units 2 and 3: Reactor Coolant
System Alloy 600 Material Management Program Plan
A-11 ATTACHMENT
Section 1R07A: Heat Sink Performance
SO23-I-8.94 Component Cooling Water Heat Exchanger Cleaning and Revision 8
Inspection
Action Requests
071000587 071200968
Maintenance Orders
06040726000
Section 1R11: Licensed Operator Requalification
Procedures
Lesson Plan Reactor Startup (Simulator) Revision 1
2RS767
Lesson Plan Plant Startup - Power Ascension from Mode 2 to 20% Revision 1
2RS768 Power (Simulator)
Action Requests
071000587
Maintenance Orders
06040726000
Section 1R12: Maintenance Effectiveness (Quarterly)
Procedures
SO23-3-3.23 Diesel Generator Monthly and Semi-annual Testing Revision 30
Action Requests
070300161
A-12 ATTACHMENT
Maintenance Orders
070300161-02 070300161-04
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
SO23-5-1.4 Plant Shutdown to Hot Standby Revision 13
SO23-5-1.3.1 Plant Startup from Hot Standby to Minimum Load Revision 26
Shutdown Nuclear Defense in Depth Planning Sheets Unit 3 Cycle 14 Fall Revision 0
Safety Program Midcycle Outage
SO23-5-1.8.1 Shutdown Nuclear Safety Revision 16
SO123-VIII-1 Recognition and Classification of Emergencies Revision 26
SO123-XX-6 Operator Work Around Program Revision 5
SO23-15-52.A Annunciator Panel 52A - FWCS/SBCS Revision 7
SO23-3-2.10 Main Steam Isolation Valve Operation Revision 16
SD-SO23-110 220 kV Switchyard System Revision 16
SSSPG-SO123- Assessment of Offsite Capabilities Following a Natural Revision 0
G-10 Disaster
Drawings and Calculations
SO23-507-6A-3-3 MSIV, FWIV, and FWBV Hydraulic Dump Valve Revision M
SO23-507-6A-5-3 MSIV, FWIV, and FWBV Hydraulic Dump Valve Revision M
40156FSO3 High Pressure Feedwater System Feedwater Isolation Revision 13
Valve 3HV4051 Electro-Hydraulic Actuation System
40141GSO3 Main Steam System Electro-Hydraulic Valve 3HV-8204 Revision 15
System
40141G Main Steam System Electro-Hydraulic Valve 2HV-8204 Revision 17
System
M3C14 DID #1 Barrier Map - Unit 3 Auxiliary Building (El. 50') Revision 0
M3C14 DID #1 Barrier Map - Unit 3 Safety Equipment Building (El. 15'- Revision 0
6" & 5'-3")
A-13 ATTACHMENT
M3C14 DID #3 Barrier Map - Train A Shutdown Cooling - Unit 3 Revision 0
Auxiliary Building (El. 50')
M3C14 DID #3 Barrier Map - Train A Shutdown Cooling - Unit 3 Safety Revision 0
Equipment Building (El. 15'-6" & 5'-3")
M3C14 DID #3 Barrier Map - Train B Shutdown Cooling - Unit 3 Revision 0
Auxiliary Building (El. 50')
M3C14 DID #3 Barrier Map - Train B Shutdown Cooling - Unit 3 Safety Revision 0
Equipment Building (El. 15'-6" & 5'-3")
UFSAR Fig. 8.2-1 One line Diagram - Switchyards Revision 16
Action Requests
071000609 070500815 071100595 071201499 071000250
Section 1R15: Operability Evaluations
Procedures
SO23-2-16 Operation of Waste Water systems Revision 20
SO23-20-4 Auxiliary Feedwater System Operation Revision 22
Vendor Spec Kanaline SR PVC Hose undated
Vendor Spec Prosser Standard-Line Submersible Dewatering Pumps June 2003
Series: 9-01000 & 9-01300"
Vendor Spec Prosser Standard-Line Submersible Dewatering Pumps March 2001
Series: 9-50000"
SO23-3-3.31.6 Main Feedwater System Valve Test Revision 7
SO23-3-3.31.4 Main Steam Valve Testing - Offline Revision 7
SO123-XV-5.1 Temporary Modification Control Revision 8
SO23-2-16 Use of Temporary Sump Pumps Revision 20
SO123-XV-52 Functionality Assessments and Operability Revision 7
Determinations
SO23-3-3.60.4 Saltwater Cooling Pump and Valve Testing Revision 9
Drawings and Calculations
40160A Auxiliary Feedwater System Revision 43
A-14 ATTACHMENT
40160B Auxiliary Feedwater Steam Supply System Revision 21
DCP 52 Plant design package to add trench eductor to TDAFW Revision 0
Action Requests
070500586 051200901 070500815 071100965 071000309 070500578
071000901
Section 1R17: Permanent Plant Modifications (71111.17A)
Engineering Change Packages
060400474-40 Modify required actions in procedure SO23-5-1.7 to Revision
require MODE 3 entry for 1-3 inoperable MSSVs per 09/27/2006
060800177-07 Replacement of Diesel Generator Temperature Switch Revision 00
per SEE 000036
061001379-84 Install CCW Bypass Flow around the Unit 3 Letdown Revision 00
Heat Exchanger
061001842-16 Replace Existing TOL for Breaker 2BZ17 Revision 00
061001842-46 Replace Existing TOL for Breaker 3BZ25
Drawings
S3-1023-ML-229, Letdown Heat Exchanger, Line 100: Valve 3TV-0223 Revision 15
Sht 1
S3-1203-ML-498, Component Cooling Water Line S3-1203-ML-498-4"-D- Revision 0
Sht 1 LL1 Sys 1203
S3-1203-ML-228, S3-1203-ML-228-8"-D-LL1, From Line 099 Valve 138 to Revision 13
Sht 1 Letdown Heat Exchanger
40123BS03 Reactor Coolant Chemical & Volume Control System Revision 29
No. 1208
Permanent Plant Modifications
020701289-37 Fix Position of Condensate Return Valve 2/3FV7546 01/15/2007
and Remove 2/3FIC-7546
040400696-17 Add ECP vent line at AFW pump motor outboard 09/25/2007
bearing housing to eliminate oil leak
A-15 ATTACHMENT
050901044-40 Technical specification bases change to allow 11/01/2005
substituting B00X for battery B007 and B008 for
temporary battery outage
051200901-07 Installation of a flow orifice downstream of 2PCV4716 07/25/2006
060500211-21 Replace vertical air tank S31319MV048 05/18/2006
060800603-29 Replace existing R3, R4 potentiometers with a new 03/07/2007
061101272-04 Install Pad Eye on beam over Safety Valve 3PSV0200 08/28/2007
Procedures
SO123-XV-44 10 CFR 50.59 and 72.48 Program Revision 8
Tech Spec Amendments
PCN 576 Request to revise Main Steam Safety Valve 11/07/2006
Requirements and Actions (T.S. 3.7.1)
Section 1R19: Postmaintenance Testing
Procedures
SO23-3-3.31.4 Main Steam Isolation Valve-Offline Testing Revision 7
SO23-3-3.31.6 Main Feedwater System Valve Test Revision 7
SO23-XXVII- Procedure for the Phased Array Ultrasonic Examination of Revision 1
33.14 Weld Overlaid Similar and Dissimilar Metal Welds
WSI 104125-TR- SONGS Pressurizer Surge Nozzle Repair Work Steps Revision 0
004
SO23-3-3.60.4 Saltwater Cooling Pump and Valve Testing Revision 9
SO23-3-3.31.10 Reactor Coolant Gas Vent System Test Revision 13
Miscellaneous
006-07 Repair/Replacement Plan for Weld Overlay Repair to Revision 0
Pressurizer Surge Nozzle
WPS -03-08-T-804- Weld Procedure Specification for Inconel to Stainless Revision 0
Bottom Steel
A-16 ATTACHMENT
WPS-08-08-T-001- Weld Procedure Specification for Stainless Steel Butter Revision 0
ButterSS
WPS-08-08-T-001-ButterSS Bead Log
WPS-03-08-T-804-Bottom Bead Log
Section 1R20: Refueling and Outage Activities
Procedures
SO23-5-1.4 Plant Shutdown to Hot Standby Revision 13
SO23-5-1.5 Plant Shutdown from Hot Standby to Cold Shutdown Revision 28
SO23-3-1.8 Draining the Reactor Coolant System Revision 26
SO23-5-1.8 Shutdown Operations (Mode 5 and 6) Revision 17
SO23-3-3.29 Determination of Reactor Shutdown Margin Revision 18
SO23-3-2.6 Shutdown Cooling System Operation Revision 24
SO23-I-3.5 Refueling Sequence Revision 14
SO23-5-1.3 Plant Startup from Cold Shutdown to Hot Standby Revision 30
SO23-5-1.7 Operating Instruction Revision 35
SO23-13-15 Loss Of Shutdown Cooling Revision 16
SO23-V-8.15 Containment Boric Acid Inspection Revision 2
M3C14 Defense In Depth Planning Sheets Revision 0
Action Requests
071200870 071200486
Section 1R22: Surveillance Testing
Procedures
SO23-3-3.30.8 Normal HVAC and Radiation Monitor Online Valve Test Revision 5
SO23-3-3.30.3 Component Cooling Water Seismic Makeup Valve Test Revision 11
SO23-3-3.30.2 Train A Saltwater Cooling Valve Test Revision 5
SO23-3-3.60.1 High Pressure Safety Injection Pump 2MP-018 Testing Revision 7
A-17 ATTACHMENT
SO23-3-3.60.3 Component Cooling Water Pump 2MP-024 Test Revision 8
SO23-3-3.60 Inservice Pump Testing Program Revision 8
Section 1R23: Temporary Plant Modifications
Procedures
ECP-07100097-3 Replace grounded pressurizer heater S31201ME616 Revision 0
with pressurizer heater S31201ME614"
Drawings and Calculations
32631 Elementary diagram reactor pressurizer backup heaters Revision 13
E124"
32632 Elementary diagram reactor pressurizer backup heaters Revision 27
E128"
32171 One line diagram pressurizer heaters distribution panels Revision 16
SO23-919-2- Heater element assembly Revision 4
D58
Section 1EP6 Drill Evaluation
Procedures
SO123-VIII-1 Emergency plan implementing procedures Revision 26
Emergency plan Drill 0704" October 3, 2007
SONGS Emergency Plan Revision 16
SO123-0-A7 Notification and Reporting of Significant Events Revision 5
Section 2OS1: Access Controls to Radiologically Significant Areas (71121.01)
Action Request Documents
061001562, 061100484, 061101431, 070700048, 070700545, 070701137, 070701389,
070800826, 071000512, 071000551, 071000551, 071100267, 071100759, 071100760
Audits, Self-Assessments, Observations, and Surveillance Reports
Health Physics Division Self-Assessment Reports for First, Second, and Third Quarter 2007
Leader Observation Program Records from May through November 2007
SCES-006-07
A-18 ATTACHMENT
Procedures
HP-I-2 Reactor Mode Change Checklist, Revision 14
SO123-VII-20 Health Physics Program, Revision 12
SO123-VII-20.6.1 Calculation of Dose from Skin Contamination, Revision 4
SO123-VII-20.7 Monitoring Internal Radiation Exposure, Revision 6
SO123-VII-20.9 Radiological Surveys, Revision 8
SO123-VII-20.9.6 Laboratory Analysis of Health Physics Air Samples, Revision 2
SO123-VII-20.11 Access Control Program, Revision 9
SO123-VII-20.11.1 Radiological Posting, Revision 8
Radiation Exposure Permits
A0707562000/200159, A0727070026, A0727070032/200101-12, A0819970001/200117-8
Miscellaneous
Selected Radiological Surveys during initial entry to Unit 2 Containment Refueling Outage
Unit 2 Shutdown Cooling Posting Plan
Section 2OS2: ALARA Planning and Controls (71121.02)
Action Request Documents
070400180, 070401109, 070401115, 070501042, 070600855, 070800568, 071101117,
071101118, 071101120, 071101121, 071101122, 071101124
Audits, Self-Assessments, Observations, and Surveillance Reports
Health Physics Division Self-Assessment Reports for First, Second, and Third Quarter 2007
Leader Observation Program Records from May through November 2007
SCES-006-07 and SOS-007-07
Procedures
HP-I-2 Reactor Mode Change Checklist, Revision 14
SO123-VII-20 Health Physics Program, Revision 11
SO123-VII-20.4 ALARA Program, Revision 4
SO123-VII-20.4.1 ALARA Design Change Reviews, Revision 4
SO123-VII-20.10 Radiological Work Planning and Controls, Revision 10
Radiation Exposure Permits
A0727070026, A1018940021
Miscellaneous
Reactor Coolant System Cobalt-58 Clean Up Curve for Unit 3 Midcycle 14
A-19 ATTACHMENT
Unit 2 Refueling Cycle 15 ALARA Daily Current Performance for November 26 through 29, 2007
Section 4OA1: Performance Indicator Verification (71151)
Procedures
SO23-XV-24 Quarterly NRC Performance Indicator (PI) Process, Revision 5
San Onofre Nuclear Generating Station; Station 2nd Quarter
Performace Report 2007
San Onofre Nuclear Generating Station; Station 3rd Quarter
Performace Report 2007
Miscellaneous
Quarterly Radiation Doses at the Site Boundary (Effluent Releases) for 2006 and 2007
Worker exposure records for radiological controlled area entries greater than 100 millirem
Section 4OA2: Identification and Resolution of Problems
Procedures
Policy Note 14 Human Performance Strategic Plan November 9,
2007
LIST OF ACRONYMS
ALARA as low as reasonably achievable
AR Action Request
AVR Automatic Voltage Regulator
BACC boric acid corrision control
CAP Corrective Action Program
CFR Code of Federal Regulations
EDG emergency diesel generator
EPRI Electric Power Research Institute
LER Licensee Event Report
NCV noncited violation
NDE nondestructive examination
SSC structure, system, and component
TS Technical Specification
UFHA Updated Fire Hazards Analysis
UFSAR Updated Final Safety Analysis Report
VUHP vessel upper head penetration
A-20 ATTACHMENT