ML081220464
ML081220464 | |
Person / Time | |
---|---|
Site: | Diablo Canyon |
Issue date: | 05/01/2008 |
From: | Vincent Gaddy NRC/RGN-IV/DRP |
To: | Conway J Pacific Gas & Electric Co |
References | |
FOIA/PA-2011-0221 IR-08-002 | |
Download: ML081220464 (11) | |
See also: IR 05000275/2008002
Text
UNITED STATES
NUC LE AR RE G UL AT O RY C O M M I S S I O N
R E GI ON I V
612 EAST LAMAR BLVD , SU I TE 400
AR LI N GTON , TEXAS 76011-4125
May 1, 2008
John T. Conway
Site Vice President and Chief Nuclear Officer
Pacific Gas and Electric Company
P.O. Box 3
Mail Code 104/6/601
Avila Beach, California 93424
SUBJECT: DIABLO CANYON POWER PLANT - NRC INTEGRATED INSPECTION
REPORT 05000275/2008002 AND 05000323/2008002
Dear Mr. Conway:
On March 31, 2008, the U.S. Nuclear Regulatory Commission completed an inspection at your
Diablo Canyon Power Plant, Units 1 and 2, facility. The enclosed integrated report documents
the inspection findings that were discussed on April 1, 2008, with Mr. James Becker and
members of your staff.
This 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.
Based on the results of this inspection, three NRC-identified findings of very low safety
significance (Green) were identified in this report. These findings involved violations of NRC
requirements. However, because of their very low risk significance and because they are
entered into your corrective action program, the NRC is treating these three findings as noncited
violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest
any NCV in this report, 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 the
Diablo Canyon Power Plant.
Pacific Gas and Electric Company -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document
system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Vince G. Gaddy, Chief
Project Branch B
Division of Reactor Projects
Dockets: 50-275
50-323
Licenses: DPR-80
Enclosure:
NRC Inspection Report 05000275/2008002
and 05000323/2008002
w/attachment: Supplemental Information
cc w/enclosure:
Sierra Club San Lucia Chapter
ATTN: Andrew Christie
P.O. Box 15755
San Luis Obispo, CA 93406
Nancy Culver
San Luis Obispo
Mothers for Peace
P.O. Box 164
Pismo Beach, CA 93448
Chairman
San Luis Obispo County
Board of Supervisors
1055 Monterey Street, Suite D430
San Luis Obispo, CA 93408
Truman Burns\Robert Kinosian
California Public Utilities Commission
505 Van Ness Ave., Rm. 4102
San Francisco, CA 94102
Pacific Gas and Electric Company -3-
Diablo Canyon Independent Safety Committee
Attn: Robert R. Wellington, Esq.
Legal Counsel
857 Cass Street, Suite D
Monterey, CA 93940
Director, Radiological Health Branch
State Department of Health Services
P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414
City Editor
The Tribune
3825 South Higuera Street
P.O. Box 112
San Luis Obispo, CA 93406-0112
James D. Boyd, Commissioner
California Energy Commission
1516 Ninth Street (MS 31)
Sacramento, CA 95814
James R. Becker, Site Vice President
Diablo Canyon Power Plant
P.O. Box 56
Avila Beach, CA 93424
Jennifer Tang
Field Representative
United States Senator Barbara Boxer
1700 Montgomery Street, Suite 240
San Francisco, CA 94111
Chief, Radiological Emergency Preparedness Section
National Preparedness Directorate
Technological Hazards Division
Department of Homeland Security
1111 Broadway, Suite 1200
Oakland, CA 94607-4052
Pacific Gas and Electric Company -4-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Michael.Peck@nrc.gov)
Branch Chief, DRP/B (Vincent.Gaddy@nrc.gov)
Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov)
Team Leader, DRP/TSS (Chuck.Paulk@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Only inspection reports to the following:
J. Adams, OEDO RIV Coordinator (John.Adams@nrc.gov)
P. Lougheed, OEDO RIV Coordinator (Patricia.Lougheed@nrc.gov)
ROPreports
DC Site Secretary (Agnes.Chan@nrc.gov)
SUNSI Review Completed: __yes___ ADAMS: Yes No Initials: __VGG_
- Publicly Available Non-Publicly Available Sensitive ; Non-Sensitive
R:\_REACTORS\_DC\2008\DC2008-02RP-MSP.wpd ML 0181220464
RIV:SRI:DRP/B C:DRS/OB C:DRS/PSB C:DRS/EB2
MSPeck RLantz MShannon LSmith
/RA/ e-mailed /RA/ /RA/ /RA/
4/29/08 4/10/08 4/14/08 4/14/08
C:DRS/EB1 C:DRP/B
RBywater VGaddy
/RA/ /RA/
4/11/08 4/30/08
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Dockets: 50-275, 50-323
Report: 05000275/2008002
Licensee: Pacific Gas and Electric Company
Facility: Diablo Canyon Power Plant, Units 1 and 2
Location: 71/2 miles NW of Avila Beach
Avila Beach, California
Dates: January 1 through March 31, 2008
Inspectors: M. Peck, Senior Resident Inspector
M. Brown, Resident Inspector
Lee Ellershaw, Senior Reactor Inspector, Region IV
C. Graves, Health Physicist
J. Groom, Resident Inspector, Callaway Plant
B. Henderson, Reactor Inspector, Region IV
Jared Nadel, Reactor Inspector, Region IV
J. Melfi, Resident Inspector, Palo Verde
A. Sanchez, Senior Resident Inspector, Arkansas Nuclear One
Approved By: V. Gaddy, Chief, Projects Branch B
Division of Reactor Projects
-1- Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS .....................................................................................................- 3 -
REPORT DETAILS.................................................................................................................- 6 -
REACTOR SAFETY ...............................................................................................................- 6 -
1R01 Adverse Weather.............................................................................................- 6 -
1R04 Equipment Alignments.....................................................................................- 7 -
1R05 Fire Protection .................................................................................................- 8 -
1R11 Licensed Operator Requalification.................................................................- 14 -
1R12 Maintenance Effectiveness............................................................................- 15 -
1R13 Maintenance Risk Assessments and Emergent Work Control .......................- 17 -
1R15 Operability Evaluations..................................................................................- 18 -
1R18 Plant Modifications ........................................................................................- 19 -
1R19 Postmaintenance Testing ..............................................................................- 21 -
1R20 Refueling and Other Outage Activities ...........................................................- 22 -
1R22 Surveillance Testing ......................................................................................- 23 -
1EP6 Emergency Preparedness Evaluation............................................................- 27 -
RADIATION SAFETY ...........................................................................................................- 27 -
2OS1 Access Control To Radiologically Significant Areas .......................................- 28 -
2OS2 ALARA Planning and Controls.......................................................................- 30 -
OTHER ACTIVITIES ............................................................................................................- 32 -
4OA2 Identification and Resolution of Problems......................................................- 34 -
4OA6 Meetings, Including Exit.................................................................................- 41 -
ATTACHMENT: SUPPLEMENTAL INFORMATION .............................................................- 42 -
Key Points of Contact ..1
List of Items Opened, Closed, and Discussed ............................................................................ 1
List of Documents Reviewed ...................................................................................................... 2
-2- Enclosure
SUMMARY OF FINDINGS
IR 05000275/2008002, 05000323/2008002; 1/1 - 3/31/08; Diablo Canyon Power Plant, Units 1
and 2; Fire Protection, Maintenance Effectiveness, and Occupational Radiation Safety.
This report covered a 13-week period of inspection by resident inspectors and announced
inspections in radiation protection. Three NRC-identified, Green, noncited violations were
identified. The significance of most findings is indicated by their color (Green, White, Yellow, or
Red) using Inspection Manual Chapter 0609 Significance Determination Process. Findings for
which the Significance Determination Process does not apply may be Green or be assigned a
severity level after NRC management review. The NRCs program for overseeing the safe
operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. On February 17, 2008, the inspectors identified a noncited violation of
Technical Specification 5.4.1.d, Fire Protection Program, after Pacific Gas and
Electric failed to maintain the integrity of an auxiliary building fire door. The
inspectors identified that the latching mechanism on Fire Door 348 was degraded
and not engaged. The unlatched fire door resulted in a reduction in fire
confinement capability. The door was required to provide a 11/2-hour fire barrier
between two plant fire areas. The licensee had several prior opportunities to
identify the degraded fire door. Security and operations personnel passed
through the affected fire area several times each day.
This finding is greater than minor because the degraded fire barrier affected the
mitigating systems cornerstone external factors attribute objective to prevent
undesirable consequences due to fire. Using Manual Chapter 0609, Appendix F,
Fire Protection Significance Determination Process, the inspectors determined
this finding is within the fire confinement category and the fire barrier was
moderately degraded because the door latch was not functional. The inspectors
concluded that this finding is of very low safety significance because a non-
degraded automatic full area water based fire suppression system was in place
in the exposing fire area. This finding was entered into the corrective action
program as Action Request A0719774. This finding has a crosscutting aspect in
the area of problem identification and resolution associated with the corrective
action program component because plant personnel did not maintain a low
threshold for identifying issues. P.1(a) (Section 1R05)
- Green. The inspectors identified a noncited violation of 10 CFR 50.65(a)(2), after
Pacific Gas and Electric Company failed to effectively control performance
monitoring of the Unit 2 containment atmosphere particulate radiation monitor
through appropriate preventive maintenance. Eight functional failures of the
radiation monitor occurred between November 2006 and January 2008. The
-3- Enclosure
licensee did not categorize any of these failures as Maintenance Rule functional
failures.
This finding is greater than minor because it is associated with the mitigating
systems cornerstone attribute of equipment performance and it affects the
cornerstone objective to ensure the availability, reliability, and capability of the
systems that respond to initiating events to prevent undesirable consequences.
The inspectors evaluated the significance of this finding using Inspection Manual
Chapter 0609, Significance Determination Process, Phase 1, Appendix A. The
inspectors determined that this finding was of very low safety significance
because this is not a design or qualification deficiency, does not represent a loss
of a system safety function or safety function of a single train, and does not
screen as potentially risk significant due to external events. The inspectors also
determined that this finding has a crosscutting aspect in the area of human
performance associated with the work practices component because engineering
staff failed to follow the November 2006 revision to the licensee maintenance rule
procedure that would have required each failure to be counted as a maintenance
rule functional failure. Engineering staff incorrectly concluded that the revision
was not applicable to the radiation monitors and therefore did not implement the
change H.4(b) (Section 1R12).
Cornerstone: Occupational Radiation Safety
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1 for failure to follow a licensee procedure. Specifically, while
touring the Unit 2 spent fuel pool on February 13, 2008, the inspectors observed
workers performing fuel inspections on the fuel bridge. The inspectors noted that
the physical location of a continuous air monitor, an AMS-4, was in the southeast
corner of the floor. Ventilation flow in this area was north to south with negative
ventilation centered on the spent fuel pool. Section 2.2 of Procedure RCP D-430
states, in part, the purpose of the continuous air monitors was to alert personnel
to changes in radiological conditions and that locations are selected based on
their potential as contributors to airborne activity. The location of the continuous
air monitor was not appropriate to alert the workers of changing radiological
conditions. During review of this occurrence, the inspectors were made aware of
a similar issue. Specifically, Action Request A0666110 was opened on
May 3, 2006, to evaluate the adequacy of AMS-4 placement in the fuel building
during fuel moves. This action request was currently open with a resolution date
of December 15, 2008.
This finding is greater than minor because it is associated with the occupational
radiation safety program and process attribute and affected the cornerstone
objective, in that the failure to monitor for radioactive material in the air had the
potential to increase personnel dose. This occurrence involves workers
unplanned, unintended or potential for such dose; therefore, this finding was
evaluated using the occupational radiation safety significance determination
process. The inspectors determined that this finding was of very low safety
significance because it did not involve: (1) an as low as is reasonably achievable
planning or work control issue; (2) an overexposure; (3) a substantial potential for
overexposure; or (4) an impaired ability to assess dose. This finding also has a
crosscutting aspect in the area of problem identification and resolution, corrective
-4- Enclosure
action component, because the licensee failed to take timely corrective actions to
address safety issues. P.1(d) (Section 2OS1)
-5- Enclosure
REPORT DETAILS
Summary of Plant Status
Pacific Gas and Electric Company (PG&E) was operating Diablo Canyon Unit 1 and Unit 2 at
full power at the beginning of the inspection period. On January 5, 2008, the licensee reduced
both units to 55 percent power in response to condenser fouling resulting from high sea swells.
On January 6, plant operators returned both units to full power and subsequently reduced Unit 1
to 50 percent power following high circulating water pump bearing temperature. On January 7,
plant operators returned Unit 1 to full power after repairing a failed bearing temperature sensor.
PG&E shut down Unit 2 on February 3 for refueling and steam generator replacement. Unit 2
remained down for the remainder of the inspection period.
1. REACTOR SAFETY
Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness
1R01 Adverse Weather (71111.01)
.1 Winter Seasonal Readiness Preparations
a. Inspection Scope
The inspectors conducted a review of PG&E preparations for seasonal susceptibilities
involving high wind and heavy rains on January 3, 2008. The inspectors completed this
review to verify that the plants design features and procedures were sufficient to protect
mitigating systems from the effects of adverse weather. Documentation for selected risk
significant systems was reviewed to ensure that these systems would remain functional
when challenged by inclement weather. During the inspection, the inspectors focused
on plant specific design features and the licensees procedures used to mitigate or
respond to adverse weather conditions. Additionally, the inspectors reviewed the Final
Safety Analysis Report (FSAR) and performance requirements for systems selected for
inspection and verified that operator actions were appropriate as specified by plant
specific procedures. The inspectors also reviewed corrective action program items to
verify that the licensee was identifying adverse weather issues at an appropriate
threshold and entering them into their corrective action program in accordance with
station corrective action procedures. Specific documents reviewed during this inspection
are listed in the attachment.
This inspection constitutes one seasonal readiness preparations sample as defined in
Inspection Procedure 71111.01-05.
b. Findings
No findings of significance were identified.
-6- Enclosure
.2 Readiness for Bio-fouling Concerns
a. Inspection Scope
During the week of January 1, 2008, the inspectors observed licensee activities
associated with expected condenser and ultimate heat sink heat exchanger fouling
resulting from high sea swells. The inspectors observed pre-job briefings, pre-shift
briefings and control room briefings to determine whether the briefings met licensee
standards. The inspectors reviewed Procedure OP O-28, Intake Management,
Revision 10, to verify reactor power reduction prerequisites were met. Finally, during the
remainder of the inspection period, the inspectors periodically reviewed licensee
activities and data collection as specified by licensee procedures to determine whether
increasing condenser circulation water pressure was properly monitored. The inspectors
also reviewed corrective action program items to verify that the licensee was identifying
adverse weather issues at an appropriate threshold and entering them into their
corrective action program in accordance with station corrective action procedures.
This inspection constitutes one readiness for imminent adverse weather condition
sample as defined in Inspection Procedure 71111.01-05.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignments (71111.04)
.1 Quarterly Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- Unit 2, Spent fuel pool cooling system during core offload, February 14, 2008
- Unit 1, Component cooling water pump and heat Exchanger 1-1, March 21, 2008
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, FSAR, Technical Specification requirements, Administrative Technical
Specifications, outstanding work orders, condition reports, and the impact of ongoing
work activities on redundant trains of equipment in order to identify conditions that could
have rendered the systems incapable of performing their intended functions. The
inspectors also walked down accessible portions of the systems to verify system
components and support equipment were aligned correctly and were operable. The
inspectors examined the material condition of the components and observed operating
parameters of equipment to verify that there were no obvious deficiencies. The
inspectors also verified that the licensee had properly identified and resolved equipment
alignment problems that could cause initiating events or impact the capability of
-7- Enclosure
mitigating systems or barriers and entered them into the corrective action program with
the appropriate significance characterization. Specific documents reviewed during this
inspection are listed in the attachment.
These activities constitute two partial system walkdown samples as defined by
Inspection Procedure 71111.04-05.
b. Findings
No findings of significance were identified.
.2 Semi-Annual Complete System Walkdown
a. Inspection Scope
On January 22, 2008, the inspectors performed a complete system alignment inspection
of the Unit 2 high head injection system to verify the functional capability of the system.
This system was selected because it was considered both safety-significant and risk-
significant in the licensees probabilistic risk assessment. The inspectors walked down
the system to review mechanical and electrical equipment alignment, electrical power
availability, system pressure and temperature indications, as appropriate, component
labeling, component lubrication, component and equipment cooling, hangers and
supports, operability of support systems, and to ensure that ancillary equipment or
debris did not interfere with equipment operation. A review of a sample of past and
outstanding work orders was performed to determine whether any deficiencies
significantly affected the system function. In addition, the inspectors reviewed the
corrective action program database to ensure that system equipment alignment
problems were being identified and appropriately resolved. The documents used for the
walkdown and issue review are listed in the attachment.
These activities constitute one complete system walkdown sample as defined by
Inspection Procedure 71111.04-05.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
Quarterly Inspection
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk significant
plant areas:
- Fire Area 8-A, Unit 1, Computer room, January 15, 2008
- Fire Area 8-D, Unit 2, Computer room, January 15, 2008
- Fire Area 14-D, Unit 1, 140' Turbine deck, January 15, 2008
- Fire Area 19-D, Unit 2, 140' Turbine deck, January 15, 2008
-8- Enclosure
- Fire Area 22-C, Unit 2, Diesel generator corridor, January 29, 2008
- Fire Area 24-D, Unit 2, Excitation switchgear room, January 29, 2008
- Fire Area 3-X, Auxiliary building 100 foot level, February 10, 2008
- Fire Area 3-T-2, Unit 2, Motor-driven auxiliary feed pump, February 10, 2008
- Fire Area 3-BB, Unit 1, Containment penetration room, February 17, 2008
The inspectors reviewed areas to assess if the licensee had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant, effectively maintained fire detection and suppression capability, maintained
passive fire protection features in good material condition, and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire
protection equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to impact equipment which could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event. Using
the documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed, that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute nine quarterly fire protection inspection samples as defined by
Inspection Procedure 71111.05-05.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1.d, Fire Protection Program, after PG&E failed to maintain the
integrity of an auxiliary building fire door.
Description. On February 17, 2008, the inspectors identified a noncited violation of
Technical Specification 5.4.1.d, Fire Protection Program, after PG&E failed to maintain
the integrity of an auxiliary building fire door. The inspectors identified that the latching
mechanism on Fire Door 348 was not engaged. The degraded door latch resulted in a
reduction in the confinement capability of the fire barrier. The door was required to
provide a 11/2-hour fire barrier between Fire Areas 3-BB and 3-AA. The licensee had
several opportunities to identify the degraded fire door. Security personnel passed into
the affected fire area at least three times each day and operations personnel passed
through the fire area at least once each shift. Procedure OM8.ID2, Fire System
Impairment, Revision 13, required plant personnel to notify the operations shift foreman
and ensure an action request is generated after discovering a fire protection system
impairment. The inspectors verified that licensee personnel had neither communicated
to the operations shift foreman nor had an action request been generated for the
degraded fire door. The inspectors previously identified that the latches on Fire
Doors 258-2, 174-A, and 350-2 were degraded on February 10, 2008. The failure of
licensee personnel to identify these degraded fire doors was entered into the corrective
action program as Action Requests A0718944, A0718946, and A0718947.
-9- Enclosure
Analysis. The failure of PG&E to maintain the integrity of Fire Door 348 is a
performance deficiency. This finding is more than minor because the degraded fire
barrier affected the mitigating systems cornerstone external factors attribute objective to
prevent undesirable consequences due to fire. The inspectors used the Inspection
Manual Chapter 0609, Appendix F, Fire Protection Significance Determination
Process, to analyze this finding. The inspectors determined this finding was a fire
confinement category and that the fire barrier was moderately degraded because the
door latch was not functional. The inspectors concluded that this finding is of very low
safety significance because a non-degraded automatic full area water based fire
suppression system was in placed in the exposing fire area. This finding has a
crosscutting aspect in the area of problem identification and resolution associated with
the corrective action program component because plant personnel did not maintain a
low threshold for identifying issues P.1(a).
Enforcement. Technical Specification 5.4.1.d required that PG&E implement a Fire
Protection Program. The Fire Protection Program requirements, as described by FSAR
Appendix 9.5a, Fire Hazards Analysis, required that Fire Door 348 be maintained as a
fire area boundary. Contrary to the above, on February 17, 2008, the inspectors
identified that plant personnel failed to maintain Fire Door 348 as a fire boundary.
Because this finding is of very low safety significance and was entered into the
corrective action program as Action Request A0719774, this violation is being treated as
a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000275/2008002-01, Failure to Identify a Degraded Fire Barrier.
1R08 Inservice Inspection Activities (71111.08)
02.01 Inspection Activities Other Than Steam Generator Tube Inspection, PWR 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 pressure boundary. Also review one or two examinations with recordable
indications that have been accepted by the licensee for continued service. In addition
the inspectors also reviewed welding and NDE activities associated with the steam
generator replacement to fulfill the inspection requirements of Inspection
Procedure 50001, Steam Generator Replacement Inspection.
The inspectors directly observed the following nondestructive examinations:
System Identification Exam Type Result
Pressurizer Surge WIB-438-439 O.L. PT No Relevant
Indications
Pressurizer Spray WIB-345-346 O.L. UT No Relevant
Indications
- 10 - Enclosure
Main Steam 2-K15-228-28V VT-3 No Relevant
Hanger 2020-1V Indications
Main Steam 2-K15-228-28 MT No Relevant
Attachment 2020-1V Indications
(6 lugs)
Reactor Pressure Vent line ET No relevant
Vessel Upper Head indications
Chemical & Volume Pipe Weld 2033-1 UT No relevant
Control System indications
The inspectors reviewed records for the following nondestructive examinations:
System Identification Exam Type Result
Pressurizer Safety B WIB-422A-423 O.L. UT and PT No Relevant
Nozzle (WOR) indications
Pressurizer Spray WIB-345-346 O.L. PT No Relevant
Line Nozzle Indications
Reactor Pressure CRDMs 6,10, 14, Bare Metal Visual No Relevant
Vessel Upper Head 15, 18, 22, 23, 30, Remote, robotic indications
31, 32, 37, 38, 42, camera
43, 51, 54, 55, 56,
62
Reactor Pressure CRDM 19,33,39,58 UT, ET No relevant
Vessel Upper Head indications
Steam Generator FW-4 and FW-4R1 RT No Relevant
2-4 Feedwater Line Indications
Reactor Coolant WIB-RC-2-1 (SE) UT No Relevant
System Hot Leg Dissimilar Metal Indications
Outlet Nozzle Weld
Reactor Coolant WIB-RC-3-16 (SE) UT No Relevant
System Cold Leg Dissimilar Metal Indications
Inlet Nozzle Weld
During the review and observation of each examination, the inspectors verified that
activities were performed in accordance with American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code requirements and applicable
procedures. Indications were compared with previous examinations and dispositioned in
accordance with ASME Code and approved procedures. The qualifications of all
nondestructive examination technicians performing the inspections were verified to be
current.
- 11 - Enclosure
No NDE examinations with relevant indications were accepted by the licensee for
continued service.
Three examples of welding on the reactor coolant system pressure boundary and one
example of welding on the chemical and volume control system were examined through
direct observation and/or record review as follows:
System Component/Weld Identification
Chemical & Volume Control Charging Pump 2-2, discharge line pipe-to-fitting
System Weld 7
Reactor Coolant System Pressurizer Safety Valve B Nozzle WOL
Reactor Coolant System Pressurizer Spray Line/WIB-345-346 WOL
Reactor Coolant System Pressurizer Surge Line/WIB-438-439 WOL
Welding procedures and nondestructive examination of the welding repair conformed to
ASME Code requirements and licensee requirements.
The inspectors verified, by review, that the welding procedure specifications and the
welders had been properly qualified in accordance with ASME Code,Section IX,
requirements. The inspectors also verified, through observation and record review, that
essential variables for the gas tungsten arc welding process (machine and manual) and
the shielded metal arc welding process were identified, recorded in the procedure
qualification record, and formed the bases for qualification of the welding procedure
specifications.
The inspectors completed one sample under Section 02.01.
b. Findings
No findings of significance were identified.
02.02 Vessel Upper Head Penetration (VUHP) Inspection Activities
a. Inspection Scope
The licensee performed NDE of 100 percent of reactor vessel upper head penetrations.
The inspector directly observed a sample consisting of the examinations listed below:
System Component ID Examination Method Result
VUHP Vent Line ET No relevant indications
- 12 - Enclosure
The inspectors reviewed the following sample of examinations in which indications were
observed, evaluated and determined not to be relevant indications using stored
electronic data or review of printed records:
System Component ID Examination Method Result
VUHP CRDM 19,33,39,58 UT,ET No relevant indications
The NDE inspections were performed in accordance with the requirements of NRC
Order EA-03-009. Qualifications of NDE personnel were reviewed and verified to be
current.
The inspectors completed one sample under Section 02.02.
b. Findings
No findings of significance were identified.
02.03 Boric Acid Corrosion Control Inspection Activities
a. Inspection Scope
The inspectors observed a sample of boric acid corrosion control inspection activities
and verified that visual inspections emphasized locations where boric acid leaks can
cause degradation of safety significant components.
The inspectors also reviewed one instance where boric acid deposits were found on
reactor coolant system piping components:
Component Number Description Action Request
CVCS-2-8148 Boric acid deposits on 1 of 6 body-to-bonnet A070014
studs and nuts
The condition of all the components was appropriately entered into the licensee=s
corrective action program, and corrective actions taken were consistent with ASME code
requirements. An engineering evaluation was conducted and the affected nut and stud
were removed and examined. The bolting material is stainless steel and is not
susceptible to corrosion from boric acid solution. No evidence of wastage, corrosion or
damage was found, and the bolting was returned to service.
The inspectors completed one sample under Section 02.03.
b. Findings
No findings of significance were identified.
- 13 - Enclosure
02.04 Steam Generator Tube Inspection Activities
a. Inspection Scope
Unit 2 steam generators were replaced during this outage and steam generator tubes
were not inspected.
b. Findings
No findings of significance were identified.
02.05 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 inspectors reviewed 17 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.
1R11 Licensed Operator Requalification (71111.11)
a. Inspection Scope
On January 3, 2008, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification training to verify that operator
performance was adequate, evaluators were identifying and documenting crew
performance problems, and training was being conducted in accordance with licensee
procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- 14 - Enclosure
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate Technical Specification actions and
Emergency Plan actions and notifications.
The crews performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements. Documents reviewed
by the inspectors included Instructor Lesson Guide R075S2, 2007 Continuing Operator
Training, dated November 29, 2007.
This inspection constitutes one quarterly licensed operator requalification program
sample as defined in Inspection Procedure 71111.11.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
Routine Quarterly Evaluations 71111.12Q
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems:
- Unit 2, Containment atmosphere particulate Radioactivity Monitor RM-11 paper
drive assembly failures, January 22, 2008
- Unit 2, Component cooling water Valve CCW-2-695 local leak rate test failure,
February 27, 2008
The inspectors reviewed events where ineffective equipment maintenance has resulted
in valid or invalid automatic actuations of engineered safeguards systems and
independently verified the licensee's actions to address system performance or condition
problems in terms of the following:
- implementing appropriate work practices;
- identifying and addressing common cause failures;
- scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
- characterizing system reliability issues for performance;
- charging unavailability for performance;
- 15 - Enclosure
- trending key parameters for condition monitoring;
- ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and
- verifying appropriate performance criteria for structures, systems, and
components functions classified as (a)(2) or appropriate and adequate goals and
corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization. Specific documents reviewed during this inspection are
listed in the attachment.
This inspection constitutes two quarterly maintenance effectiveness samples as defined
in Inspection Procedure 71111.12-05.
b. Findings
Introduction. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(2),
after PG&E failed to effectively monitor performance of the Unit 2, containment
atmosphere particulate radioactivity monitor through appropriate preventive
maintenance.
Description. Eight functional failures of the Unit 2, containment atmosphere particulate
radiation monitor occurred between November 2006 and January 2008. Each failure
required entry into Technical Specification Action 3.4.15, Reactor Coolant System
Leakage Detection Instrumentation. The licensee did not consider any of the radiation
monitor failures as Maintenance Rule functional failures. Beginning in November 2006,
Procedure MA1.ID17, Maintenance Rule Monitoring Program, required that the
licensee declare a maintenance rule functional failure for failed scoped components that
also required an unplanned entry into a Technical Specification action.
Technical Specification bases for 3.4.15 stated that reactor coolant leakage detection
systems met Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage
Detection Systems. Regulatory Guide 1.45 stated that the particulate radiation monitor
provides a separate and diverse method for detection, classification, and location of
reactor leakage throughout the plant operating cycle. The inspectors concluded that the
numerous failures of the particulate radiation monitor should have been evaluated
against the licensees performance criteria and resulted in placement of system into
Maintenance Rule (a)(1) status.
Analysis. The failure of PG&E to effectively control performance monitoring of the
Unit 2, containment particulate radioactivity monitor in accordance with
10 CFR 50.65(a)(2) was a performance deficiency. This finding is more than minor
because it is associated with the equipment performance attribute of the mitigating
systems cornerstone and affected the cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. The inspectors evaluated the significance of this finding
using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1,
Appendix A. The inspectors determined that this finding was of very low safety
- 16 - Enclosure
significance (Green) because this finding is not a design or qualification deficiency, does
not represent a loss of a system safety function or safety function of a single train, and
does not screen as potentially risk significant due to external events. The inspectors
also determined that this finding has a crosscutting aspect in the area of human
performance associated with the work practices component because engineering staff
failed to follow the November 2006 revision to the licensee maintenance rule procedure
that would have required each failure to be counted as a maintenance rule functional
failure. Engineering staff inaccurately concluded that the revision was not applicable to
the radiation monitors and therefore did not implement the change H.4(b).
Enforcement. 10 CFR 50.65(a)(1), requires, in part, that the holders of an operating
license shall monitor the performance or condition of structures, systems, and
components within the scope of the rule as defined by 10 CFR 50.65(b), against
licensee-established goals, in a manner sufficient to provide reasonable assurance that
structures, systems, and components are capable of fulfilling their functions.
Paragraph (a)(2) of 10 CFR 50.65 states, in part, that monitoring as specified in
10 CFR 50.65(a)(1) is not required where it has been demonstrated that the
performance or condition of an structures, systems, and components is effectively
controlled through the performance of appropriate preventive maintenance such that the
systems, structures, and components remains capable of performing its intended
function.
Contrary to the above, PG&E did not demonstrate that the performance or condition of
the Unit 2 containment atmosphere particulate radioactivity monitor had been effectively
controlled through the performance of appropriate preventive maintenance and did not
monitor against licensee-established goals. Specifically, repetitive failures associated
with Unit 2 containment atmosphere particulate radioactivity monitor from
November 2006 to January 2008 demonstrated that the Unit 2 containment atmosphere
particulate radioactivity monitor performance was not being effectively controlled per
10 CFR 50.65(a)(2). Because this issue is of very low safety significance (Green) and is
entered into PG&Es corrective action program as Action Request A0717009, this
violation is being treated as a noncited violation consistent with Section VI.A.1 of the
NRC Enforcement Policy: NCV 05000275,05000323/2008002-02, Failure to
Demonstrate a Containment Atmosphere Particulate Radiation Monitor Performance
was Effectively Controlled.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the
maintenance and emergent work activities affecting risk-significant and safety-related
equipment listed below to verify that the appropriate risk assessments were performed
prior to removing equipment for work:
- Technical Specification Sheet T0061921, Unit 1, Residual Heat Removal
Pump 1-2 planned maintenance, January 9, 2008
- TSS T0062026, Unit 2, Trip risk during scaffolding construction,
January 15, 2008
- 17 - Enclosure
- TSS T0062095, Unit 1, Surveillance testing of excore instrumentation,
January 30, 2008
- TSS T0062365, Unit 1, Failure of generator seal oil pump, February 26, 2008
- TSS T0062438, Unit 1, Phase duct cooler out of service for corrective
maintenance, March 11, 2008
- TSS T0062492, Unit 1, Removal of Vital Battery Charger 1-1 for planned
maintenance, March 24, 2008
These activities were selected based on their potential risk significance relative to the
reactor safety cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed Technical
Specification requirements and walked down portions of redundant safety systems,
when applicable, to verify risk analysis assumptions were valid and applicable
requirements were met.
These activities constituted six samples as defined by Inspection
Procedure 71111.13-05.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- Action Request A0687787, Units 1 and 2, Degraded seismic qualification of the
fuel handling building, January 8, 2008
- Action Request A0714564, Unit 2, Degraded auxiliary building supply Fan S-46,
January 16, 2008
- Action Request A0717989, Unit 2, High reactor coolant system radioiodine due to
failed fuel, January 17, 2008
- Action Request A0717034, Unit 2, High motor current on containment fan cooling
units, January 28, 2008
- 18 - Enclosure
- Action Request A0717677, Unit 1, Component cooling water Pump 1-2 motor oil
leak, January 30, 2008
- Action Request A0720656, Units 1 and 2, Cyclic fatigue of the emergency diesel
generator fuel lines, March 8, 2008
- Action Request A0722963, Units 1 and 2, Emergency diesel generator
tachometer failed to reset during power transfer, March 10, 2008
- Action Request A0721019, Unit 1, Emergency Diesel Generator 1-01 primary
fuel filter leak, February 27, 2008
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that Technical Specification operability was
properly justified and the subject component or system remained available such that no
unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the Technical Specifications and FSAR to
the licensees evaluations, to determine whether the components or systems were
operable. Where compensatory measures were required to maintain operability, the
inspectors determined whether the measures in place would function as intended and
were properly controlled. The inspectors determined, where appropriate, compliance
with bounding limitations associated with the evaluations. Additionally, the inspectors
also reviewed a sampling of corrective action documents to verify that the licensee was
identifying and correcting any deficiencies associated with operability evaluations.
Specific documents reviewed during this inspection are listed in the attachment.
This inspection constitutes eight samples as defined in Inspection
Procedure 71111.15-05.
b. Findings
No findings of significance were identified.
1R18 Plant Modifications (71111.18)
.1 Permanent Plant Modifications
a. Inspection Scope
The following engineering design package was reviewed and selected aspects were
discussed with engineering personnel:
- Design Change Package C-49857, Replacement of the containment recirculation
sump strainer, Revision 1
This document and related documentation were reviewed for adequacy of the
associated 10 CFR 50.59 safety evaluation screening, consideration of design
parameters, implementation of the modification, post-modification testing, and relevant
procedures, design, and licensing documents were properly updated. The inspectors
- 19 - Enclosure
observed ongoing and completed work activities to verify that installation was consistent
with the design control documents. The modification increases the emergency core
cooling recirculation sump net positive suction head in response to Generic Letter 2004-
02, Potential Impact of Debris Blockage on Emergency Recirculation during Design
Basis Accidents at Pressurized-Water Reactors. Specific documents reviewed during
this inspection are listed in the attachment.
This inspection constitutes one permanent modification sample as defined in Inspection
Procedure 71111.18.
b. Findings
No findings of significance were identified.
.2 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed the following temporary modifications:
- Action Request A0709926, Unit 2, Temporary modification to separate loose
parts monitoring system common power supply as part of Unit 2, steam
generator replacement project, January 23, 2008
- Action Request A0710453, Unit 2, Temporary modification to store and use
selected materials inside Unit 2 containment during Modes 1-4 prior to the Unit 2
Refueling Outage 14 steam generator replacement, January 24, 2008
The inspectors compared the temporary configuration changes and associated
10 CFR 50.59 screening and evaluation information against the design basis, the FSAR,
and the Technical Specifications, as applicable, to verify that the modification did not
affect the operability or availability of the affected systems. The inspectors also
compared the licensees information to operating experience information to ensure that
lessons learned from other utilities had been incorporated into the licensees decision to
implement the temporary modification. The inspectors, as applicable, performed field
verifications to ensure that the modifications were installed as directed; the modifications
operated as expected; modification testing adequately demonstrated continued system
operability, availability, and reliability; and that operation of the modifications did not
impact the operability of any interfacing systems. Lastly, the inspectors discussed the
temporary modification with operations, engineering, and training personnel to ensure
that the individuals were aware of how extended operation with the temporary
modification in place could impact overall plant performance. Specific documents
reviewed during this inspection are listed in the attachment.
This inspection constitutes two temporary modification samples as defined in Inspection
Procedure 71111.18.
b. Findings
No findings of significance were identified.
- 20 - Enclosure
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following postmaintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- Postmaintenance Test R0307896, Unit 2, Residual heat removal Pump 2-2
preventive maintenance, January 7, 2008
- Postmaintenance Test R0299924, Unit 1, Component cooling water Pump 1-1
preventive maintenance, January 22, 2008
- Postmaintenance Test R0308581, Unit 1, Auxiliary Feedwater Pump 1-1
preventive maintenance, January 31, 2008
- Postmaintenance Test C0217599, Unit 2, Containment Penetration 22 and 23
following repair of Valve CCW-2-695, February 23, 2008
- Postmaintenance Test WO R0270299, Unit 2, Containment Penetration 30
following repair of Valve CS-2-9011B, February 27, 2008
- Postmaintenance Test C0214829, Unit 2, Containment Penetration 50
following corrective maintenance, February 27, 2008
- Postmaintenance Test R0285525, Unit 1, Vital Battery Charger 1-1 preventative
maintenance, March 26, 2008
- Postmaintenance Test C0219100, Unit 2, Vital 4kV Bus H relay troubleshooting
and corrective maintenance, March 29, 2008
These activities were selected based upon the structure, system, or component's ability
to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate
for the maintenance performed; acceptance criteria were clear and demonstrated
operational readiness; test instrumentation was appropriate; tests were performed as
written in accordance with properly reviewed and approved procedures; equipment was
returned to its operational status following testing (temporary modifications or jumpers
required for test performance were properly removed after test completion), and test
documentation was properly evaluated. The inspectors evaluated the activities against
Technical Specifications, the FSAR, 10 CFR Part 50 requirements, licensee procedures,
and various NRC generic communications to ensure that the test results adequately
ensured that the equipment met the licensing basis and design requirements. In
addition, the inspectors reviewed corrective action documents associated with
postmaintenance tests to determine whether the licensee was identifying problems and
entering them in the corrective action program and that the problems were being
corrected commensurate with their importance to safety. Specific documents reviewed
during this inspection are listed in the attachment.
- 21 - Enclosure
This inspection constitutes eight samples as defined in Inspection Procedure 71111.19.
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the Unit 2,
refueling outage, between February 3 and March 31, 2008, to confirm that the licensee
had appropriately considered risk, industry experience, and previous site-specific
problems in developing and implementing a plan that assured maintenance of defense-
in-depth. During the refueling outage, the inspectors observed portions of the shutdown
and cooldown processes and monitored licensee controls over the outage activities
listed below. The inspectors also reviewed activities associated with the steam
generator replacement to fulfill the inspection requirements of Inspection
Procedure 50001, Steam Generator Replacement Inspection.
- Licensee configuration management, including maintenance of defense-in-depth
commensurate with the outage safety plan for key safety functions and
compliance with the applicable Technical Specifications when taking equipment
out of service
- Implementation of clearance activities and confirmation that tags were properly
hung and equipment appropriately configured to safely support the work or
testing
- Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error
- Controls over the status and configuration of electrical systems to ensure that
Technical Specifications and Outage Safety Plan requirements were met, and
controls over switchyard activities
- Monitoring of decay heat removal processes, systems, and components
- Controls to ensure that outage work was not impacting the ability of the operators
to operate the spent fuel pool cooling system
- Reactor water inventory controls including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss
- Controls over activities that could affect reactivity
- Maintenance of secondary containment as required by Technical Specifications
- Refueling activities, including fuel handling and sipping to detect fuel assembly
leakage
- 22 - Enclosure
- Licensee identification and resolution of problems related to refueling outage
activities
Specific documents reviewed during this inspection are listed in the attachment.
This inspection constitutes one refueling outage sample as defined in Inspection
Procedure 71111.20-05.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
.1 Routine Surveillance Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specification requirements:
- Surveillance R0284198-01, Unit 1, Phase A slave relays, February 4, 2008
- Surveillance R0289352, Unit 2, Low temperature overpressure protection
system, February 4, 2008
- Routine Unit 2, Shift checks required by licenses, February 6, 2008
- Surveillance R0311207, Unit 1, Auxiliary saltwater flow monitoring, February 11,
2008
- Surveillance R031125-01, Unit 1, Diesel generator, February 19, 2008
- Surveillance R0288943, Unit 2, 4kV Bus F auto-transfer, March 19, 2008
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; the calibration frequency was in accordance with Technical
Specifications, the FSAR, procedures, and applicable commitments; measuring and test
equipment calibration was current; test equipment was used within the required range
and accuracy; applicable prerequisites described in the test procedures were satisfied;
test frequencies met Technical Specification requirements to demonstrate operability
and reliability; tests were performed in accordance with the test procedures and other
- 23 - Enclosure
applicable procedures; jumpers and lifted leads were controlled and restored where
used; test data and results were accurate, complete, within limits, and valid; test
equipment was removed after testing; where applicable, test results not meeting
acceptance criteria were addressed with an adequate operability evaluation or the
system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of
the safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program. Specific documents
reviewed during this inspection are listed in the attachment.
This inspection constitutes six routine surveillance testing samples as defined in
Inspection Procedure 71111.22.
b. Findings
No findings of significance were identified.
.2 Inservice Testing Surveillance
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specification requirements:
- Surveillance R0309435, Unit 1, Turbine-driven auxiliary feedwater steam stop
Valve FCV-95, January 31, 2008
- Surveillance R0286556, Unit 1, Steam supply to turbine-driven auxiliary
feedwater turbine Valves FCV-37 and FCV-38, January 31, 2008
- Surveillance R0309344, Unit 1, Auxiliary feedwater pump discharge
Valves LCV-106, 107, 108, and 109, January 31, 2008
- Surveillance R0308743-01, Auxiliary saltwater Pump 1-2 crosstie
Valve FCV-495, February 11, 2008
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left set-points
were within required ranges; and the calibration frequency was in accordance with
- 24 - Enclosure
Technical Specifications, the FSAR, procedures, and applicable commitments;
measuring and test equipment calibration was current; test equipment was used within
the required range and accuracy; applicable prerequisites described in the test
procedures were satisfied; test frequencies met Technical Specification requirements to
demonstrate operability and reliability; tests were performed in accordance with the test
procedures and other applicable procedures; jumpers and lifted leads were controlled
and restored where used; test data and results were accurate, complete, within limits,
and valid; test equipment was removed after testing; where applicable for inservice
testing activities, testing was performed in accordance with the applicable version of
Section XI, American Society of Mechanical Engineers Code, and reference values were
consistent with the system design basis; where applicable, test results not meeting
acceptance criteria were addressed with an adequate operability evaluation or the
system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of its
safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program. Specific documents
reviewed during this inspection are listed in the attachment.
This inspection constitutes four inservice inspection samples as defined in Inspection
Procedure 71111.22.
b. Findings
No findings of significance were identified.
.3 Reactor Coolant System Leak Detection Inspection Surveillance
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specifications requirements:
- Routine daily checks required by licensees, Unit 1, March 24, 2008
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left set-points
were within required ranges; and the calibration frequency was in accordance with
Technical Specifications, the FSAR, procedures, and applicable commitments;
measuring and test equipment calibration was current; test equipment was used within
the required range and accuracy; applicable prerequisites described in the test
- 25 - Enclosure
procedures were satisfied; test frequencies met Technical Specifications requirements to
demonstrate operability and reliability; tests were performed in accordance with the test
procedures and other applicable procedures; jumpers and lifted leads were controlled
and restored where used; test data and results were accurate, complete, within limits,
and valid; test equipment was removed after testing; where applicable, test results not
meeting acceptance criteria were addressed with an adequate operability evaluation or
the system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of its
safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program. Specific documents
reviewed during this inspection are listed in the attachment.
This inspection constitutes one reactor coolant system leak detection inspection sample
as defined in Inspection Procedure 71111.22.
b. Findings
No findings of significance were identified.
.4 Containment Isolation Valve Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specification requirements:
- Local Leak Rate Test R0286798, Unit 2, Containment Penetrations 22 and 23,
February 10 through 22, 2008
- Local Leak Rate Test R0264996, Unit 2, Containment Penetration 50,
February 19 through 24, 2008
- Local Leak Rate Test R0286800, Unit 2, Containment Penetration 30,
February 27, 2008
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; and the calibration frequency was in accordance with
Technical Specifications, the FSAR, procedures, and applicable commitments;
measuring and test equipment calibration was current; test equipment was used within
- 26 - Enclosure
the required range and accuracy; applicable prerequisites described in the test
procedures were satisfied; test frequencies met Technical Specifications requirements to
demonstrate operability and reliability; tests were performed in accordance with the test
procedures and other applicable procedures; jumpers and lifted leads were controlled
and restored where used; test data and results were accurate, complete, within limits,
and valid; test equipment was removed after testing; where applicable, test results not
meeting acceptance criteria were addressed with an adequate operability evaluation or
the system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of its
safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program. Specific documents
reviewed during this inspection are listed in the attachment.
This inspection constitutes three containment isolation valve inspection samples as
defined in Inspection Procedure 71111.22.
b. Findings
No findings of significance were identified.
1EP6 Emergency Preparedness Evaluation (71114.06)
Training Observation
a. Inspection Scope
The inspectors observed a simulator training evolution for licensed operators on
January 3, 2008, which required emergency plan implementation by a licensee
operations crew. This evolution was evaluated and included in performance indicator
data regarding drill and exercise performance. The inspectors observed event
classification and notification activities performed by the crew. The inspectors also
attended the post-evolution critique for the scenario. The focus of the inspectors
activities was to note any weaknesses and deficiencies in the crews performance and
ensure that the licensee evaluators noted the same issues and entered them into the
corrective action program. As part of the inspection, the inspectors reviewed Emergency
Plan Training Scenario, Session 07-5.
This inspection constitutes one sample as defined in Inspection Procedure 71114.06-05.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
- 27 - Enclosure
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 inspectors used the
requirements in 10 CFR Part 20, Technical Specifications, and the licensees procedures
required by Technical Specifications as criteria for determining compliance. The
inspectors also reviewed activities associated with the steam generator replacement to
fulfill the inspection requirements of Inspection Procedure 50001, Steam Generator
Replacement Inspection. During the inspection, the inspectors interviewed the radiation
protection manager, radiation protection supervisors, and radiation workers. The
inspectors 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 three radiation, high radiation, or
airborne radioactivity areas
- Radiation work permits, procedures, engineering controls, and air sampler
locations
- Conformity of electronic personal dosimeter alarm setpoints 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 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 work permit briefings and worker instructions
- 28 - Enclosure
- Adequacy of radiological controls, such as required surveys, radiation protection
job coverage, and contamination control during job performance
- 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 rates - high radiation
areas and very high radiation areas
- Radiation worker and radiation protection technician performance with respect to
radiationprotection work requirements
Specific documents reviewed during this inspection are listed in the attachment.
This inspection constitutes 20 samples as defined in Inspection Procedure 71121.01.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1 for failure to follow a licensee procedure.
Description. While touring the Unit 2 spent fuel pool on February 13, 2008, the
inspectors observed workers performing fuel inspections on the fuel bridge. Radiation
Work Permit 08-2019-00 requires a continuous air monitor be operating in the fuel
building, with an appropriate alarm setpoint to alert workers and provides actions for
workers to take upon receiving an alarm. The inspectors noted that the physical location
of the continuous air monitor, an AM-4, was in the southeast corner of the floor. The
function of the continuous air monitor is to monitor for airborne radioactive materials
while fuel inspection is performed. Furthermore, Site Procedure RCP D-430, Plant
Airborne Radioactivity Surveillance, Section 2.2.3 states, in part, the purpose of the
continuous air monitors is to alert personnel to changes in radiological conditions.
Ventilation flow in this area is from north to south with the exhaust intakes centered with
the spent fuel pool. The continuous air monitor was approximately 18 feet away from
the nearest exhaust intake and approximately 50 feet away from the workers location.
The permanently installed continuous air monitor was out of service; however, it was
physically located beneath an exhaust intake. Personnel interviews indicated that the
AMS-4 was originally placed on top of the permanently installed continuous air monitor,
but then it was moved to get a better remote indication. However, the inspectors
concluded, from discussions with radiation protection supervision, that no evaluation was
made to determine if the new location was appropriate to alert workers of changing
radiological conditions.
- 29 - Enclosure
During review of this occurrence, the inspectors were made aware of a similar situation
that was identified on May 3, 2006. Specifically, Action Request A0666110 was opened
to evaluate the adequacy of AMS-4 placement in the fuel building during fuel moves.
The corrective action was initiated in response to an NRC inspectors questions during a
walkthrough. However, this action request remained open with a resolution date of
December 15, 2008.
Analysis. This finding is more than minor because it is associated with the occupational
radiation safety program and process attribute and affected the cornerstone objective, in
that the failure to monitor for radioactive material in the air had the potential to increase
personnel dose. This occurrence involves workers unplanned, unintended or potential
for such dose; therefore, this finding was evaluated using the occupational radiation
safety significance determination process. The inspectors determined that this finding
was of very low safety significance because it did not involve: (1) an as low as is
reasonably achievable (ALARA) planning or work control issue; (2) an overexposure;
(3) a substantial potential for overexposure; or (4) an impaired ability to assess dose.
This finding also has a crosscutting aspect in the area of problem identification and
resolution, corrective action component, because the licensee failed to take timely
corrective actions to address personnel safety issues. P.1(d)
This finding was identified by NRC because the NRC inspectors questioned the position
of the AMS-4.
Enforcement. Technical Specification 5.4.1 requires procedures be established,
implemented, and maintained covering the applicable procedures recommended in
Regulatory Guide 1.33, Appendix A. Section 7 of Appendix A recommends radiation
protection procedures for airborne radioactivity monitoring. The licensee implementing
Procedure RCP D-430, Plant Airborne Radioactivity Surveillance, Section 2.2 states, in
part, the purpose of the continuous air monitors is to alert personnel to changes in
radiological conditions and that locations are selected based on their potential as
contributors to airborne activity. Contrary to this requirement, the licensee failed to
implement this procedure because the selected location of the continuous air monitor did
not provide adequate coverage to alarm and alert the workers of changes in radiological
conditions. Because this failure to follow a procedure is of very low safety significance
and has been entered into the licensees corrective action program, Action
Request A0719338, this violation is being treated as a noncited violation, consistent with
Section VI.A of the NRC Enforcement Policy: NCV 05000323/2008002-03, Failure to
Follow Procedures.
2OS2 ALARA Planning and Controls (71121.02)
a. Inspection Scope
The inspectors assessed licensee performance with respect to maintaining individual
and collective radiation exposures as low as is reasonably achievable (ALARA). The
inspectors used the requirements in 10 CFR Part 20 and the licensees procedures
required by technical specifications as criteria for determining compliance. The
inspectors also reviewed activities associated with the steam generator replacement to
fulfill the inspection requirements of Inspection Procedure 50001, Steam Generator
Replacement Inspection. The inspectors interviewed licensee personnel and reviewed:
- 30 - Enclosure
- Five outage or online maintenance work activities scheduled during the
inspection period and associated work activity exposure estimates which were
likely to result in the highest personnel collective exposures
- 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
- 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 followup
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
Specific documents reviewed during this inspection are listed in the attachment.
This inspection constitutes 12 samples of ALARA planning and controls as defined in
Inspection Procedure 71121.02.
b. Findings
No findings of significance were identified.
- 31 - Enclosure
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the Fourth
Quarter 2008 performance indicators for any obvious inconsistencies prior to its public
release in accordance with IMC 0608, Performance Indicator Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b. Findings
No findings of significance were identified.
.2 Unplanned Scrams per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned scrams per 7000 critical
hours performance indicator for Units 1 and 2 for the first through fourth quarters of
2007. To determine the accuracy of the performance indicator data reported during
those periods, performance indicator definitions and guidance contained in Revision 5 of
the Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment
Performance Indicator Guideline, were used. The inspectors reviewed the licensees
operator narrative logs, issue reports, event reports and NRC Inspection reports for the
period of first through fourth quarters of 2007 to validate the accuracy of the submittals.
The inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified.
This inspection constitutes one unplanned scrams per 7000 critical hours sample as
defined by Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
.3 Unplanned Scrams with Complications
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned scrams with
complications performance indicator for Units 1 and 2 for the first through fourth quarters
of 2007. To determine the accuracy of the performance indicator data reported during
those periods, performance indicator definitions and guidance contained in Revision 5 of
the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline,
- 32 - Enclosure
were used. The inspectors reviewed the licensees operator narrative logs, issue
reports, event reports and NRC integrated inspection reports for the period of first
through fourth quarters of 2007 to validate the accuracy of the submittals. The
inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified.
This inspection constitutes one unplanned scrams with complications sample as defined
by Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
.4 Unplanned Transients per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned transients per 7000
critical hours performance indicator for Units 1 and 2 for the first through fourth quarters
of 2007. To determine the accuracy of the performance indicator data reported during
those periods, performance indicator definitions and guidance contained in Revision 5 of
the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline,
were used. The inspectors reviewed the licensees operator narrative logs, issue
reports, maintenance rule records, event reports and NRC integrated inspection reports
for the period of the first through fourth quarters of 2007 to validate the accuracy of the
submittals. The inspectors also reviewed the licensees issue report database to
determine if any problems had been identified with the performance indicator data
collected or transmitted for this indicator and none were identified.
This inspection constitutes one unplanned transients per 7000 critical hours sample as
defined by Inspection Procedure 71151.
b. Findings
No findings of significance were identified.
.5 Occupational Radiation Safety
a. Inspection Scope
The inspectors reviewed licensee documents from October 1, 2007 through
December 31, 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 NEI 99-02, "Regulatory Assessment Indicator
Guideline," Revision 5). Additional records reviewed included ALARA records and whole
body counts of selected individual exposures. The inspectors interviewed licensee
personnel that were accountable for collecting and evaluating the performance indicator
data. In addition, the inspectors toured plant areas to verify that high radiation, locked
high radiation, and very high radiation areas were properly controlled. Performance
- 33 - Enclosure
indicator definitions and guidance contained in NEI 99-02, Revision 5, were used to
verify the basis in reporting for each data element.
This inspection constitutes one occupational radiation safety sample as defined by
b. Findings
No findings of significance were identified.
.6 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual
Radiological Effluent Occurrences
a. Inspection Scope
The inspectors reviewed licensee documents from October 1, 2007 through
December 31, 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
inspectors 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.
This inspection constitutes one sample of radiological effluent technical
specification/offsite dose calculation manual radiological effluent occurrences as defined
by Inspection Procedure 71151.
b. Findings
No findings of significances were identified.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical
Protection
.1 Routine Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspector routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. Attributes reviewed included: the complete and accurate identification of the
problem; that timeliness was commensurate with the safety significance; that evaluation
- 34 - Enclosure
and disposition of performance issues, generic implications, common causes,
contributing factors, root causes, extent of condition reviews, and previous occurrences
reviews were proper and adequate; and that the classification, prioritization, focus, and
timeliness of corrective actions were commensurate with safety and sufficient to prevent
recurrence of the issue. Minor issues entered into the licensees corrective action
program as a result of the inspectors observations are included in the attached list of
documents reviewed.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure they were considered an
integral part of the inspections performed during the quarter and are documented in
Section 1 of this report.
Specific documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for followup, the inspectors performed a daily screening of
items entered into the licensees corrective action program. This review was
accomplished through inspection of the stations daily condition report packages.
These daily reviews were performed by procedure as part of the inspectors daily plant
status monitoring activities and, as such, did not constitute any separate inspection
samples.
b. Findings
No findings of significance were identified.
.3 Selected Issue Followup Inspection
a. Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors completed an in-depth review of:
- Action Request A0716519, NRC problem identification adverse trend,
January 15, 2008
- Action Request A0717510, Inattentive operator, January 29, 2008
- Identification and resolution of problems associated with the steam generator
replacement project
- 35 - Enclosure
The above constitutes completion of three in-depth problem identification and resolution
samples.
b. Findings
No findings of significance were identified.
.4 Occupational Radiation Safety
a. Inspection Scope
The inspectors 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
Section 2OS1 describes a finding with crosscutting aspects associated with problem
identification and resolution.
4OA5 Other
A. Temporary Instruction 2515/166, APressurized Water Reactor Containment Sump
Blockage@, Diablo Canyon Units 1 and 2 (Closed)
Temporary Instruction 2515/166 was performed at Diablo Canyon Power Plant, Unit 1
during May 2007, and documented in Inspection Report 05000275/2007003.
Subsequent inspection of Diablo Canyon Power Plant Unit 2 is documented in this
report. The inspection phase of Temporary Instruction 2515/166 for Units 1 and 2 is
complete.
O3.01 Verify the implementation of the plant modifications and procedure changes committed
to by the licensee in their Generic Letter 2004-02 responses. Listed below are the
commitments and actions taken by Diablo Canyon Unit 1 and 2:
1. Install larger sump screens.
Actions Taken
Installed and documented in Diablo Canyon Procedure C-50844 and DCP
C - 50857, Action Request 0701461
2. Modify reactor cavity door (Door 278-2)
Actions Taken
Work completed and documented in AR A0648630.
3. Add three 18-inch high perforated plate debris interceptors on doors 275-2, 276-2
and 277-2 in the crane wall.
- 36 - Enclosure
Actions Taken
Work completed and documented in AR A0687983.
4. Install RMI and/or other approved encapsulated fibrous insulation on the
replacement steam generators and the steam generator belly bands.
Actions Taken
Work completed and documented in DEP M-50754 and AR A0642989.
5. Remove cable tray fire stops inside the crane wall which are inside the pipe
break zone of influence.
Actions Taken
Work completed and documented in AR A0676978 and WO C0213262-01
and C0214501-01.
6. Install multiple banding on cal-sil piping insulation inside the pipe break zone of
influence.
Actions Taken
Work completed and documented in AR A0693591.
7. Install stainless steel jacketing on Temp-Mat piping insulation inside the pipe
break zone of influence.
Actions Taken
Work completed and documented in AR A0693786.
8. Install tray covers to protect the pressurizer heater cable insulation in cable trays.
Actions Taken
Work completed and documented in AR A0688131.
9. Install encapsulated Temp-Mat insulation on the inlet to Pressurizer Safety
Valves 8010A, 8010B and 8010C.
Actions Taken
Work completed and documented in AR A0693786.
10. Conduct an evaluation of downstream debris ingestion effects.
Actions Taken
Evaluation completed and documented in AR A0703421-05.
11. Conduct downstream effects evaluation for erosive wear on ECCS and CSS
valves.
Actions Taken
Evaluation completed with satisfactory results and documented in
AR A0703421-06.
- 37 - Enclosure
12. Conduct a downstream effects evaluation of auxiliary equipment.
Actions Taken
Evaluation completed with satisfactory results and documented in
AR A0703421-07.
13. Conduct an evaluation of the ECCS pumps disaster bushing leakage.
Actions Taken
Evaluation completed with satisfactory results and documented in
Calculation M-1113 R0
14. Conduct a fuel blockage evaluation.
Actions Taken
Evaluation completed with satisfactory results and documented in
AR A0703421-04.
15. Conduct a LOCA deposition model fuel evaluation.
Actions Taken
Evaluation completed with satisfactory results and documented in
AR A0703421-70.
16. Change procedure EOP E-1.3, Transfer to Cold-leg Recirculation.
Actions Taken
Change implemented and documented in AR A0701461.48.
17. Change procedure EOP E-1. Loss of Reactor or Secondary Coolant.
Actions Taken
Change implemented and documented in AR A0701461.48.
18. Change procedure EOP ECA-1.3, Sump Blockage Guideline.
Actions Taken
Change implemented and documented in AR A0701461.48.
19. Change procedure PEP EN-1, Post Accident Mitigation Diagnostic Aids and
Guidelines.
Actions Taken
Change implemented and documented in AR A0720403-03.
20. Change procedure STP R-20, Boric Acid Inventory.
Action Taken
Change implemented and documented in AR A0690337-10.
- 38 - Enclosure
21. Change procedure STP M-45A, Containment Inspection Prior to Establishing
Containment Integrity.
Action Taken
Change implemented and documented in AR A0701461-75.
22. Change procedure STP M-45B, Containment Inspection When Containment
Integrity is Established.
Action Taken
Change implemented and documented in AR A0718227-03.
23. Change procedure STP M-45C, Outage Management Containment Inspection.
Action Taken
Change implemented and documented in AR A0718227-04.
24. Change Procedure CF3.ID9, Design Change Development.
Action Taken
Change implemented and documented in CF3.ID9 R32.
25. Change Procedure MIP C-4.0, Thermal Insulation.
Action Taken
Change implemented and documented in MIP C-4.0 R4.
26. Change Procedure AD7.DC8, Work Control.
Action Taken
Change implemented and documented in AD7.DC8 R27.
29. Change Procedure AD4.ID9, Containment Housekeeping and Material
Controls.
Action Taken
Change implemented and documented in AR A0718227-05.
30. Change Technical Specification 3.5.4, Refueling Water Storage Tank and
Surveillance Requirement 3.5.4.2, to increase the minimum required borated
water volume from equal to or greater than 400,000 gallons (81.5 percent
indicated level) to equal to or greater than 455,300 gallons.
Action Taken
Technical specification amendment submitted and approved by NRC on
March 26, 2008.
B. Temporary Instruction 2515-172, Reactor Coolant System Dissimilar Metal Butt Welds
- 39 - Enclosure
Temporary Instruction TI 2515/172, Reactor Coolant System Dissimilar Metal Butt
Welds was performed at Diablo Canyon during Refueing Outage 2R14 in February and
March 2008.
O3.01 Licensees Implementation of the MRP-139 Baseline Inspections
a. MRP-139 baseline inspections:
The inspectors observed performance and reviewed records of structural weld
overlays and nondestructive examination activities associated with the Diablo
Canyon Unit 2 pressurizer structural weld overlay mitigation effort. The baseline
inspections of the pressurizer dissimilar metal butt welds (DMBWs) were
completed during the spring 2008 refueling outage.
b. At the present time, the licensee is not planning to take any deviations from the
baseline inspection requirements of MRP-139, and all other applicable DMBWs
are scheduled in accordance with MRP-139 guidelines.
03.02 Volumetric Examinations
a. There were no inspections of unmitigated pressurizer DMBWs performed during
this outage. The inspectors reviewed the ultrasonic examination records of the
unmitigated hot leg and cold leg DMBWs (Welds WIB-RC-2-1[SE] and
WIB-RC-3-16[SE]), respectively, performed on April 29, 2006. These
examinations were conducted in accordance with the MRP-139 guidelines
(i.e., personnel, procedures, and equipment qualified in accordance with ASME
Code,Section XI, Supplement VIII [PDI] requirements).
No relevant conditions or deficiencies were identified during the examinations of
the hot and cold leg unmitigated DMBWs, or the mitigated pressurizer DMBWs.
b. Inspectors directly observed and/or reviewed records of NDE performed on
pressurizer weld overlays. This effort is documented in Section 1R08 of this
inspection report.
For each weld overlay inspected the licensee submitted and received NRC
approval by letter dated February 6, 2008, for the use of Relief Request REP-1
U2, The Application of Weld Overlay on Dissimilar Metal Welds of Pressurizer
Nozzles, Revision 1.
Inspection coverage met requirements of MRP-139.
No relevant conditions were identified.
c. The certification records of ultrasonic examination personnel used in the
examination of the unmitigated hot and cold legs DMBWs, and the mitigated
pressurizer DMBWs were reviewed. All personnel records showed that they
were qualified under the EPRI Performance Demonstration Initiative.
d. No deficiencies were identified during the NDE.
- 40 - Enclosure
03.03 Weld Overlays
a. The inspectors observed structural weld overlay welding and reviewed records
pertaining to the pressurizer nozzles and determined that welding was performed
in accordance with ASME Code Section IX requirements. Welding inspections
are documented in section 1R08 of this inspection report.
b. The licensee submitted and received NRC approval by letter dated February 6,
2008, for the use of Relief Request REP-1 U2, The Application of Weld Overlay
on Dissimilar Metal Welds of Pressurizer Nozzles, Revision 1.
c. The qualification records of welders were reviewed and all qualifications were
current.
d. No relevant conditions were identified.
03.04 Mechanical Stress Improvement
This item is not applicable because the licensee did not employ a mechanical stress
improvement process.
03.05 Inservice inspection program
The licensee MRP-139 inservice inspection program has basically been controlled
through the Action Request Program to assure that requirements identified in the
MRP-139 guidelines are not inadvertently missed. As such, the MRP-139 inservice
inspection program is in-process, although it was recognized that this may not be the
most appropriate way to control DMBW locations and scheduling requirements. The
licensee initiated Action Request AR A0725407 to update MRP-139 tracking and
planning documents, and to create an appropriate scheduling mechanism. This item will
receive further in-office inspection at a later date.
The inspectors review determined that the hot leg and cold leg DMBWs are
appropriately categorized in accordance with MRP-139 requirements. Categorization of
all other DMBWs will receive further in-office inspection at a later date.
With the exception of the pressurizer nozzle DMBWs, which were categorized as H, no
other DMBWs were categorized as either H or I. The structural weld overlay
mitigation effort removed the pressurizer nozzles from Category H.
The licensees MRP-139 Inservice Inspection Program will receive additional in-office
review at a later date.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On March 28, 2008, the inspectors presented the results of this inservice inspection to
Mr. Jim Becker, Site Vice President, and other members of licensee management.
Licensee management acknowledged the inspection findings. The inspectors returned
proprietary material examined during the inspection.
- 41 - Enclosure
On April 1, 2008, the inspectors presented the inspection results to Mr. J. Becker, and
other members of your staff. The licensee acknowledged the issues presented. The
inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
On February 15, 2008, the inspectors presented the occupational radiation safety
inspection results to Mr. M. Somerville, Radiation Protection Manager, and other
members of your staff who acknowledged the findings. On March 14, 2008, the
inspectors presented the inspection results to Mr. L. Parker, Acting Regulatory Services
Manager, and other members of your staff who acknowledged the findings by
teleconference. The inspectors confirmed that proprietary information was not provided
or examined during the inspection.
ATTACHMENT: SUPPLEMENTAL INFORMATION
- 42 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
PG&E Personnel
J. Becker, Vice President - Diablo Canyon Operations and Station Director
R. Brown
W. Cote
C. Dougherty
R. Hite, Manager, Radiation Protection
D. Gonzalez
S. Ketelsen, Manager, Regulatory Services
K. Langdon, Director, Operations Services
M. Meko, Director, Site Services
K. Peters, Director, Engineering Services
K. Shatell
M. Somerville, Manager, Radiation Protection
S. Zawalick
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000275; Failure to Maintain the Integrity of an Auxiliary Building Fire
NCV 05000323/2008002-01 Door (Section 1R05)
05000275; Failure to Demonstrate that the Unit 2 Containment
05000323/2008002-02 Atmosphere Particulate Radioactivity Monitor Performance
was Being Effectively Controlled per 10 CFR 50.65(a)(2)
(Section 1R12)
05000275; Failure to Follow Procedures, per Technical
NCV 05000323/2008002-03 Specification 5.4.1 (Section 2OS1)
LIST OF DOCUMENTS REVIEWED
1R01: Adverse Weather
Procedures
CP M-12, Stranded Plant, Revision 3A
Action Requests
A0700848 A0713166 A0713716 A0714757 A0715124
A-1 Attachment
Other Documents
Meeting notes, Operational Decision Making Meeting, January 3, 2008
1R04: Equipment Alignment
Procedures
OP F-2:1, Component Cooling Water System, Make Available, Revision 29
Action Requests
A0581569 A0709594 A0661827
Drawings
106714, Unit 1 Component Cooling Water System, Sheet 1, Revision 59
106714, Unit 1 Component Cooling Water System, Sheet 2, Revision 56
106714, Unit 1 Component Cooling Water System, Sheet 3, Revision 49
108008, Unit 2 Chemical & Volume Control System, Sheet 1, Revision 83
108008, Unit 2 Chemical & Volume Control System, Sheet 2, Revision 11
108008, Unit 2 Chemical & Volume Control System, Sheet 3, Revision 89
108008, Unit 2 Chemical & Volume Control System, Sheet 4, Revision 80
108008, Unit 2 Chemical & Volume Control System, Sheet 4A, Revision 78
108008, Unit 2 Chemical & Volume Control System, Sheet 4B, Revision 93
108008, Unit 2 Chemical & Volume Control System, Sheet 4C, Revision 0
108008, Unit 2 Chemical & Volume Control System, Sheet 5, Revision 67
108008, Unit 2 Chemical & Volume Control System, Sheet 5A, Revision 54
108008, Unit 2 Chemical & Volume Control System, Sheet 5B, Revision 85
108008, Unit 2 Chemical & Volume Control System, Sheet 5C, Revision 72
108008, Unit 2 Chemical & Volume Control System, Sheet 6, Revision 44
108008, Unit 2 Chemical & Volume Control System, Sheet 7, Revision 84
108008, Unit 2 Chemical & Volume Control System, Sheet 8, Revision 84
108008, Unit 2 Chemical & Volume Control System, Sheet 9, Revision 77
108008, Unit 2 Chemical & Volume Control System, Sheet 10, Revision 6
108008, Unit 2 Chemical & Volume Control System, Sheet 11, Revision 7
108008, Unit 2 Chemical & Volume Control System, Sheet 12, Revision 17
108008, Unit 2 Chemical & Volume Control System, Sheet 13, Revision 38
108008, Unit 2 Chemical & Volume Control System, Sheet 14, Revision 55
108008, Unit 2 Chemical & Volume Control System, Sheet 15, Revision 73
108008, Unit 2 Chemical & Volume Control System, Sheet 16, Revision 83
Other Documents
Diablo Canyon Nuclear Power Plant Units 1 and 2, Design Criteria Memorandum, S-8 and
Volume Control System, Revision 30B
A-2 Attachment
1R05: Fire Protection
Procedure
OM8.ID2, Fire System Impairment, Revision 13
Work Order
Roving Fire Watch Check Lists completed for February 9, 10, 16, and 17, 2008
Action Request
A0718292
1R08: Inservice Inspection Activities
Procedures
WDI-ET-008, IntraSpect Eddy Current Inspection of Vessel Head Penetration J-Welds and Tube
OD Surfaces, Revision 8
WDI-ET-013, IntraSpect UT Analysis Guidelines, Revision 12
ISI X-CRDM, Reactor Vessel Top and Bottom Head Visual Inspections, Revision 4A
CF5-DC2, Welding Filler Material Control, Revision 10
NDE PDI-UT-2, Ultrasonic Examinations of Austenitic Piping
54-ISI-838-09, Manual Ultrasonic Examination of Weld Overlaid Similar and Dissimilar Metal
Welds, Revision 3
PDI-UT-8, Generic Procedure for the Ultrasonic Examination of Weld Overlaid Similar and
Dissimilar Metal Welds, Revision F
54-PT-200-07, Color Contrast Solvent Removable Liquid Penetrant Examinations of
Components, Revision 7
PDI-ISI-254-SE, Ultrasonic Examination of Dissimilar Welds, Revision 2
Calculation
CN-NCE-DCPPRSG-12, Feedwater Nozzle and Thermal Sleeve Analysis, Revision 1
A-3 Attachment
Corrective Action Documents
A0717850 A0719528 A0718124 A0674071
A0718292 A0719824 A0719065 A0725407
A0718661 A0720014 A0719829
A0719033 A0716746 A0712487
A0719321 A0717199 A0712484
Drawings
2-2-48, Charging Injection - Out, Revision 2
8019491D, Diablo Canyon Unit 2 Pressurizer Spray Nozzle Overlay Implementation, Revision 2
8019493D, Diablo Canyon Unit 2 Pressurizer Safety and Relief Nozzle Overlay Implementation.
Revision 2
8023646B, Diablo Canyon Unit 2 Pressurizer Spray Nozzle SWOL Contour Template,
Revision 0
8023647B, Diablo Canyon Unit 2 Pressurizer Surge Nozzle SWOL Contour Template,
Revision 0
8019492D, Diablo Canyon Unit 2 Pressurizer Surge Nozzle Overlay Implementation, Revision 2
Miscellaneous
Relief Request RR REP-1 U2, Application of Weld Overlay on Dissimilar and Similar Metal
Welds of the Pressurizer Relief Valve, Safety Vaves, Spray Line, and Surge Line Nozzles for
the Third 10-year ISI Interval at DCPP Unit 2, Revision 1
ESH-102, Safety Evaluation by the Office of Nuclear Reactor Regulation Request for relief from
the AMSE Boiler and Pressure Vessel Code,Section XI, ISI Program Pacific Gas & Electric Co.
Diablo Canyon Power Plant, Unit 2, Docket 50-323, Revision 0
Alloy 600 Program Review, 9/5/06
A-4 Attachment
Welding Procedure Specifications and their Supporting Procedure Qualification Records
Welding Procedure Specification 11, Welding of P8 Materials with GTAW and/or SMAW, ASME
I, ASME III, ANSI B31.1, and AWS D5.2, Procedure Qualification Records 201, 235, and 499,
Revison 8
Welding Procedure Specification 3/8/F43OLTBSCa3, Machine Temper Bead Overlay GTAW,
Procedure Qualification Records 7164, 7213, 7280, and 7281, Revision 3
1R12: Maintenance Effectiveness
Issue Report
RPE Number P-7401 Rev 00 RC-2: C&S Design Class I Duo Check Valve Parts
Procedure
MA1.ID17, Maintenance Rule Monitoring Program, Revision 18
Work Order
C0217599
Action Requests
A0718996 A0584087 A0584097 A0671226 A0697363 A0709074
A0709405 A0712454 A0712518 A0717009 A0717151 A0716671
1R15: Operability Evaluations
Procedures
STP M-51, Routine Surveillance Test of Containment Fan Cooler Units, Revision 15A
STP M-93A, Refueling Interval Surveillance - Containment Fan Cooler System, Revision 20
AR PK01-16, Annunciator Response - Containment Environment PPC, Revision 4
OM7.ID12, Operability Determination, Revision 11
STP-86, Leak Reduction of Systems Outside Containment Likely to Contain Radioactive
Materials Following an Accident, Revision 19
STP M-21-ENG.1, Diesel Generator Inspection, Revision 8
MP M-54.1, Bolt Fabrication and Tensioning, Revision 20
Action Requests
A0407497 A0411426 A0709301 A0709957 A0714266 A0718586
A-5 Attachment
Calculation
Fuel Handling Building Steel Superstructure, Revision 4
Other Document
USNRC Information Notice 2007-27 dated August 6, 2007, Recurring Events Involving
Emergency Diesel Generator Operability
1R18: Plant Modifications
Procedures
CF4.ID7, Temporary Modifications, Revision 19
STP M-45B, Containment Inspection When Containment Integrity is Established, Revision 12
Action Request
A0643070
Work Order
C0216374-1, Build Frames/Stage Scaff Matl IAW EM-TMOD, January 7, 2008
C0216374-2, Stage Cables, El. Panels, Transfmrs, IAW EM-TMOD, January 23, 2008
C0216374-3, Stage Joboxs, Harnesses & A-Frame IAW EM-TMOD, January 8, 2008
C0216374-4, Stage Lead Shielding in Boxes IAW EM-TMOD, January 18, 2008
C0216374-5, Stage Machining Equipment IAW EM-TMOD, January 24, 2008
C0216405-1, Stage Sump Material in Containment IAW EM-TMOD, January 28, 2008
Drawing
452418, Rear View Loose Parts Monitoring Rack, Revision 14
Calculations
Unit 2 Design Calculation, N-217, Containment Coatings Tracking, Revision 8
SGRP Project Letter, SGRP-07-1057, Temporary Modification A0710453 Installation
Instructions and Applicability Determination, November 20, 2007
Calculation ALION-REP-DCPP-2830-001, Diablo Canyon Characterization of Events that May
Lead to ECCS Recirculation, Revision 0
Calculation GE-NE-0000-0064-1369-P-R2, May 2007, Residual Heat Removal Pump ECCS
Trainer System S0100 Hydraulic Sizing Report
Calculation M-580, Determination of Post LOCA Flood Water Levels Inside Containment Units 1
and 2, Revision 4
Calculation M-591, Determination of the Head Loss Across the Recirculation Sump Screen
Structure, Revision 28
A-6 Attachment
Calculation N-100, Maximum Flow From ECCS Pumps an Minimum Flow to Containment Spray
Header, Revision 2
Calculation N-22b7, Post-LOCA Minimum Containment Sump Level, Revision 3
Containment Recirculation Sump Strainer Diablo Canyon Power Plant Units 1 and 2,
Contract 3500736064, September 29, 2006
Specification 10070-M-NPG, Diablo Canyon Power Plant Units 1 and 2,,Containment
Recirculation Sump Strainer Specification, September 29, 2006
1R19: Post Maintenance Testing
Procedures
STP P-RHR-22, Routine Surveillance Test of RHR Pump 2-2, Revision 19
STP P-AFW-11, Routine Surveillance Test of Turbine-Driven Auxiliary Feedwater Pump 1-1,
Revision 24
OP F-2:II, Component Cooling Water System Changing Over Pumps and Common
Components, Revision10
STP M-16, Integrated Test of Engineered Safeguards and Diesel Generators, Revision 40
STP-650, Penetration 50 Containment Isolation Valve Leak Test, Revision 11
CF3.ID13, Replacement or New Part Evaluation, AT-RPE AR and CITE, Revision 19A
STP-623, Penetration 22 and 23 Containment Isolation Valve Leak Test, Revision 7
Action Requests
A0715884 A0725117 A0718341 A0718996 A0720488
Drawings
Dual Plate Check Valve Assembly Drawing, Revision 2
8- 130 Swing Check Valve Cast Stain STL- Butt Weld Ends Stellite Trim, Revision 6
Miscellaneous
Valve 9011B Leak Rate History
Generic Check Valve Inspection As Found for CS-2-9011B (MP M-51.14)
Generic Check Valve Inspection As Left for CS-2-9011B (MP M-51.14)
Generic Check Valve Inspection for CCW-2-695 (MP M-51.14)
RPE Number: P-7401 Revised August 11, 2002
A-7 Attachment
1R20: Outage Activities
Procedures
OP O-32, Unit Attachment 3, Charging pump 2-1, Revision 0
AP SD-0, Loss of, or Inadequate Decay Heat Removal, Revision 11A
AD8.DC55, Unit 2 Outage Safety Checklist - Core Offloaded, Revision 27
AD8.DC51, Outage Safety Management Control of Off-Site Power Supplies to Vital Buses,
Revision 12A
Other Documents
Unit 2, Fuel Assemble NN66 Movement History March 3, 2008
Diablo Canyon Power Plant 2R14 Outage Safety Plan, Revision 1
Nuclear Management and Resource Council, NUMARC Guidelines for Industry Actions to
Assess Shutdown Management, December 1991
Action Requests
A0719298 A0719285 A0719294
1R22: Surveillance Testing
Procedures
STP V-3R5, Exercising Steam Supply to Auxiliary Feedwater Pump Turbine Stop Valve,
FCV-95, Revision 19
STP V-3R6, Exercising Steam Supply to Auxiliary Feedwater Pump Turbine Isolation Valves,
FCV-37 and FCV-38, Revision 10
STP V-3P5, Exercising Valves LCV-106, 107, 108, and 109 Auxiliary Feedwater Pump
Discharge, Revision 20
STP V-623, Penetration 22 and 23 Containment Isolation Valve Leak Test, Revision 7
STP V-650, Penetration 50 Containment Isolation Valve Leak Testing, Revision 11
CF3.ID13, Replacement or New Part Evaluation (RPE), AT-RPE AR and CITE, Revision 19A
STP I-1A, Routine Shift Checks required by Licenses, Revision 109
STP V-3F1, Exercising Valve FCV-495, ASW Pump 2 Crosstie Valve, Revision 23
STP M-26, Auxiliary Saltwater Flow System Monitoring, Revision 2
A-8 Attachment
STP M-9A, Diesel Generator Routine Surveillance Test, Revision76A
STP M-13F, 4kV Bus F Non-SI Auto-Transfer Test, Revision 36
STP M-16U, Slave Relay Test of Trains A and B, K605, Revision 6
2OS1: Access Controls to Radiologically Significant Areas and 2OS2: ALARA Planning
and Controls
Procedures
RCP D-200, Writing Radiation Work Permits and ALARA Planning, Revision 41
RCP D-220, Control of Access to High, Locked High, and Very High Radiation Areas,
Revision 35
RCP D-240, Radiological Posting, Revision 18
RCP D-420, Sampling and Measuring of Airborne Radioactivity, Revision 20B
RCP D-430, Plant Airborne Radioactivity Surveillance, Revision 18
RCP D-500, Routine and Job Coverage Surveys, Revision 23
RP1, Radiation Protection, Revision 4A
RP1.ID9, Radiation Work Permits, Revision 9
AWPO-002, NRC Performance Indicator: RETS/ODCM Radiological Effluent Occurrences,
Revision 9
AWPO-003, NRC Performance Indicators: Occupational Exposure Control Effectiveness,
Revision 5
Action Requests
A0666110 A0714302 A0711672 A0713281 A0713540 A0703336
A0703351 A0706806 A0081493 A0716527 A0714302 A0649226
A0711318 A0711502 A0719338 A0716506 A0716528 A0713703
A0716120 A0716272 A0716535 A0716640 A0716656
Audits and Self-Assessment
Diablo Canyon Power Plant Quality Performance Assessment Report, 3rd Period 2007
Radiation Work Permits
08-2015-00 08-0007-00 08-2140-00 08-2041-00 08-2104-00 08-2106-00
08-2001-00 08-2066-00 08-2019-00
A-9 Attachment
4OA2: Problem Identification and Resolution
Miscellaneous
Quality Verification Observation Report, January 31-February 7, 2008
Quality Verification 2R14 Short Form Assessment 080380014, February 7, 2008,
Quality Verification Department Bi-Weekly Observation Report EDMS 080030003,
February 11, 2008
Generation Nuclear Quality Verification Diablo Canyon Power Plant, Short Form
Assessment 080660004
Quality Verification, 2R14 Mid-Outage Human Performance Assessment, Short Form
Assessment 080660003
Quality Verification Department, Bi-weekly Observation Report, February 12-28, 2008,
EDMS 080030004
Quality Verification Real Time Report, February 21, 2008
Plant Performance Improvement Report, December 2007
Quality Verification, Real Time Report February 28, 2008
DCPP Observation Program Report, FileNet: 080660012, March 6, 2008
DCPP Observation Program Report, FileNet: 080730027, March 13, 2008
DCPP Observation Program Report, FileNet: 080790015, March 20, 2008
DCPP Observation Program Report, FileNet: 080860006, March 27, 2008
Section 4OA5: TI 2515/166, PWR Containment Sump
Amendment 200, License DPR-82, Pacific Gas and Electric Company, Docket 50-323, Diablo
Canyon Nuclear Power Plant, Unit 2, Amendment to Facility Operating License, Safety
Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment 199 to Facility
Operating License DPR-80, and Amendment 200 to Facility Operating License DPR-82, Pacific
Gas and Electric Company, Diablo Canyon Nuclear Power Plant, Units 1 and 2.
PG&E Letter, DCL-08-002, U.S. Nuclear Regulatory Commission, Supplemental response to
Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation
During Design Basis Accidents at Pressurized Water Reactors.
A-10 Attachment
LIST OF ACRONYMS
ALARA as low as is reasonably achievable
ASME American Society of Mechanical Engineers
CFR Code of Federal Regulation
FSAR Final Safety Analysis Report
NCV noncited violation
NDE nondestructive examination
NEI Nuclear Energy Institute
PG&E Pacific Gas and Electric
VUHP vessel upper head penetration
A-11 Attachment