ML100360782
| ML100360782 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/05/2010 |
| From: | Clark J NRC/RGN-IV/DRP/RPB-E |
| To: | Kowalewski J Entergy Operations |
| References | |
| IR-09-005 | |
| Download: ML100360782 (59) | |
See also: IR 05000382/2009005
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
February 5, 2010
Joseph Kowalewski, Vice President, Operations
Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3
17265 River Road
Killona, LA 70057-0751
SUBJECT: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED
INSPECTION REPORT 05000382/2009005
Dear Mr. Kowalewski:
On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Waterford Steam Electric Station, Unit 3. The enclosed integrated inspection
report documents the inspection findings, which were discussed on January 11, 2010, with you
and other members of your staff.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents three self-revealing findings of very low safety significance (Green). All
of these findings were determined to involve violations of NRC requirements. Additionally, a
licensee-identified violation, which was determined to be of very low safety significance, is listed
in this report. However, because of the very low safety significance and because they are
entered into your corrective action program, the NRC is treating these findings as noncited
violations, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the
violations or the significance of the noncited violations, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with
copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E.
Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement,
U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident
Inspector at the Waterford Steam Electric Station, Unit 3 facility. In addition, if you disagree with
the characterization of any finding in this report, you should provide a response within 30 days
of the date of this inspection report, with the basis for your disagreement, to the Regional
Administrator, Region IV, and the NRC Resident Inspector at Waterford Steam Electric Station,
Entergy Operations, Inc. -2-
Unit 3. The information you provide will be considered in accordance with Inspection Manual
Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letter, and its
enclosure, will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records component of NRCs document system (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the
Public Electronic Reading Room).
Sincerely,
/RA/
Jeffrey A. Clark, P.E.
Chief, Project Branch E
Division of Reactor Projects
Docket: 50-382
License: NPF-38
Enclosure:
NRC Inspection Report 05000382/2009005
w/Attachment: Supplemental Information
cc w/Enclosure:
Senior Vice President
Entergy Nuclear Operations
P. O. Box 31995
Jackson, MS 39286-1995
Senior Vice President and
Chief Operating Officer
Entergy Operations, Inc.
P. O. Box 31995
Jackson, MS 39286-1995
Vice President, Operations Support
Entergy Services, Inc.
P. O. Box 31995
Jackson, MS 39286-1995
Senior Manager, Nuclear Safety
and Licensing
Entergy Services, Inc.
P. O. Box 31995
Jackson, MS 39286-1995
Entergy Operations, Inc. -3-
Site Vice President
Waterford Steam Electric Station, Unit 3
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057-0751
Director
Nuclear Safety Assurance
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057-0751
General Manager, Plant Operations
Waterford 3 SES
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057-0751
Manager, Licensing
Entergy Operations, Inc.
17265 River Road
Killona, LA 70057-0751
Chairman
Louisiana Public Service Commission
P. O. Box 91154
Baton Rouge, LA 70821-9154
Parish President Council
St. Charles Parish
P. O. Box 302
Hahnville, LA 70057
Director, Nuclear Safety & Licensing
Entergy, Operations, Inc.
440 Hamilton Avenue
White Plains, NY 10601
Louisiana Department of Environmental
Quality, Radiological Emergency Planning
and Response Division
P. O. Box 4312
Baton Rouge, LA 70821-4312
Entergy Operations, Inc. -4-
Chief, Technological Hazards
Branch
FEMA Region VI
800 North Loop 288
Federal Regional Center
Denton, TX 76209
Entergy Operations, Inc. -5-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Chuck.Casto@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRP Deputy Director (Anton.Vegel@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Mark Haire@nrc.gov)
Resident Inspector (Dean.Overland@nrc.gov)
Branch Chief, DRP/E (Jeff.Clark@nrc.gov)
Senior Project Engineer, DRP/E (Ray.Azua@nrc.gov)
WAT Site Secretary (Linda.Dufrene@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
Regional State Liaison Officer (Bill.Maier@nrc.gov)
NSIR/DPR/EP (Eric.Schrader@nrc.gov)
NSIR/DPR/EP (Steve.LaVie@nrc.gov)
Inspection Reports/MidCycle and EOC Letters to the following:
ROPreports
Only inspection reports to the following:
DRS/TSB STA (Dale.Powers@nrc.gov)
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)
R:\File located: R:\REACTORS\_WAT\2009\WAT 2009004 RP-MSH.doc ML 100360782
ADAMS: No : Yes : SUNSI Review Complete Reviewer Initials: JAC
- Publicly Available : Non-Sensitive
Non-publicly Available Sensitive
RIV:SRI:DRP/E RI:DRP/E SPE:DRP/E C:DRS/EB1 C:DRS/EB2
MSHaire DHOverland RVAzua TFarnholtz NFOKeefe
/RA/E-mailed /RA/E-mailed /RA/ /RA/ /RA/
01/25/2010 01/25/2010 01/18/2010 01/19/2010 1/19/2010
C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/E
SMGarchow MPShannon GEWerner JAClark
/RA/ /RA/ /RA/ /RA/
01/25/2010 01/25/2010 1/19/2010 02/5/2010
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000382
License: NFP-38
Report: 05000382/2009005
Licensee: Entergy Operations, Inc.
Facility: Waterford Steam Electric Station, Unit 3
Location: Hwy. 18
Killona, LA
Dates: October 8 through December 31, 2009
Inspectors: M. Haire, Senior Resident Inspector
D. Overland, Resident Inspector
S. Anderson, General Engineer
R. Azua, Senior Project Engineer
M. Bloodgood, Senior Reactor Inspector
T. Buchanan, Reactor Inspector
P. Elkmann, Senior Emergency Preparedness Inspector
L. Ricketson, P.E., Senior Health Physicist
N. Greene, Health Physicist
Approved By: Jeff Clark, P.E., Chief, Project Branch E
Division of Reactor Projects
-1- ENCLOSURE
SUMMARY OF FINDINGS
IR 05000382/2009005; October 8, 2009 through December 31, 2009; Waterford Steam Electric
Station, Unit 3, Identification and Resolution of Problems, Access Control to Radiologically
Significant Areas
The report covered a 3-month period of inspection by resident inspectors and announced
baseline inspections by regional based inspectors. Three Green noncited violations of NRC
requirements were identified. The significance of most findings is indicated by their color
(Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance
Determination Process. Findings for which the significance determination process does not
apply may be Green or be assigned a severity level after NRC management review. The NRC's
program for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Initiating Events
- Green. A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, was identified for the licensees failure to promptly correct a
condition adverse to quality. Specifically, the licensee did not promptly correct
reactor coolant pump vapor seal leakage that resulted in boric acid accumulation
on the component cooling water heat exchanger and cover areas of three reactor
coolant pumps. Corrective actions for this condition were implemented during
Refueling Outage 15, but these corrective actions failed to correct the condition
and the vapor seal leakage continued through operating Cycle 16. This resulted
in some additional boric acid corrosion and degradation to reactor coolant pump
covers and carbon steel component cooling water flanges. The licensee
implemented a design modification to correct the condition and documented the
condition in Condition Report CR-WF3-2009-5501.
The licensees failure to promptly correct a condition adverse to quality is more
than minor because it is associated with the equipment performance attribute of
the Initiating Events Cornerstone and affects the cornerstone objective to limit the
likelihood of those events that upset plant stability. The finding has very low
safety significance because, although the finding contributes to the likelihood of a
reactor trip, mitigation equipment was still available. This finding had a
crosscutting aspect in the area of human performance associated with work
control in that the licensee did not effectively plan for the resources necessary to
implement the postmaintenance testing associated with the corrective actions
implemented during Refueling Outage 15, and therefore failed to discover that
those corrective actions were inadequate to correct the condition H.3(a)
(Section 4OA2).
- Green. A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B,
Criterion V, was identified for the licensees failure to prescribe an activity
-2- ENCLOSURE
affecting quality by documented instructions, procedures, or drawings
appropriate to the circumstance. Specifically, for all reactor coolant pump heat
exchanger to pump cover bolted connection gasket replacements between the
refueling outage of 1986 (Refueling Outage 1) and the refueling outage of 2009
(Refueling Outage 16), the licensee prescribed the wrong gasket material, gasket
size, and fastener preload because they had failed to incorporate a design
change implemented during Refueling Outage 1 into their instructions,
procedures, or drawings. Station Modification Package SMP-1427, an
engineering change implemented during Refueling Outage 1 in response to
industry operating experience, called for a thicker gasket, different gasket
material, and an increased bolt preload in order to increase gasket compression
and reduce the probability of leakage. As a consequence of failing to incorporate
Station Modification Package SMP-1427 changes into procedures, all heat
exchanger gasket replacements since Refueling Outage 1, four gasket
replacements in total, have utilized thinner gaskets with less than the vendor
recommended compression. The licensee documented this condition in
Condition Report CR-WF3-2009-5501.
The licensees failure to prescribe appropriate gasket replacement requirements
is more than minor because it is associated with the equipment performance
attribute of the Initiating Events Cornerstone and affects the cornerstone
objective to limit the likelihood of those events that upset plant stability. The
finding has very low safety significance because, although the finding contributes
to the likelihood of a reactor trip, mitigation equipment is still available. This
finding had a crosscutting aspect in the area of problem identification and
resolution associated with operating experience in that the licensee did not
institutionalize operating experience through changes to the station procedures
P.2(b) (Section 4OA2).
Cornerstone: Occupational Radiation Safety
- Green. The inspectors reviewed a self-revealing noncited violation of Technical
Specification 6.8.1 which resulted from a worker failing to follow radiation
protection procedures. A contract radiation worker went to work near steam
generator 1 rather than the area for which he/she was briefed and received
multiple electronic dosimeter dose rate alarms, but did not leave the area until
receiving a continuous dose alarm. In response, the licensee investigated the
occurrence and restricted the individuals access. Additional actions were being
evaluated. This issue was entered into the licensees corrective action program
as Condition Reports CR-WF3-2009-05648 and WF3-2009-06852.
This finding is greater than minor because it involved the program attribute of
exposure control and affected the cornerstone objective in that the failure of the
worker to follow procedural guidance resulted in the worker being
unknowledgeable to the dose rates in all areas entered. The inspectors used the
Occupational Radiation Safety Significance Determination Process and
determined the finding had very low safety significance because it was not:
(1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure,
-3- ENCLOSURE
(3) a substantial potential for overexposure, or (4) an inability to assess dose.
The finding had a crosscutting aspect in the area of human performance, work
practices component, because the worker failed to use human error prevention
techniques such as self and peer checking H.4(a) (Section 2OS1).
B. Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee, has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation and corrective
action tracking numbers (condition report numbers) are listed in Section 4OA7.
-4- ENCLOSURE
REPORT DETAILS
Summary of Plant Status
The plant began the inspection period on October 8, 2009, at 100 percent power and remained
at approximately 100 percent power until October 19, 2009, when the plant was shutdown in
preparation of the licensees planned Refueling Outage 16. The plant remained shutdown until
December 1, 2009, when the reactor was placed back online and the licensee began increasing
power. On December 6, 2009, the plant reached 100 percent power and continued to operate
at this level for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather Protection (71111.01)
.1 Readiness to Cope with External Flooding
a. Inspection Scope
The inspectors evaluated the design, material condition, and procedures for coping with
the design basis probable maximum flood. The evaluation included a review to check
for deviations from the descriptions provided in the Updated Final Safety Analysis Report
for features intended to mitigate the potential for flooding from external factors. As part
of this evaluation, the inspectors checked for obstructions that could prevent draining,
checked that the roofs did not contain obvious loose items that could clog drains in the
event of heavy precipitation, and determined that barriers required to mitigate the flood
were in place and operable. Additionally, the inspectors performed a walkdown of the
protected area to identify any modification to the site that would inhibit site drainage
during a probable maximum precipitation event or allow water ingress past a barrier.
The inspectors also reviewed the abnormal operating procedure for mitigating the design
basis flood to ensure it could be implemented as written. Specific documents reviewed
during this inspection are listed in the attachment.
These activities constitute completion of one external flooding sample as defined in
Inspection Procedure 71111.01-05.
b. Findings
No findings of significance were identified.
-5- ENCLOSURE
1R04 Equipment Alignments (71111.04)
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- October 8, 2009, Essential chiller train B
- October 14, 2009, Low pressure safety injection train B
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could affect the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Final Safety Analysis Report, technical specification
requirements, administrative technical specifications, outstanding work orders, 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 operable. The inspectors examined the material condition of the
components and observed operating parameters of equipment to verify that there were
no obvious deficiencies. The inspectors also verified that the licensee had properly
identified and resolved equipment alignment problems that could cause initiating events
or impact the capability of mitigating systems or barriers and entered them into the
corrective action program with the appropriate significance characterization. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of two partial system walkdown samples as
defined in Inspection Procedure 71111.04-05.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
.1 Quarterly Fire Inspection Tours
a. Inspection Scope
The inspectors conducted fire protection walkdowns that were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- November 2, 2009, Fuel handling building
- November 10, 2009, Fire zones RAB 37, 38, and 39
-6- ENCLOSURE
- November 28, 2009, Reactor containment building
- December 15, 2009, Battery and switchgear areas
The inspectors reviewed areas to assess if licensee personnel had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant; effectively maintained fire detection and suppression capability; maintained
passive fire protection features in good material condition; and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire
protection equipment, systems, or features, in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to affect equipment that could initiate or mitigate a plant
transient, or their impact on the plants ability to respond to a security event. Using the
documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed, that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four quarterly fire-protection inspection samples
as defined in Inspection Procedure 71111.05-05.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures (71111.06)
a. Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report, the flooding analysis,
and plant procedures to assess susceptibilities involving internal flooding; reviewed the
corrective action program to determine if licensee personnel identified and corrected
flooding problems; and verified that operator actions for coping with flooding can
reasonably achieve the desired outcomes. The inspectors also walked down the area
listed below to verify the adequacy of equipment seals located below the flood line, floor
and wall penetration seals, watertight door seals, common drain lines and sumps, sump
pumps, level alarms, and control circuits, and temporary or removable flood barriers.
- October 14, 2009, Reactor Auxiliary Building -35 foot elevation
This inspection procedure also requires an annual review of risk-significant cables
located in underground bunkers/manholes. Waterford Steam Electric Station, Unit 3, by
design, does not have any safety-related cables that are located in underground
bunkers/manholes; however, there are 17 manholes in which cables associated with
maintenance rule related equipment were located. The inspectors inspected
-7- ENCLOSURE
Manholes M301-NA, M346-NB, and M347-NA and determined that all three contained
maintenance rule related cables submerged in water. The submerged cables did not
show visible deterioration. The licensee has documented this condition in Condition
Report CR-WF3-2009-3925, and is developing a cable monitoring program. Specific
documents reviewed during this inspection are listed in the attachment.
This activity constitutes completion of two flood protection measures inspection samples
as defined in Inspection Procedure 71111.06-05.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07)
a. Inspection Scope
The inspectors reviewed licensee programs, verified performance against industry
standards, and reviewed critical operating parameters and maintenance records for the
steam generators. The inspectors verified that performance tests were satisfactorily
conducted for heat exchangers/heat sinks and reviewed for problems or errors; the
licensee utilized the periodic maintenance method outlined in EPRI Report NP 7552,
Heat Exchanger Performance Monitoring Guidelines; the licensee properly utilized
biofouling controls; the licensees heat exchanger inspections adequately assessed the
state of cleanliness of their tubes; and the heat exchanger was correctly categorized
under 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants. Specific documents reviewed during this inspection are listed in
the attachment.
These activities constitute completion of one heat sink inspection sample as defined in
Inspection Procedure 71111.07-05.
b. Findings
No findings of significance were identified.
1R08 In-service Inspection Activities (71111.08)
Completion of Sections .1 through .5, below, constitutes completion of one sample as
defined in Inspection Procedure 71111.05-05.
.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water
Reactor Vessel Upper Head Penetration Inspections, and Boric Acid Corrosion Control
(71111.08-02.01)
a. Inspection Scope
The inspectors reviewed two types of nondestructive examination activities and two
welds on the reactor coolant system pressure boundary.
-8- ENCLOSURE
The inspectors directly observed the following nondestructive examinations:
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Safety Injection RCS 2A Safety Injection Nozzle Ultrasonic Testing
System (Weld No.12-009)
Reactor Coolant RCS 1A Cold leg Suction Line Ultrasonic Testing
System (Weld No.07-005)
Reactor Coolant RCS 2A Cold Leg Suction Line Ultrasonic Testing
System (Weld No.11-002)
Reactor Coolant RCS 2A Cold Leg Suction Line Visual Inspection VT-1&2
System (Weld No.11-002)
The inspectors reviewed records for the following nondestructive examinations:
SYSTEM IDENTIFICATION EXAMINATION TYPE
Safety Injection RCS 2A Safety Injection Nozzle Ultrasonic Testing
System (Weld No.12-009)
Reactor Coolant RCS 1A Cold leg Suction Line Ultrasonic Testing
System (Weld No.07-005)
Reactor Coolant RCS 2A Cold Leg Suction Line Ultrasonic Testing
System (Weld No.11-002)
Reactor Coolant RCS 2A Cold Leg Suction Line Visual Inspection VT-1&2
System (Weld No.11-002)
Reactor Coolant RCS 12 Hot Leg Surge Line Ultrasonic Testing
System (Weld No.15-009)
During the review and observation of each examination, the inspectors verified that
activities were performed in accordance with the ASME Code requirements and
applicable procedures. The inspectors also verified the qualifications of all
nondestructive examination technicians performing the inspections were current.
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 welding process were identified, recorded in the procedure
qualification record, and formed the basis for qualification of the welding procedure
specifications. Specific documents reviewed during this inspection are listed in the
attachment.
These actions constitute completion of the requirements for Section 02.01.
-9- ENCLOSURE
b. Findings
No findings of significance were identified.
.2 Vessel Upper Head Penetration Inspection Activities (71111.08-02.02)
a. Inspection Scope
The inspectors reviewed the results of licensee personnels visual inspection of
pressure-retaining components above the reactor pressure vessel head to verify that
there was no evidence of leaks or boron deposits on the surface of the reactor pressure
vessel head or related insulation. The inspectors verified that the personnel performing
the visual inspection were certified as Level II and Level III VT-2 examiners. Specific
documents reviewed during this inspection are listed in the attachment.
The inspectors also reviewed the results of licensee personnels volumetric inspection of
pressure-retaining components above the reactor pressure vessel head to verify that
there were no flaws in the welds associated with these penetrations. The inspectors
observed data acquisition and analysis of one penetration. The inspector verified that
the personnel performing the inspections were current in their certification as Level II or
Level III ultrasonic testing examiners. Specific documents reviewed during this
inspection are listed in the attachment.
These actions constitute completion of the requirements for Section 02.02.
b. Findings
No findings of significance were identified.
.3 Boric Acid Corrosion Control Inspection Activities (71111.08-02.03)
a. Inspection Scope
The inspectors evaluated the implementation of the licensees boric acid corrosion
control program for monitoring degradation of those systems that could be adversely
affected by boric acid corrosion. The inspectors reviewed the documentation associated
with the licensees boric acid corrosion control walkdown as specified in Procedure
NOECP-107, Boric Acid Corrosion Control Program (BACCP), Revision 1. The
inspectors also reviewed the visual records of the components and equipment. The
inspectors verified that the visual inspections emphasized locations where boric acid
leaks could cause degradation of safety-significant components. The inspectors also
verified that the engineering evaluations for those components where boric acid was
identified gave assurance that the ASME Code wall thickness limits were properly
maintained. The inspectors confirmed that the corrective actions performed for evidence
of boric acid leaks were consistent with requirements of the ASME Code. Specific
documents reviewed during this inspection are listed in the attachment.
- 10 - ENCLOSURE
These actions constitute completion of the requirements for Section 02.03.
b. Findings
No findings of significance were identified.
.4 Steam Generator Tube Inspection Activities (71111.08-02.04)
a. Inspection Scope
The inspectors assessed the in-situ screening criteria to assure consistency between
assumed nondestructive examination flaw sizing accuracy and data from the Electric
Power Research Institute (EPRI) examination technique specification sheets. No
conditions were identified that warranted in-situ pressure testing. The inspectors did,
however, review the licensees Steam Generator Degradation Assessment and Repair
Criteria for RF15, dated April 2008, and compared the in-situ test screening parameters
to the guidelines contained in the EPRI document In Situ Pressure Test Guidelines,
Revision 2. This review determined that the screening parameters were consistent with
the EPRI guidelines.
In addition, the inspectors reviewed both the licensee site-validated and qualified
acquisition and analysis technique sheets used during this refueling outage and the
qualifying EPRI examination technique specification sheets to verify that the essential
variables regarding flaw sizing accuracy, tubing, equipment, technique, and analysis had
been identified and qualified through demonstration. The inspectors reviewed
acquisition technique and analysis technique data sheets.
The inspection procedure specified comparing the estimated size and number of tube
flaws detected during the current outage against the previous outage operational
assessment predictions to assess the licensees prediction capability. The inspectors
compared the previous outage operational assessment predictions with the flaws
identified during the current steam generator tube inspection effort. The number of
identified indications fell below the range of prediction but was consistent with historical
predictions.
The inspection procedure specified confirmation that the steam generator tube eddy
current test scope and expansion criteria meet technical specification requirements,
EPRI guidelines, and commitments made to the NRC. The inspectors compared the
recommended test scope to the actual test scope and found that the licensee had
accounted for all known flaws and had, as a minimum, established a test scope that met
technical specification requirements, EPRI guidelines, and commitments made to the
NRC. The scope of the licensees eddy current examinations of tubes in both steam
generators included:
- 100 percent bobbin examination full length of tubing
- 100 percent hot leg top of tube sheet
- 100 percent Rows 1 and 2 u-bend rotating pancake coil
- 100 percent dented tube supports at egg crates greater than 2 Volts
- 11 - ENCLOSURE
- 20 percent dented diagonal bar and vertical strap greater than 2 Volts
- 20 percent free span dings greater than 5 Volts
- Cold leg top of tube sheet periphery exam for loose parts
The inspection procedure specified that, if new degradation mechanisms were identified,
the licensee would verify the analysis fully enveloped the problem of the extended
conditions including operating concerns and that appropriate corrective actions were
taken before plant startup. No new degradation mechanisms were identified.
The inspection procedure required confirmation that the licensee inspected all areas of
potential degradation, especially areas that were known to represent potential eddy
current test challenges (e.g., top-of-tubesheet, tube support plates, and U-bends). The
inspectors confirmed that all known areas of potential degradation were included in the
scope of inspection and were being inspected.
The inspection procedure further required verification that repair processes being used
were approved in the technical specifications. The inspectors confirmed that the repair
processes being used were consistent with the technical specifications requirements.
The inspection procedure also required confirmation of adherence to the technical
specification plugging limit, unless alternate repair criteria have been approved. The
inspection procedure further requires determination whether depth sizing repair criteria
were being applied for indications other than wear or axial primary water stress corrosion
cracking in dented tube support plate intersections. The inspectors determined that the
technical specification plugging limits were being adhered to (i.e., 40 percent maximum
through-wall indication).
If steam generator leakage greater than 3 gallons per day was identified during
operations or during post shutdown visual inspections of the tubesheet face, the
inspection procedure required verification that the licensee had identified a reasonable
cause based on inspection results and that corrective actions were taken or planned to
address the cause for the leakage. The inspectors did not conduct any assessment
because this condition did not exist.
The inspection procedure required confirmation that the eddy current test probes and
equipment were qualified for the expected types of tube degradation and an assessment
of the site-specific qualification of one or more techniques. The inspectors observed
portions of the eddy current tests. During these examinations, the inspectors verified
that: (1) the probes appropriate for identifying the expected types of indications were
being used, (2) probe position location verification was performed, (3) calibration
requirements were adhered, and (4) probe travel speed was in accordance with
procedural requirements. The inspectors performed a review of site-specific
qualifications of the techniques being used.
These actions constitute completion of the requirements of Section 02.04.
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b. Findings
No findings of significance were identified.
.5 Identification and Resolution of Problems (71111.08-02.05)
a. Inspection scope
The inspectors reviewed 27 condition reports which dealt with inservice inspection
activities and found the corrective actions were appropriate. The specific condition
reports 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. Specific documents reviewed during this inspection are listed in
the attachment.
These actions constitute completion of the requirements of Section 02.05.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11)
a. Inspection Scope
On November 24, 2009, the inspectors observed a crew of licensed operators in the
plants simulator to verify that operator performance was adequate, evaluators were
identifying and documenting crew performance problems, and training was being
conducted in accordance with licensee procedures. The inspectors evaluated the
following areas:
- Licensed operator performance
- Crews clarity and formality of communications
- Crews ability to take timely actions in the conservative direction
- Crews prioritization, interpretation, and verification of annunciator alarms
- Control board manipulations
- Oversight and direction from supervisors
The inspectors compared the crews performance in these areas to pre-established
operator action expectations. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of one quarterly licensed-operator requalification
program sample as defined in Inspection Procedure 71111.11.
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b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12Q)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems:
- November 20, 2009, Effects of voiding on the functionality of low pressure safety
injection system
- December 8, 2009, Effects of excessive leakage on functionality of containment
isolation valves
The inspectors reviewed events such as 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)
- Characterizing system reliability issues for performance
- Charging unavailability for performance
- Trending key parameters for condition monitoring
- Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)
- Verifying appropriate performance criteria for structures, systems, and
components classified as having an adequate demonstration of performance
through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as
requiring the establishment of appropriate and adequate goals and corrective
actions for systems classified as not having adequate performance, as described
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization. Specific documents reviewed during this inspection are
listed in the attachment.
- 14 - ENCLOSURE
These activities constitute completion of two quarterly maintenance effectiveness
samples as defined in Inspection Procedure 71111.12-05.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed licensee personnel's evaluation and management of plant risk
for the maintenance and emergent work activities affecting risk-significant and
safety-related equipment listed below to verify that the appropriate risk assessments
were performed prior to removing equipment for work:
- October 24, 2009, Scheduled plant refuel outage with reactor coolant system
water level reduced to approximately 19 feet to support reactor vessel head
removal during mode 6 operations
- November 30, 2009, Scheduled activity to take the reactor coolant system solid
and draw a bubble in the pressurizer following the refueling outage
- December 13, 2009, Scheduled plant protection system channel B functional test
The inspectors selected these activities based on potential risk significance relative to
the reactor safety cornerstones. As applicable for each activity, the inspectors verified
that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)
and that the assessments were accurate and complete. When licensee personnel
performed emergent work, the inspectors verified that the licensee personnel promptly
assessed and managed plant risk. The inspectors reviewed the scope of maintenance
work, discussed the results of the assessment with the licensee's probabilistic risk
analyst or shift technical advisor, and verified plant conditions were consistent with the
risk assessment. The inspectors also reviewed the technical specification requirements
and inspected portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of three maintenance risk assessments and
emergent work control inspection samples as defined in Inspection
Procedure 71111.13-05.
b. Findings
No findings of significance were identified.
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1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- October 29, 2009, Log power nuclear instrument channel B
- November 16, 2009, Station battery train B total allowable resistance
- November 19, 2009, Broken in-core nuclear instrumentation E-13
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that technical specification operability was
properly justified and the subject component or system remained available such that no
unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the technical specifications and Updated
Safety Analysis Report to the licensees evaluations, to determine whether the
components or systems were operable. Where compensatory measures were required
to maintain operability, the inspectors determined whether the measures in place would
function as intended and were properly controlled. The inspectors determined, where
appropriate, compliance with bounding limitations associated with the evaluations.
Additionally, the inspectors also reviewed a sampling of corrective action documents to
verify that the licensee was identifying and correcting any deficiencies associated with
operability evaluations. Specific documents reviewed during this inspection are listed in
the attachment.
These activities constitute completion of three operability evaluations inspection samples
as defined in Inspection Procedure 71111.15-05.
b. Findings
No findings of significance were identified.
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:
- November 9, 2009, S6X (41 second load block relay for emergency diesel
generator B sequencer) loose terminal adjustments retested during Operating
Procedure OP-903-116
- November 10, 2009, Removal, inspection, stroke test, and re-installment 3 plus
a safety injection sump outlet header B check valve SI-604B
- November 17, 2009, Replacement of station battery 3-AB-S due to end of useful
life
- 16 - ENCLOSURE
- November 19, 2009, Adjustment to closing force for reactor coolant loop 1
shutdown cooling outside containment isolation valve SI-407B to correct
excessive leakage
(Operating Procedure OP-903-046)
- December 7, 2009, Replacement of station battery 3-A-S due to end of useful life
- December 9, 2009, Change setpoints and adjust limit stop setting on containment
vacuum relief differential pressure switch CVRIDPIS5220A
The inspectors selected these activities based upon the structure, system, or
component's ability to affect risk. The inspectors evaluated these activities for the
following (as applicable):
- The effect of testing on the plant had been adequately addressed; testing was
adequate for the maintenance performed
- Acceptance criteria were clear and demonstrated operational readiness; test
instrumentation was appropriate
The inspectors evaluated the activities against the technical specifications, the Updated
Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and
various NRC generic communications to ensure that the test results adequately ensured
that the equipment met the licensing basis and design requirements. In addition, the
inspectors reviewed corrective action documents associated with postmaintenance tests
to determine whether the licensee was identifying problems and entering them in the
corrective action program and that the problems were being corrected commensurate
with their importance to safety. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of seven postmaintenance testing inspection
samples as defined in Inspection Procedure 71111.19-05.
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 3
refueling outage, conducted October 19, 2009, through December 4, 2009, to confirm
that licensee personnel had appropriately considered risk, industry experience, and
previous site-specific problems in developing and implementing a plan that assured
maintenance of defense-in-depth. During the refueling outage, the inspectors observed
portions of the shutdown and cooldown processes and monitored licensee controls over
the outage activities listed below.
- 17 - ENCLOSURE
- Configuration management, including maintenance of defense-in-depth, is
commensurate with the outage safety plan for key safety functions and
compliance with the applicable technical specifications when taking equipment
out of service
- Clearance activities, including confirmation that tags were properly hung and
equipment appropriately configured to safely support the work or testing
- Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error
- Status and configuration of electrical systems to ensure that technical
specifications and outage safety-plan requirements were met, and controls over
switchyard activities
- Monitoring of decay heat removal processes, systems, and components
- Verification that outage work was not impacting the ability of the operators to
operate the spent fuel pool cooling system
- Reactor water inventory controls, including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss
- Controls over activities that could affect reactivity
- Refueling activities, including fuel handling and sipping to detect fuel assembly
leakage
- Startup and ascension to full power operation, tracking of startup prerequisites,
walkdown of the primary containment to verify that debris had not been left which
could block emergency core cooling system suction strainers, and reactor
physics testing
- Licensee identification and resolution of problems related to refueling outage
activities
- Review of Operating Experience Smart Sample FY2007-03, crane and heavy lift
inspection
- Review of Operating Experience Smart Sample FY2007-01, related to
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one refueling outage and other outage
inspection sample as defined in Inspection Procedure 71111.20-05.
- 18 - ENCLOSURE
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report, procedure
requirements, and technical specifications to ensure that the two surveillance activities
listed below demonstrated that the systems, structures, and/or components tested were
capable of performing their intended safety functions. The inspectors either witnessed or
reviewed test data to verify that the significant surveillance test attributes were adequate
to address the following:
- Preconditioning
- Evaluation of testing impact on the plant
- Acceptance criteria
- Test equipment
- Procedures
- Jumper/lifted lead controls
- Test data
- Testing frequency and method demonstrated technical specification operability
- Test equipment removal
- Restoration of plant systems
- Fulfillment of ASME Code requirements
- Updating of performance indicator data
- Engineering evaluations, root causes, and bases for returning tested systems,
structures, and components not meeting the test acceptance criteria were correct
- Reference setting data
- Annunciators and alarms setpoints
The inspectors also verified that licensee personnel identified and implemented any
needed corrective actions associated with the surveillance testing.
- 19 - ENCLOSURE
- November 9, 2009, Train B integrated emergency diesel generator/engineering
safety features test (Operating Procedure OP-903-116)
- November 18, 2009, Leak test on reactor coolant loop 1 shutdown cooling
outside containment isolation valve SI-407B
- December 14, 2009, Annulus negative pressure valves ANP-101 and ANP-102
surveillance test (Operating Procedure OP-903-120)
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of three surveillance testing inspection samples as
defined in Inspection Procedure 71111.22-05.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
.1 Inoffice Review, Revision 23
a. Inspection Scope
The inspector performed an in-office review of Emergency Plan Implementing Procedure
EP-001-001, Revision 23, Recognition and Classification of Emergency Conditions,
submitted August 19, 2009. This revision
- Added information to emergency action level CU1 to clarify that steam generator
leakage is considered to be identified reactor coolant leakage
- Added information to emergency action level RCB2 to clarify that manual
initiation of emergency core cooling systems to compensate for a steam
generator tube leak/rupture meets the intent of the emergency action level
- Added information to emergency action level HU6 to clarify that entry conditions
are not met until hurricane force winds are projected for the site occurring in less
than or equal to twelve hours
This revision was compared to its previous revision, to the criteria of NUREG-0654,
Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1, to Nuclear Energy
Institute Report 99-01, Emergency Action Level Methodology, Revision 5, and to the
standards in 10 CFR 50.47(b) to determine if the revision adequately implemented the
requirements of 10 CFR 50.54(q). This review was not documented in a safety
evaluation report and did not constitute approval of licensee-generated changes;
therefore, this revision is subject to future inspection.
- 20 - ENCLOSURE
These activities constitute completion of one sample as defined in Inspection
Procedure 71114.04-05.
b. Findings
No findings of significance were identified.
.2 Inoffice Review, Revision 24
a. Inspection Scope
The inspector performed an in-office review of the Waterford Steam Electric Station
Emergency Plan, Revision 38, and Emergency Plan Implementing
Procedure EP-001-001, Recognition and Classification of Emergency Conditions,
Revision 24, submitted October 23, 2009. These revisions
- Deleted emergency action level CU4, fuel clad degradation
- Changed the initiating conditions of Emergency Action Level SU9, Fuel Clad
Degradation, from greater than 1.0 µCi/g DEI or greater than 100 over E-Bar
µCi/g, to greater than 60 µCi/g DEI or greater than 1.0 µCi/g DEI for more than a
continuous 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period or greater than 100 over E-Bar µCi/g
- Removed fuel clad degradation from the list of Unusual Event conditions on the
Emergency Plan Table 4-1, Summary of Initiating Conditions, and the index of
initiating conditions for cold shutdown conditions in Procedure EP-001-001
The NRC approved the licensees changes to emergency action levels CU4 and SU9 in
a Safety Evaluation Report and letter dated October 13, 2009 (Agency Document and
Management System Accession Number ML092600263).
These revisions were compared to the Safety Evaluation Report dated
October 13, 2009, to determine if the revisions adequately implemented the
requirements of 10 CFR 50.54(q).
These activities constitute completion of two samples as defined in Inspection
Procedure 71114.04-05.
b. Findings
No findings of significance were identified.
- 21 - ENCLOSURE
1EP6 Drill Evaluation (71114.06)
.1 Training Observations
a. Inspection Scope
The inspectors observed a training evolution for licensed operators on
December 21, 2009, which required emergency plan implementation by a licensee
operations crew. This evolution was planned to be evaluated and included in
performance indicator data regarding drill and exercise performance. The inspectors
reviewed the event scenarios and crew briefings for two scenarios. The inspectors
observed event classification and notification activities performed by the crew. The
inspectors also attended the postevolution 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.
These activities constitute completion of one sample as defined in Inspection
Procedure 71114.06-05.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational and Public Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
a. Inspection Scope
This area was inspected to assess licensee personnels 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, the technical specifications, and the licensees
procedures required by technical specifications as criteria for determining compliance.
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 radiation, high radiation, or
airborne radioactivity areas
- 22 - ENCLOSURE
- Radiation work permits, procedures, engineering controls, and air sampler
locations
- Conformity of electronic personal dosimeter alarm set points with survey
indications and plant policy; workers knowledge of required actions when their
electronic personnel dosimeter noticeably malfunctions or alarms
- Barrier integrity and performance of engineering controls in airborne radioactivity
areas
- 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
- Adequacy of radiological controls, such as required surveys, radiation protection
job coverage, and contamination control during job performance
- Dosimetry placement in high radiation work areas with significant dose rate
gradients
- Changes in licensee procedural controls of high dose rate - high radiation areas
and very high radiation areas
- Controls for special areas that have the potential to become very high radiation
areas during certain plant operations
- Posting and locking of entrances to all accessible high dose rate - high radiation
areas and very high radiation areas
- Radiation worker and radiation protection technician performance with respect to
radiation protection work requirements
Either because the conditions did not exist or an event had not occurred, no
opportunities were available to review the following items:
- Adequacy of the licensees internal dose assessment for any actual internal
exposure greater than 50 millirem committed effective dose equivalent
- 23 - ENCLOSURE
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of 21 of the required 21 samples as defined in
Inspection Procedure 71121.01-05.
b. Findings
Introduction. The inspectors reviewed a Green self-revealing, noncited violation of
Technical Specification 6.8.1 which resulted from a worker failing to follow radiation
protection procedures.
Description. On November 17, 2009, a contract radiation worker went to work near
steam generator 1 and received multiple electronic dosimeter dose rate alarms, but did
not leave the area until receiving a continuous dose alarm. In response, the licensee
investigated and found the worker indicated to radiation protection access control
personnel he would be going to the D-ring to work. The radiation protection technician
providing the radiological briefing showed the worker a map of reactor coolant pump 1A
and asked if that was where individual would be working. The worker acknowledged it
was, and the radiation protection technician used the survey map associated with
Radiation Work Permit 618, Task 1, Remove/Replace Insulation in the Reactor
Containment Building, to brief the worker on the radiological conditions. The worker
then signed onto Radiation Work Permit 618, Task 1, which provided a dose alarm
setpoint of 50 millirem and dose rate setpoint of 350 millirem per hour, and went to work
near steam generator 1, where dose rates were higher than the area for which the
worker was briefed. The licensee determined the worker entered a maximum dose rate
of 763 millirem per hour and received a dose of 50.8 millirem. Radiation protection
representatives stated the appropriate radiation work permit for the work area was
Radiation Work Permit 618, Task 2. Through examination of the electronic dosimeter
histogram, the licensee verified the worker received multiple dose rate alarms. The
worker mistakenly thought the dose rate alarms were generated by the workers
powered air purifying respirator signaling low air flow. Additional corrective actions were
being considered at the time of the inspection.
Analysis. The failure to follow radiation protection procedural requirements for entry into
the radiological controlled area was a performance deficiency. This finding is greater
than minor because it involved the program attribute of exposure control and affected
the cornerstone objective in that the failure of the worker to follow procedural guidance
resulted in the worker being unknowledgeable of the dose rates in all areas entered.
The inspectors used the Occupational Radiation Safety Significance Determination
Process and determined the finding had very low safety significance because it was not:
(1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure,
(3) a substantial potential for overexposure, or (4) an inability to assess dose. The
finding had a crosscutting aspect in the area of human performance, work practices
component, because the worker failed to use human error prevention techniques such
as self and peer checking H.4.a].
- 24 - ENCLOSURE
Enforcement. Technical Specification 6.8.1 requires written procedures be established,
implemented, and maintained covering the applicable procedures recommended in
Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Appendix A lists
procedures for access control to radiation areas. Procedure EN-RP-100, Radworker
Expectations, Revision 3, Section 5.3[9], requires the radiation work permit to be read,
understood, and obeyed as a condition of radiologically controlled area access.
Section 5.4[3](h) requires the worker know where to properly perform his/her task.
Section 5.3[17] requires the worker be briefed and sign on the appropriate radiation work
permit. Section 5.3[11] requires the worker know the radiological conditions in the work
area. The contract worker violated these requirements when the worker did not know
where to perform his/her task, did not sign the appropriate radiation work permit and
task, and did not know the radiological conditions in the work area as evidenced by the
multiple electronic dosimeter dose rate alarms. Because this failure to follow radiation
protection procedural guidance when entering the radiological controlled area was of
very low safety significance and has been entered into the licensees corrective action
program in Condition Reports WF3-2009-05648 and WF3-2009-06852, this violation is
being treated as an noncited violation, consistent with Section VI.A of the NRC
Enforcement Policy: NCV 05000382/2009005-01; Failure to Follow Radiation
Protection Procedural Requirements.
2OS2 ALARA Planning and Controls (71121.02)
a. Inspection Scope
The inspectors assessed licensee personnels performance with respect to maintaining
individual and collective radiation exposures as low as is reasonably achievable. 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 interviewed licensee personnel and reviewed the following:
- Integration of ALARA requirements into work procedure and radiation work
permit (or radiation exposure permit) documents
- Shielding requests and dose/benefit analyses
- Use of engineering controls to achieve dose reductions and dose reduction
benefits afforded by shielding
- Workers use of the low dose waiting areas
- Radiation worker and radiation protection technician performance during work
activities in radiation areas, airborne radioactivity areas, or high radiation areas
- Corrective action documents related to the ALARA program and follow-up
activities, such as initial problem identification, characterization, and tracking
Specific documents reviewed during this inspection are listed in the attachment.
- 25 - ENCLOSURE
These activities constitute completion of two of the required 15 samples and four of the
optional samples as defined in Inspection Procedure 71121.02-05.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the third
quarter 2009 performance indicators for any obvious inconsistencies prior to its public
release in accordance with Inspection Manual Chapter 0608, Performance Indicator
Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b. Findings
No findings of significance were identified.
.2 Reactor Coolant System Specific Activity
a. Inspection Scope
The inspectors sampled licensee submittals for the reactor coolant system specific
activity performance indicator for the period from the third quarter 2008 through the third
quarter 2009. To determine the accuracy of the performance indicator data reported
during those periods, the inspectors used definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5.
The inspectors reviewed the licensees reactor coolant system chemistry samples,
technical specification requirements, issue reports, event reports, and NRC integrated
inspection reports for the period of the third quarter 2008 through the third quarter 2009
to validate the accuracy of the submittals. The inspectors also reviewed the licensees
issue report database to determine if any problems had been identified with the
performance indicator data collected or transmitted for this indicator and none were
identified. In addition to record reviews, the inspectors observed a chemistry technician
obtain and analyze a reactor coolant system sample. Specific documents reviewed are
described in the attachment to this report.
These activities constitute completion of one reactor coolant system specific activity
sample as defined in Inspection Procedure 71151-05.
- 26 - ENCLOSURE
b. Findings
No findings of significance were identified.
.3 Reactor Coolant System Leakage
a. Inspection Scope
The inspectors sampled licensee submittals for the reactor coolant system leakage
performance indicator for the period from the third quarter 2008 through the third quarter
2009. To determine the accuracy of the performance indicator data reported during
those periods, the inspectors used definitions and guidance contained in NEI Document
99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5. The
inspectors reviewed the licensees operator logs, reactor coolant system leakage
tracking data, issue reports, event reports, and NRC integrated inspection reports for the
period of the third quarter 2008 through the third quarter 2009 to validate the accuracy of
the submittals. The inspectors also reviewed the licensees issue report database to
determine if any problems had been identified with the performance indicator data
collected or transmitted for this indicator and none were identified. Specific documents
reviewed are described in the attachment to this report
These activities constitute completion of one reactor coolant system leakage sample as
defined in Inspection Procedure 71151-05.
b. Findings
No findings of significance were identified.
.16 Occupational Exposure Control Effectiveness (OR01)
a. Inspection Scope
The inspectors sampled licensee submittals for the Occupational Radiological
Occurrences performance indicator for the third quarter 2009. To determine the
accuracy of the performance indicator data reported during those periods, performance
indicator definitions and guidance contained in NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors
reviewed the licensees assessment of the performance indicator for occupational
radiation safety to determine if indicator related data was adequately assessed and
reported. To assess the adequacy of the licensees performance indicator data
collection and analyses, the inspectors discussed with radiation protection staff, the
scope and breadth of its data review, and the results of those reviews. The inspectors
independently reviewed electronic dosimetry dose rate and accumulated dose alarm and
dose reports and the dose assignments for any intakes that occurred during the time
period reviewed to determine if there were potentially unrecognized occurrences. The
inspectors also conducted walkdowns of numerous locked high and very high radiation
area entrances to determine the adequacy of the controls in place for these areas.
- 27 - ENCLOSURE
These activities constitute completion of the occupational radiological occurrences
sample as defined in Inspection Procedure 71151-05.
b. Findings
No findings of significance were identified.
.17 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual
Radiological Effluent Occurrences (PR01)
a. Inspection Scope
The inspectors sampled licensee submittals for the Radiological Effluent Technical
Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences
performance indicator for the third quarter 2009. To determine the accuracy of the
performance indicator data reported during those periods, performance indicator
definitions and guidance contained in NEI Document 99-02, Regulatory Assessment
Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the
licensees issue report database and selected individual reports generated since this
indicator was last reviewed to identify any potential occurrences such as unmonitored,
uncontrolled, or improperly calculated effluent releases that may have impacted offsite
dose.
These activities constitute completion of the radiological effluent technical
specifications/offsite dose calculation manual radiological effluent occurrences sample
as defined in Inspection Procedure 71151-05.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. The inspectors reviewed attributes that included the complete and accurate
identification of the problem; the timely correction, commensurate with the safety
- 28 - ENCLOSURE
significance; the evaluation and disposition of performance issues, generic implications,
common causes, contributing factors, root causes, extent of condition reviews, and
previous occurrences reviews; and the classification, prioritization, focus, and timeliness
of corrective. Minor issues entered into the licensees corrective action program
because of the inspectors observations are included in the attached list of documents
reviewed.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure, they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b. Findings
No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees corrective action program. The inspectors
accomplished this through review of the stations daily corrective action documents.
The inspectors performed these daily reviews as part of their daily plant status
monitoring activities and, as such, did not constitute any separate inspection samples.
b. Findings
No findings of significance were identified.
.3 Selected Issue Follow-up Inspection
a. Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors reviewed conditions surrounding reactor coolant system leakage and boric
acid corrosion related to reactor coolant pumps. The inspectors considered the following
during the review of the licensees actions: (1) complete and accurate identification of
problems in a timely manner; (2) evaluation and disposition of operability/reportability
issues; (3) consideration of extent of condition, generic implications, common cause, and
previous occurrences; (4) classification and prioritization of the resolution of the problem;
(5) identification of root and contributing causes of the problem; (6) identification of
corrective actions; and (7) completion of corrective actions in a timely manner.
These activities constitute completion of one in-depth problem identification and
resolution sample as defined in Inspection Procedure 71152-05.
- 29 - ENCLOSURE
b. Findings
i. Introduction. A self-revealing Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct
a condition adverse to quality. Specifically, the licensee did not promptly correct
reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on
the component cooling water heat exchanger and cover areas of three reactor
coolant pumps. Corrective actions for this condition were implemented during
Refueling Outage 15, but these corrective actions failed to correct the condition and
the vapor seal leakage continued through Operating Cycle 16. This resulted in some
additional boric acid corrosion and degradation to reactor coolant pump covers and
carbon steel component cooling water flanges.
Description. The reactor coolant pumps are designed to direct vapor stage seal
leakage to the reactor drain tank via installed piping which includes a check valve to
prevent back flow from the drain line to the vapor seal. For several cycles, the
licensee has recognized that vapor stage seal leakage has not been draining to the
reactor drain tank as designed but has instead been backing up in the line and
spilling into the pump shroud region. It was theorized that this failure of the vapor
stage leakage to flow to the reactor drain tank was due to the normally positive
pressure in the reactor drain tank and that a design change was needed. During
Refueling Outage 15, the licensee implemented Engineering Change EC-6256 to
redirect all reactor coolant pump vapor seal leakage flow to a floor drain instead of
the reactor drain tank. However, the design change did not consider the flow
restriction effects of an existing check valve in each of the reactor coolant pump
vapor stage leakage piping, and made the modification downstream of each of those
existing check valves such that vapor stage leakage no longer faced the back
pressure from the reactor drain tank, but still had to pass through the existing check
valves in order to reach the target floor drain.
The postmaintenance test prescribed by Engineering Change EC-6256 to verify flow
through the modified vapor stage leakage piping from the seal, through the leak-off
piping (including the installed check valve) to the floor drain was not implemented as
specified. Instead, because of schedule and resource impacts (it would have been
difficult, resource intensive, and intrusive to conduct the test as prescribed), a
substitute postmaintenance test was performed that only verified flow through the
portion of the piping that was modified. This meant that the postmaintenance test did
not verify that water would actually flow from the vapor stage seal, through the
existing check valves, through the new piping modification and into the floor drain.
Operating Cycle 16 proceeded following Refueling Outage 15 with the newly
modified and inadequately tested vapor stage leakage line in operation. At the
conclusion of Operating Cycle 16, Mode 3 walkdowns at the beginning of Refueling
Outage 16 identified more boric acid accumulation on three of four reactor coolant
pumps, indicating continued reactor coolant pump vapor stage leakage out onto the
heat exchanger and pump cover. The licensees root cause analysis determined that
Engineering Change EC-6256 was ineffective. A test similar to the postmaintenance
test originally prescribed by Engineering Change EC-6256 was performed on reactor
- 30 - ENCLOSURE
coolant pump 2B (which had experienced the most boric acid accumulation) and it
identified that the installed check valve RC-511B was incapable of passing flow as
intended by design. The valve was a 3/4 Velan spring loaded check valve in which
the pressure required to overcome the spring load was more than the static head of
water between the vapor stage seal and the check valve could develop. Both the
original design and the subsequent design modification implemented by Engineering
Change EC-6256 were incapable of passing flow as intended by design because the
vapor stage leakage line between the seal and the check valve could not develop
enough static head to lift the check valve before backing up and spilling over onto the
pump heat exchanger and cover. If the postmaintenance test prescribed by
Engineering Change EC-6256 had been implemented as prescribed during Refueling
Outage 15, this design flaw associated with the check valve would have been
detected and the design could have been modified to correct this condition at that
time. However, because that postmaintenance test was not properly implemented,
the condition adverse to quality (the vapor stage leakage onto the reactor coolant
pump heat exchanger and pump cover and associated boric acid accumulation and
associated corrosion) continued to exist for another operating cycle.
Analysis. The licensees failure to promptly correct a condition adverse to quality is a
performance deficiency. The finding is more than minor because it is associated with
the equipment performance attribute of the Initiating Events Cornerstone and affects
the cornerstone objective to limit the likelihood of those events that upset plant
stability. Using the Manual Chapter 0609, Attachment 4, Phase 1 screening
worksheet, the issue screened as having very low safety significance because,
although the finding contributes to the likelihood of a reactor trip, mitigation
equipment is still available. This finding had a crosscutting aspect in the area of
human performance associated with work control in that the licensee did not
effectively plan for the resources necessary to implement the postmaintenance
testing per Engineering Change EC 6256 H.3(a).
Enforcement. Title10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, that
measures shall be established to assure that conditions adverse to quality are
promptly identified and corrected. Contrary to the above, the licensee failed to
promptly correct a condition adverse to quality. Specifically, the licensee failed to
correct the reactor coolant pump vapor seal leakage with the corrective actions it
implemented during Refueling Outage 15 (ending May 31, 2008), and the vapor seal
leakage continued through operating cycle 16 until corrected during Refueling
Outage 16 (ending December 4, 2009). Because this finding was of very low safety
significance and has been entered into the licensees corrective action program as
Condition Report CR-WF3-2009-5501, it is being treated as a noncited violation
consistent with Section VI.A.1 of the NRC Enforcement Policy:
NCV 05000382/2009005-02, Reactor Coolant Pump Vapor Seal Leakage.
ii. Introduction: A self-revealing Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, was identified for the licensees failure to prescribe an
activity affecting quality by documented instructions, procedures, or drawings
appropriate to the circumstance. Specifically, for all reactor coolant pump heat
exchanger to pump cover bolted connection gasket replacements between the
- 31 - ENCLOSURE
refueling outage of 1986 (Refueling Outage 1) and the refueling outage of 2009
(Refueling Outage 16), the licensee prescribed the wrong gasket material, gasket
size, and fastener preload because they had failed to incorporate a design change
implemented during Refueling Outage 1 into their instructions, procedures, or
drawings. Station Modification Package SMP-1427, an engineering change
implemented during Refueling Outage 1 in response to industry operating
experience, called for a thicker gasket, different gasket material, and an increased
bolt preload in order to increase gasket compression and reduce the probability of
leakage. As a consequence of failing to incorporate Station Modification
Package SMP-1427 changes into procedures, all heat exchanger gasket
replacements since Refueling Outage 1, four gasket replacements in total, have
utilized thinner gaskets with less than the vendor recommended compression.
Description. After the licensees first operating cycle, industry operating experience
indicated that the reactor coolant pump heat exchanger to pump cover bolted
connection had a high probability of leakage as designed and warranted a design
modification to increase gasket compression to reduce the likelihood of reactor
coolant leakage at that interface. As a result, the licensee implemented a design
modification, Station Modification Package SMP-1427, to change the required
gasket material from stainless steel/asbestos to inconel/grafoil, to change the gasket
thickness from 0.125 inches to 0.135 inches, and to change the fastening method
from 2200-foot pounds of torque (roughly equivalent to 30 ksi tensioned) to 38.7 ksi
tensioned.
All four reactor coolant pump bolted connections were modified to the new gaskets
and fastening method as prescribed in Station Modification Package SMP-1427.
However, Technical document TD-B580.0025 was not updated with the design
change at that time. As a result, all gasket replacements conducted between
Refueling Outage 1 and Refueling Outage 16 were accomplished in accordance with
the outdated and inadequate specifications that remained in TD-B580.0025. The
result was that, by the beginning of Refueling Outage 16, only reactor coolant
pump RCP-1B still retained the modifications prescribed by Station Modification
Package SMP-1427 and implemented in Refueling Outage 1.
It is noteworthy that the inspection of reactor coolant pump 1A during the midcycle
outage on October 9, 2007, identified a sizable quantity of boric acid crystals
contained in the pump shroud. The root cause analysis concluded that the boric
acid accumulation was primarily due to leakage past the reactor coolant pump heat
exchanger to pump cover gasket. However, the root cause analysis for this leakage
did not identify that operating experience associated with leakage past these gaskets
had caused the licensee to implement Station Modification Package SMP-1427 in
Refueling Outage 1, and neither did the root cause analysis identify that the thicker
gasket and modified fastening method were needed to achieve the vendors
recommended compression. Therefore, the gasket replacement on reactor coolant
pump RCP-1A was not performed in accordance with Station Modification
Package SMP-1427. In addition, it is noteworthy that boric acid accumulation
discovered on reactor coolant pump RCP-2B on October 20, 2009, prompted
another root cause analysis by the licensee which concluded that leakage past the
- 32 - ENCLOSURE
heat exchanger to pump cover gasket may have been a possible cause of a portion
of that boric acid accumulation. The root cause analysis performed in 2007 for
reactor coolant pump RCP-1A was a missed opportunity to identify the licensees
past failure to include the Station Modification Package SMP-1427 design
modifications into plant procedures. Had that opportunity not been missed, it is
postulated that the inadequate gasket and fastener configuration on reactor coolant
pump RCP-2B may have been identified and corrected before the discovery of
significant boric acid accumulation on it during Operating Cycle 16, which may have
reduced the accumulation of boric acid on that pump.
Analysis. The licensees failure to prescribe appropriate gasket replacement
requirements in instructions, procedures, or drawings is a performance deficiency.
The finding is more than minor because it is associated with the equipment
performance attribute of the Initiating Events Cornerstone and affects the
cornerstone objective to limit the likelihood of those events that upset plant stability.
Using the Manual Chapter 0609, Attachment 4, Phase 1 screening worksheet, the
issue screened as having very low safety significance because, although the finding
contributes to the likelihood of a reactor trip, mitigation equipment is still available.
This finding had a crosscutting aspect in the area of problem identification and
resolution associated with operating experience in that the licensee did not
institutionalize operating experience through changes to the station procedures
Enforcement. Title10 CFR Part 50, Appendix B, Criterion V, requires that activities
affecting quality shall be prescribed by documented instructions, procedures, or
drawings, of a type appropriate to the circumstances. Contrary to the above, the
licensee failed to prescribe an activity affecting quality by instructions, procedures, or
drawings, of a type appropriate to the circumstances. Specifically, for all reactor
coolant pump heat exchanger to pump cover bolted connection gasket replacements
between the refueling outage of 1986 (Refueling Outage 1) and the refueling outage
of 2009 (Refueling Outage 16), the licensee prescribed the wrong gasket material,
gasket size, and fastener preload because they had failed to incorporate a design
change implemented during Refueling Outage 1 into their instructions, procedures,
or drawings. Because this finding was of very low safety significance and has been
entered into the licensees corrective action program as Condition
Report CR-WF3-2009-5501, it is being treated as a noncited violation consistent with
Section VI.A.1 of the NRC Enforcement Policy: NCV 05000382/2009005-03,
Failure to Update Drawings after Design Change.
4OA5 Other Activities
.1 Temporary Instruction 2515-172, Reactor Coolant System Dissimilar Metal Butt Welds
a. Inspection Scope
The reactor coolant system for this unit is carbon steel with stainless steel cladding and
has the following dissimilar metal welds subject to the requirements of the Materials
Reliability Program 139:
- 33 - ENCLOSURE
1. One 12-inch pressurizer surge line nozzle was mitigated during a previous
outage using a weld overlay process. The weld was classified as Category F per
materials reliability program guidelines.
2. Three 6-inch pressurizer safety nozzles were mitigated during a previous outage
using a weld overlay process. The welds were classified as Category F per
materials reliability program guidelines.
3. One 4-inch pressurizer spray nozzle was mitigated during a previous outage
using a weld overlay process. The weld was classified as Category F per
materials reliability program guidelines.
4. Two 14-inch hot leg shutdown cooling nozzles were mitigated during a previous
outage using a weld overlay process. The welds were classified as Category F
per materials reliability program guidelines.
5. One 12-inch hot leg surge nozzle was mitigated during a previous outage using a
weld overlay process. The weld was classified as Category F per materials
reliability program guidelines.
6. One 2-inch hot leg drain nozzle was mitigated during a previous outage using a
weld overlay process. The weld was classified as Category F per materials
reliability program guidelines.
7. Four 12-inch safety injection nozzles were previously left unmitigated. The
licensee performed a volumetric inspection of each nozzle during the current
outage and classified the welds as Category E per materials reliability program
guidelines.
8. Four 30-inch reactor coolant pump suction piping (unmitigated as of this outage).
The licensee performed a volumetric inspection of each pipe during the current
outage and classified the welds as Category E per materials reliability program
guidelines.
9. Four 30-inch reactor coolant pump discharge piping (unmitigated as of this
outage). The licensee performed a volumetric inspection of each pipe during the
current outage and classified the welds as Category E per materials reliability
program guidelines.
All of the pressurizer and hot-leg welds have been mitigated, in previous outages, using
a full-structural overlay weld. The cold-leg-temperature welds have not been mitigated
as of this outage. The cold-leg welds have been volumetrically inspected and any
decision to mitigate these welds will be made on the basis of these and/or future
inspections.
- 34 - ENCLOSURE
03.01 Licensees Implementation of the Materials Reliability Program (MRP-139) Baseline
Inspections
a. The inspector reviewed records of structural weld overlays and nondestructive
examination activities associated with the licensees hot leg surge nozzles
structural weld overlay mitigation effort.
b. The licensee was not planning to take any deviations from the baseline
inspection requirements of Materials Reliability Program MRP-139, and all other
applicable dissimilar metal butt welds were scheduled in accordance with
Materials Reliability Program MRP-139 guidelines.
03.02 Volumetric Examinations
a. The inspector observed the phased array ultrasonic examination of two cold leg
welds that were not scheduled to be overlaid. This examination was conducted
in accordance with ASME Code,Section XI, Supplement VIII Performance
Demonstration Initiative requirements regarding personnel, procedures, and
equipment qualifications. No relevant conditions were identified during this
examination.
b. The inspector reviewed records for the nondestructive evaluations performed on
the hot leg surge nozzle weld overlay. Inspection coverage met the requirements
of Materials Reliability Program MRP-139 and no relevant conditions were
identified.
c. The certification records of ultrasonic examination personnel were reviewed for
those personnel that performed the examinations of the cold-leg welds. All
personnel records showed that they were qualified under the EPRI Performance
Demonstration Initiative.
d. No deficiencies were identified during the nondestructive examinations.
03.03 Weld Overlays
a. The inspector reviewed the welding activities associated with the weld overlay
performed on the hot leg surge nozzle.
b. The licensee submitted and received NRC authorization for the use of relief
request from the ASME code to apply weld overlays on their dissimilar metal butt
welds. Using this, the licensee performed weld overlays on all of the dissimilar
metal butt welds associated with pressurizer and hot leg temperatures. This
welding took place in previous outages. The inspector reviewed the weld records
for one of these welds to ensure the welding was performed in accordance with
the ASME code as modified by the approved relief requests.
c. No deficiencies were identified in the completed full structural weld overlays.
- 35 - ENCLOSURE
03.04 Mechanical Stress Improvement
This item was not applicable because the licensee did not have plans to employ a
mechanical stress improvement process.
03.05 Inservice Inspection Program
The inspector reviewed the licensees risk informed inservice plan and verified that all
dissimilar metal butt welds have been entered into the plan and will be examined on a
schedule consistent with Materials Reliability Program MRP-139.
b. Findings
No findings of significance were identified.
4OA6 Meetings
Exit Meeting Summary
On October 1, 2009, the inspector conducted a telephonic exit meeting to present the results of
the in-office inspection of changes to the Waterford Steam Electric Station, Unit 3s, emergency
action levels to Mr. J. Lewis, Manager, Emergency Preparedness. He acknowledged the issues
presented. The inspector asked whether any materials examined during the inspection should
be considered proprietary. No proprietary information was identified
On November 9, 2009, the inspector conducted a telephonic exit meeting to present the results
of the in-office inspection of changes to the Waterford Steam Electric Station, Unit 3, emergency
plan and emergency action levels to Mr. R. Perry, Acting Emergency Preparedness Manager.
He acknowledged the issues presented. The inspector asked whether any materials examined
during the inspection should be considered proprietary. No proprietary information was
identified.
On November 13, 2009, the inspectors presented the results of the inservice inspection to you
and other members of your staff. You acknowledged the issues presented. The inspectors
returned proprietary material examined during the inspection.
On November 20, 2009, the inspectors presented the inspection results to Mr. C. Arnone,
General Manager, Plant Operations, and other members of your staff. They acknowledged the
issues presented. The inspector asked whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
On January 11, 2010, the inspectors presented the quarterly inspection results to you and other
members of your staff. You acknowledged the issues presented. The inspectors asked whether
any materials examined during the inspection should be considered proprietary. No proprietary
information was identified.
- 36 - ENCLOSURE
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as noncited violations.
Technical Specification 6.8.1 requires written procedures be established, implemented, and
maintained covering the applicable procedures recommended in Appendix A of Regulatory
Guide 1.33, Revision 2, February 1978. Appendix A lists procedures for access control to
radiation areas. Procedure EN-RP-100, Radworker Expectations, Revision 3, Section 5.3[9]
requires the radiation work permit to be read, understood, and obeyed as a condition of
radiologically controlled area access. Procedure EN-RP-100, Radworker Expectations,
Revision 3, Section 5.4[3](h) requires the worker know where to properly perform his/her task.
Section 5.3[17] requires the worker be briefed and sign on the appropriate radiation work permit.
Section 5.3[11] requires the worker know the radiological conditions in the work area. The
licensee identified an example of a worker entering a high radiation area using an inappropriate
radiation work permit and without knowing the dose rates in the area. On October 24, 2009, a
security officer entered shutdown heat exchanger Room B and received an electronic dosimeter
dose rate alarm. The room was posted as a high radiation area and dose rates within the area
were as high as 140 millirem per hour. The officer entered the radiological controlled area using
Radiation Work Permit 2009005, Tours and Inspection in All Radiological Controlled Areas,
Except High Radiation Areas, Locked High Radiation Areas, Very High Radiation Areas, and the
Reactor Containment Building. Because the radiation work permit did not allow entry into high
radiation areas, radiation protection personnel did not anticipate the officer would enter the room
and did not brief the officer on the dose rates in the area. In response, the licensee conducted a
human performance error review and counseled the officer. This finding was of very low safety
significance because it did not involve an actual or substantial potential of an overexposure.
This finding was entered into the licensees corrective action program as Condition
Report CR-WF3-2009-05648.
- 37 - ENCLOSURE
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
C. Arnone, General Manager Plant Operations
D. Bauman, Senior Project Manager
M. Bratton, Manager, Senior Lead Technical Specialist
J. Brawley, ALARA Supervisor, Radiation Protection
B. Celeste, Lead Level III, Contractor, C&S Engineers, Inc.
K. Cook, Acting General Manager Plant Operations
L. Dauzat, Supervisor, Radiation Protection
D. Dufrene; Technician, Radiation Protection
G. Ferguson, PE, IWE Examination
J. Gobell, Project Manager
J. Houghtaling, Senior Project Manager
C. Hunsaker, Technical Specialist II
J. Kowalewski, Vice President, Operations
J. Lewis, Manager, Emergency Preparedness
R. Luter, Technical Specialist IV
M. Mason, Engineer, Licensing
R. McGaha, Technical Specialist II
M. Mason, Engineer, Licensing
R. Murillo, Manager, Licensing
K. Nichols, Director, Engineering
R. OQuinn, Senior Staff Engineer
C. Pickering, Supervisor, Mechanical Maintenance
B. Piluti, Manager, Radiation Protection
J. Polluck, Engineer, Licensing
R. Redmond, Technical Specialist, Boric Acid Corrosion Control Program
W. Sims, Manager, Major Projects I
B. Williams, Technical Specialist IV
R. Williams, ASME Section XI/ISI Senior Lead
NRC Personnel
M. Haire, Senior Resident Inspector
D. Overland, Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Failure to follow radiation protection procedural
requirements05000382/2009005-02 NCV Reactor Coolant Pump Vapor Seal Leakage
A-1 Attachment
Opened and Closed
Failure to follow radiation protection procedural
requirements05000382/2009005-02 NCV Reactor Coolant Pump Vapor Seal Leakage
05000382/2009005-03 NCV Failure to Update Drawings after Design Change
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
WSES-FSAR- Final Safety Analysis Report - Section 2.4, Hydrologic 10
UNIT-3 Engineering
OP-901-521 Off-Normal Procedure for Severe Weather and Flooding 301
Section 1R04: Equipment Alignment
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
OP-002-004 Chilled Water System 303
OP-903-063 Chilled Water Pump Operability Verification 302
SD-CHW Essential Chilled Water System Description 6
G853 Sheet 3 Chilled Water flow Diagram SH-1 December 4,
1975
SD-SI Safety Injection System Description 13
OP-009-008 Safety Injection System Operating Procedure 26
A-2 Attachment
Section 1R05: Fire Protection
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
OP-009-004 Fire Protection 305
MM-004-424 Building Fire Hose Station Inspection and Hose 10
Replacement
MM-007-010 Fire Extinguisher Inspection and Extinguisher Replacement 302
FP-001-014 Duties of a Fire Watch 14
FP-001-015 Fire Protection Impairments 302
DBD-018 Appendix R/Fire Protection
FP-001-015 Fire Protection Impairments 302
FP-001-018 Pre-fire Plan Strategies, Development, And Revision 300
UNT-007-006 Housekeeping 301
EN-DC-161 Control of Combustibles 003
UNT-007-060 Control of Loose Items 302
UNT-005-013 Fire Protection Program 010
SD-FP Fire Protection System Description 2
Section 1R06: Flood Protection Measures
CONDITION REPORTS
CR-WF3-2005-03338 CR-WF3-1996-00930 CR-WF3-2009-3925
PROCEDURE/DOCUMENTS
NUMBER TITLE REVISION /
DATE
WSES-FSAR- February
Appendix 3.6A Pipe Rupture Analysis
UNIT-3 2002
WSES-FSAR- February
Water Level (Flood) Design
UNIT-3 2002
WSES-FSAR-
System Description Plant Sumps 6
UNIT-3
OP-901-521 Severe Weather and Flooding 301
G-349 Yard Duct Runs and Outdoor Lighting Drawing 18
A-3 Attachment
Section 1R07: Heat Sink Performance
PROCEDURES/DOCUMENTS
NUMBER TITLE DATE
NOECP-257 Steam Generator Secondary Side Inspections 4
LTR-SGDA-08-129 Acceptability of Loose Batwing Section found in the May 12, 2008
Upper Central Stay Cavity Region during RF15
LTR-SGDA-09-189 Acceptability of SG Operation As a Result of an November 16,
Unattached Steam Vent and Observed Feedwater Ring 2009
Erosion
LTR-SGDA-09-188 Acceptance Criteria for Waterford Feedwater November 13,
Discharge Elbows 2009
Section 1RO8: Inservice Inspection Activities
DOCUMENTS/PROCEDURES/REPORTS
NUMBER TITLE REVISION / DATE
EN-DC-317 Entergy Steam Generator Administrative 4
Procedure
NOECP-257 Steam Generator Secondary Side Inspection 4
NOECP-252 Steam Generator Eddy Current Inspection 11
Testing
CEP-NDE-0955 Alloy 600 Visual Examination (VE) of 301
Bare-Metal Surfaces
EN-DC-319 Inspection and Evaluation of Boric Acid Leaks 4
NOECP-107 Boric Acid corrosion Control Program 3
WF3-CHEM-SEC-001- Strategic Secondary Water Chemistry Plan 6
06
WDI-PJF-1304321- Waterford 3 - RF16 - Reactor Vessel Head 0
FSR-001 Penetration Inspection Final Report.
WDI-SSP-1002 Reactor Vessel Head Penetration Inspection 3
Tool Operation for ANO 2 and Waterford 3 -
ROSA
WCAL-002 Pulser/Receiver Linearity Procedure 10
WDI-ET-003 IntraSpect Eddy Current Imaging Procedure 14
for Inspection of Reactor Vessel Head
Penetrations
WDI-ET-004 IntraSpect Eddy Current Analysis Guidelines 14
A-4 Attachment
Section 1R07: Heat Sink Performance
PROCEDURES/DOCUMENTS
NUMBER TITLE DATE
WDI-STD-1040 IntraSpect Ultrasonic Procedure for Inspection 2
of Reactor Vessel Head Penetrations, Time of
Flight Ultrasonic, Longitudinal Wave and
Shear Wave
WDI-STD-1041 IntraSpect UT Analysis Guidelines 1
WDI-STD-101 RVHI Vent Tube J-Weld Eddy Current 8
Examination
WDI-STD-114 RVHI Vent Tube ID & CS Wastage Eddy 10
Current Examination
CEP-NDE-0404 Manual Ultrasonic Examination of Ferritic 4
Piping Welds (ASME XI)
ISI-UT-09-019 UT Calibration/Examination (WO 157687) - October 31, 2009
RCS Cold Leg Loop 1A - Weld No.07-005
L-09-006 Ultrasonic Instrument Linearity - Krautkramer October 22, 2009
USN 60 SW (Serial No. 01VNCT); Transducer
Frequency 4.0 MHz (Serial No. 5746222529);
Calibration Standard (Serial No. 9634);
Couplant - Ultragel II (Batch No. 06225)
ISI-VT-09-194 Visual Examination for Boric Acid Detection October 27, 2009
(WO 159119) - RCS Loop 1A Cold Leg -
Weld No.07-002
MRP-139 Material Reliability Program: Primary System 1
Piping Butt Weld Inspection and Evaluation
Guideline
CEP-NDE-0901 VT-1 Examination 4
CEP-NDE-0902 VT-2 Examination 7
CEP-NDE-0903 VT-3 Examination 5
SI-UT-130 Procedure for the Phased Array Ultrasonic 3
Examination of Dissimilar Metal Welds
SI-NDE-06 Calibration of Ultrasonic NDE Equipment 4
SI-NDE-08 Qualification and Certification of NDE 1
Personnel for Nuclear Applications
WF3 11-002 RCP 2A Structural Integrity Associates - Phased Array October 30, 2009
Suction Nozzle Ultrasonic Examination Record Data Sheet for
Weld No. 11-002: Reactor Coolant Pump 2A
Cold Leg Suction Nozzle
A-5 Attachment
Section 1R07: Heat Sink Performance
PROCEDURES/DOCUMENTS
NUMBER TITLE DATE
WF3 11-002 RCP 2A Structural Integrity Associates - Ultrasonic October 30, 2009
Suction AX SH Phased Array Calibration Record for Weld
No. 11-002: Reactor Coolant Pump Suction
Nozzle Dissimilar Metal Weld - Wedge
Angle 36.2o (Axial Scan)
WF3 11-002 RCP 2A Structural Integrity Associates - Ultrasonic October 30, 2009
Suction Circ - 10 RL Phased Array Calibration Record for Weld
No. 11-002: Reactor Coolant Pump Suction
Nozzle Dissimilar Metal Weld - Wedge
Angle 4.0o (Circumferential Scan)
WF3 11-002 RCP 2A Structural Integrity Associates - Ultrasonic October 30, 2009
Suction Circ + 10 RL Phased Array Calibration Record for Weld No.
11-002: Reactor Coolant Pump Suction
Nozzle Dissimilar Metal Weld - Wedge
Angle 14.0o (Circumferential Scan)
WF3 11-002 RCP 2A Structural Integrity Associates - Ultrasonic October 30, 2009
Suction Flat RL Phased Array Calibration Record for Weld
No. 11-002: Reactor Coolant Pump Suction
Nozzle Dissimilar Metal Weld - Wedge Angle
14.0o (Axial & Circumferential Scan)
WF3-LIN-09-002 Structural Integrity Associates - Ultrasonic October 21, 2009
Linearity Record - Zetec/RD Tech OmniScan
MX - Version 1.4R3 (Serial No. ONMI-1983);
Transducer 115-000-613 (Serial
No. 01VTVW); Reference Block 16 AX (Serial
No. SI-16-AX-03).
Product Code 115-000- Krautkramer Phased Array Transducer September 02, 2008
566 Certificate of Compliance (Serial
No. 01VM4k-1)
SII006-07-09-28155-1 Laboratory Testing Inc. - Certified Test Report July 27, 2007
for Sonotech Ultragel II
WF3 12-009 RCP 2A Structural Integrity Associates - Phased Array October 29, 2009
Safety Injection Nozzle Ultrasonic Examination Record Data Sheet for
Weld No. 12-009: RCP 2A Safety Injection
Nozzle
WF3 12-009 RCP 2A Structural Integrity Associates - Ultrasonic October 29, 2009
Safety Injection AX SH Phased Array Calibration Record for Weld
No. 12-009: Reactor Coolant Pump 2A Safety
Injection Nozzle Dissimilar Metal Weld -
Wedge Angle 36.2o (Axial Scan)
A-6 Attachment
Section 1R07: Heat Sink Performance
PROCEDURES/DOCUMENTS
NUMBER TITLE DATE
WF3 12-009 RCP 2A Structural Integrity Associates - Ultrasonic October 29, 2009
Safety Injection AX RL Phased Array Calibration Record for Weld
No. 12-009: Reactor Coolant Pump 2A Safety
Injection Nozzle Dissimilar Metal Weld -
Wedge Angle 16.2o (Axial Scan)
WF3 12-009 RCP 2A Structural Integrity Associates - Ultrasonic October 29, 2009
Safety Injection CIRC Phased Array Calibration Record for Weld
RL No. 12-009: Reactor Coolant Pump 2A Safety
Injection Nozzle Dissimilar Metal Weld -
Wedge Angle 16.2o (Circumferential Scan)
Contract No. C-08-422 Sonaspection - Structural Integrity of December 17, 2009
Calibration Block No. SI-16-AX-03 & SI-16-
CIRC-03
WF3-LIN-09-003 Structural Integrity Associates - Ultrasonic October 21, 2009
Linearity Record - Zetec/RD Tech OmniScan
MX - Version 1.4R3 (Serial No. ONMI-1590);
Transducer 115-000-613 (Serial No.
01VTW0); Reference Block 16 AX (Serial
No. SI-16-AX-03).
Product Code 115-000- Krautkramer Phased Array Transducer August 26, 2008
613 Certificate of Conformity (Serial
No. 01VTW0-1)
ENGINEERING CHANGE REQUEST
NUMBER TITLE DATE
0000004490 Steam Generator Degradation Assessment and Repair April 2008
Criteria for RF15
0000005544 Waterford 3 Cycle 16 Steam Generator Operational April 2008
Assessment
0000005544 Waterford 3 Cycle 16 Steam Generator Operational August 2008
Assessment
0000008593 Waterford-3 RF16 Steam Generator Eddy Current Probe Revision 0
Equivalency Report
A-7 Attachment
ENGINEERING CHANGE REQUEST
NUMBER TITLE DATE
0000008594 Waterford-3 RF16 Steam Generator Inspection ECT
Data Analyst Training Manual
0000008592 RF16 Waterford-3 Steam Generator Analysis Guidelines Revision 0
0000008591 Steam Generator Degradation Assessment and Repair October 2009
Criteria for RF16
MISCELLANEOUS DOCUMENTS
NUMBER TITLE DATE
ECR-WF3-4490 Steam Generator Degradation Assessment and Repair April 2008
Criteria
W3F1-2008-0039 Steam Generator Conditions Observed at Waterford 3 May 20, 2008
During Refueling Outage 15
ECR-WF3-8593 Waterford -3 RF16 Steam Generator Eddy Current Probe November 3,
Equivalency Report 2009
ECR-WF3-8594 Document the Analysts Training Manual for RF16 SG November 6,
Eddy Current Analysts per the Requirements of NEI 97- 2009
06 and EN-DC-317
ECR-WF3-8592 RF16 Waterford-3 Steam Generator Analysis Guidelines November 5,
2009
ECR-WF3-8591 Steam Generator Degradation Assessment and Repair October 2009
Criteria for RF16
WF3-CHEM- Strategic Secondary Water Chemistry Plan 6
SEC-001-06
Inspection Report for Bare Metal Visual of Reactor
Vessel Head
BOP-VT-09-020 Visual Examination of Boric Acid Detection November 12,
2009
LTR-SGMP-09- Estimate of Through-Tube Depth of Intrados Wear Scar November 10,
179 in Waterford Steam Generator 32 2009
LO-WLO-2008- WF3 Boric Acid corrosion Control Program Self- October 6-16,
00068 Assessment 2008
LO-WLO-2006- Waterford 3 Strategic Secondary Water Chemistry Plan March 27-30,
00046 Self-Assessment 2006
A-8 Attachment
LO-WLO-2008- Benchmark of: Point Beach (PBNP) Nuclear Plant July 16-17,
0091 2009
W3F1-2008-0039 Steam Generator Conditions Observed at Waterford 3 May 20, 2008
WDI-PJF- Waterford 3 RF16 Reactor Vessel Head Penetration 0
1304321-FSR- Inspection Final Report
001
DWG U.T. Calibration Standard UT-6 (Contract No.74470) March 14,
C-246-392-2 1974
CNRO-2007-002 Mitigating Actions and Associated Schedule for Alloy
600/82/182
Weld No.12-009 Waterford 3 Dissimilar-Metal Weld Walk-Down Data
Sheet 4: 12 SI Nozzle to Safe-End
Various Personnel Certifications and Certification
Reviews
Bare Metal Visual Inspections Scheduled for RF-16
RF-16 Steam Generator Scope Summary
WELDING DATA RECORDS
2009-4293 2009-4528 2009-4588
CONDITION REPORTS
CR-WF3-2006-3966 CR-WF3-2008-2283 CR-HQN-2009-1068 CR-WF3-2009-5194
CR-WF3-2009-5501 CR-WF3-2009-5502 CR-WF3-2009-5509 CR-WF3-2009-5511
CR-WF3-2009-5514 CR-WF3-2009-5515 CR-WF3-2009-5516 CR-WF3-2009-5553
CR-WF3-2009-5554 CR-WF3-2009-5555 CR-WF3-2009-5556 CR-WF3-2009-5585
CR-WF3-2009-5662 CR-WF3-2009-5671 CR-WF3-2009-5679 CR-WF3-2009-5700
CR-WF3-2009-5716 CR-WF3-2009-5735 CR-WF3-2009-5757 CR-WF3-2009-5765
CR-WF3-2009-5769 CR-WF3-2009-5770 CR-WF3-2009-5774 CR-WF3-2009-5836
CR-WF3-2009-5838 CR-WF3-2009-5899 CR-WF3-2009-5941 CR-WF3-2009-5944
CR-WF3-2009-6486 CR-WF3-2009-6504 CR-WF3-2009-6514 CR-WF3-2009-6620
A-9 Attachment
Section 1R11: Licensed Operator Requalification Program
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
EN-TQ-114 Licensed Operator Requalification Training Program 0
Description
O-JITDIL Simulator Scenario for Dilution JIT 3
Section 1R12: Maintenance Effectiveness
CONDITION REPORTS
WF3-CR-2008-2637 WF3-CR-2008-2641 WF3-CR-2008-2689 WF3-CR-2008-2721
WF3-CR-2008-3103 WF3-CR-2008-3976 WF3-CR-2008-4012 WF3-CR-2008-4033
WF3-CR-2008-4635 WF3-CR-2008-4953 WF3-CR-2009-2189 WF3-CR-2009-2762
WF3-CR-2009-2796 WF3-CR-2009-3507 WF3-CR-2009-4066 WF3-CR-2009-4088
WF3-CR-2009-4093 WF3-CR-2009-4098 WF3-CR-2009-4155 WF3-CR-2009-5335
WF3-CR-2008-3217 WF3-CR-2008-4992 WF3-CR-2009-1901 WF3-CR-2009-2485
WF3-CR-2008-4453 WF3-CR-2008-5266 WF3-CR-2009-2077 WF3-CR-2009-4499
WF3-CR-2008-4583 WF3-CR-2009-0214 WF3-CR-2009-2096 WF3-CR-2009-5804
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
EN-DC-206 Maintenance Rule 1
NUMARC 93-01 Industry Guideline for Monitoring the Effectiveness of 3
maintenance at Nuclear Power Plants
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
EOOS Version 3.3a Schedulers Evaluation for Shutdown Version Waterford November 5,
3 Rev 3 Model 2009
A-10 Attachment
N/A RF16 Daily Outage Status Report October 24,
2009
OP-903-107 Surveillance Procedure for Plant Protection System 303
Channel Functional Test
EOOS Version 3.3a Schedulers Evaluation for Shutdown Version 12/03/2009
Waterford 3 Rev 3 Model
Section 1R15: Operability Evaluations
CONDITION REPORTS
CR-WF3-2009-6101 CR-WF3-2008-2684 CR-WF3-2008-2705 CR-WF3-2008-2730
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
EN-OP-104 Operability Determination 4
MI-003-126 Core Protection Calculator Functional 14
SD-PPS Plant Protection System Description 0
OP-903-107 Plant Protection System Channel A, B, C, D, Functional Test 303
TSTF-324 Correct logarithmic power vs. RTP 1
ECE98-001 Calculation of Maximum Allowable Battery Inter Cell 0
Connection Resistance
ECE98-001 Calculation of Maximum Allowable Battery Inter Cell 1
Connection Resistance
ME-003-220 Station Battery Bank & Charger (18 month) 303
ME-003-220 Station Battery Bank & Charger (18 month) 301
SD-NI Nuclear Instrumentation System Description 6
Section 1R19: Postmaintenance Testing
CONDITION REPORTS
CR-WF3-2009-6095 CR-WF3-2009-6412 CR-WF3-2008-2381 CR-WF3-2009-6461
CR-WF3-2009-6449 CR-WF3-2008-4179 CR-WF3-2009-6506 CR-WF3-2009-4499
WORK ORDERS
1517161 213478 187774 152910
161402 122097 212157
A-11 Attachment
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
STA-001-004 Local Leak Rate Test 303
ICE-37718 Siemens Motor Driven Relay Observed Contact Behavior 02/05/1999
OP-903-116 Train B Integrated Emergency Diesel Generator/Engineering 013
Safety Features Test
ME-003-230 Battery Service Test 306
ME-003-240 Battery Performance Test 306
ME-004-213 Battery Intercell Connections 14
ME-004-231 Station Battery Charging 19
ME-003-210 Station Battery Bank and Charger (Quarterly) 16
ME-003-220 Station Battery Bank and Charger (18 month) 303
OP-903-046 Emergency Feed Pump Operability Check - Attachment 10.3 305
Section 1R20: Refueling and Other Outage Activities
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
OP-903-027 Inspection of Containment 301
PLG-009-014 Conduct of Planned Outages 303
OP-001-003 Reactor Coolant System Drain Down 306
OI-037-000 Operations Risk Assessment Guideline 2
MM-004-201 Containment Building Polar Crane PM 303
WF3-CS-08-01 NEI Heavy Load Drop Initiative 0
UNT-007-008 Control of Loads and Lifting 302
RF-001-009 Reactor Head 303
NEI 08-05 Industry Initiative on Control of Heavy Loads 0
MM-007-003 Containment Building Polar Crane Testing 5
A-12 Attachment
Section 1R22: Surveillance Testing
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION /
DATE
OP-903-116 Train B Integrated Emergency Diesel Generator/Engineering 013
Safety Features Test
OP-903-120 Section 7.10 Annulus Negative Pressure Surveillance Test 9
Section 2OS1: Access Controls to Radiologically Significant Areas
CONDITION REPORTS
CR-WF3-2009-5492 CR-WF3-2009-5648 CR-WF3-2009-5878 CR-WF3-2009-5880
CR-WF3-2009-6767 CR-WF3-2009-6792 CR-WF3-2009-6834 CR-WF3-2009-6852
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
EN-RP-100 Radworker Expectations 3
EN-RP-101 Access Control for Radiologically 4
Controlled Areas
EN-RP-102 Radiological Control 2
EN-RP-105 Radiation Work Permits 6
EN-RP-108 Radiation Protection Posting 7
EN-RP-121 Radioactive Material Control 4
EN-RP-123 Radiological Controls for Highly 0
Radioactive Particles
HP-001-114 Control of Temporary Shielding 10
UNT-001-016 Radiation Protection 301
UNT-007-001 Control of Miscellaneous Material in the 5
Spent Fuel Pool
A-13 Attachment
AUDITS, SELF-ASSESSMENTS, AND SURVEILLANCES
PROCEDURE/DOCUMENTS
NUMBER TITLE DATE
QA-14/15-2009- Radiation Protection/Radwaste Audit September
WF3-1 2009
RADIATON WORK PERMITS
NUMBER DESCRIPTION
2009-0401 Perform UDS/Viper/Votes and/or AOV/MOV testing of contaminated
system valves
2009-0510 Install/Remove Steam Generator Nozzle Dams, Pin verification, &
closeout
2009-0512 Remove/Install Steam Generator Secondary Manways/Handholes
2009-0513 RCP 1A Motor and Driver Mount removal and replacement
2009-0603 Entries into posted LHRA of the Reactor Containment Building to
perform minor maintenance activities, walkdowns, surveillances, and
inspections
2009-0606 Perform minor maintenance activities, walkdowns, surveillances, and
inspections
2009-0628 Entries into Containment Sump to perform transmitter calibrations,
Weir Box cleaning and Under Vessel inspections
2009-0721 Entries into posted LHRA of the Reactor Containment Building to
install/remove shielding on the ICI stalks
2009-0805 Refuel 16 - Tours and inspections in all RCAs except HRA, LHRA,
A-14 Attachment
SAMPLE RESULTS AND SURVEYS
MISCELLANEOUS
NUMBER TITLE DATE
WF3-0910-0398 Survey of RAB -35 Shutdown Heat Exchangers October 23, 2009
WF3-0910-0431 Survey of RAB -35 Shutdown Heat Exchangers October 24, 2009
Section 2OS2: ALARA Planning and Controls
PROCEDURES
NUMBER TITLE REVISION
HP-002-201 Radiological Survey Techniques and Frequencies 302
EN-RP-104 Personnel Contamination Events 4
EN-RP-106 Radiological Survey Documentation 2
EN-RP-131 Air Sampling 7
EN-RP-203 Dose Assessment 3
MISCELLANEOUS
NUMBER TITLE DATE
2009-0020 Personnel Contamination Event Record October 29, 2009
2009-0045 Personnel Contamination Event Record November 3, 2009
2009-0049 Personnel Contamination Event Record November 5,2009
Section 4OA1: Performance Indicator Verification
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
NEI 99-02 Regulatory Assessment Performance Indicator Guideline 5
EN-LI-114 Performance Indicator Process 4
EN-DIR-RP-002 Radiation Protection Performance Indicator Program 0
MISCELLANEOUS DOCUMENTS
Radiological controlled area entries greater than 100 millirem
A-15 Attachment
Section 4OA2: Identification and Resolution of Problems
CONDITION REPORTS
CR-WF3-2009-5501 CR-WF3-2009-5502 CR-WF3-2009-5509 CR-WF3-2009-5511
CR-WF3-2009-5514 CR-WF3-2009-7166 CR-WF3-2009-7159
Section 4OA5: Other Activities
DOCUMENTS
NUMBER TITLE REVISION /
DATE
CEP-NDE-0955 Alloy 600 Visual Examination (VE) of Bare-Metal Surfaces 301
EC-1830 Waterford Steam Electric Station, Unit 3, Dissimilar Metal 0
Drawing No. Hot Leg Surge Nozzle Weld Overlay Design 5
WSES-19Q-05
SI-UT-130 Procedure for the Phased Array Ultrasonic Examination of 3
SI-NDE-06 Calibration of Ultrasonic NDE Equipment 4
SI-NDE-08 Qualification and Certification of NDE Personnel for 1
Nuclear Applications
CEP-NDE-0901 VT-1 Examination 4
CEP-NDE-0902 VT-2 Examination 7
CEP-NDE-0903 VT-3 Examination 5
WF3 11-002 Structural Integrity Associates - Phased Array Ultrasonic October 30,
RCP 2A Suction Examination Record Data Sheet for Weld No. 11-002: 2009
Nozzle Reactor Coolant Pump 2A Cold Leg Suction Nozzle
WF3 11-002 Structural Integrity Associates - Ultrasonic Phased Array October 30,
RCP 2A Suction Calibration Record for Weld No. 11-002: Reactor Coolant 2009
AX SH Pump Suction Nozzle Dissimilar Metal Weld - Wedge
Angle 36.2o (Axial Scan)
WF3 11-002 Structural Integrity Associates - Ultrasonic Phased Array October 30,
RCP 2A Suction Calibration Record for Weld No. 11-002: Reactor Coolant 2009
Circ + 10 RL Pump Suction Nozzle Dissimilar Metal Weld - Wedge
Angle 14.0o (Circumferential Scan)
WF3 11-002 Structural Integrity Associates - Ultrasonic Phased Array October 30,
RCP 2A Suction Calibration Record for Weld No. 11-002: Reactor Coolant 2009
Flat RL Pump Suction Nozzle Dissimilar Metal Weld - Wedge
Angle 14.0o (Axial & Circumferential Scan)
A-16 Attachment
WF3-LIN-09-002 Structural Integrity Associates - Ultrasonic Linearity October 21,
Record - Zetec/RD Tech OmniScan MX - Version 1.4R3 2009
(Serial No. ONMI-1983); Transducer 115-000-613 (Serial
No. 01VTVW); Reference Block 16 AX (Serial No. SI-16-
AX-03).
Contract No. C- Sonaspection - Structural Integrity of Calibration Block May 18, 2009
09-089 R1 No. SI-Flat-SS-4inchT-01
A-17 Attachment