ML062090477
| ML062090477 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/28/2006 |
| From: | Fredrickson P NRC/RGN-II/DRP/RPB4 |
| To: | Gannon C Carolina Power & Light Co |
| References | |
| IR-06-003 | |
| Download: ML062090477 (46) | |
See also: IR 05000400/2006003
Text
July 28, 2006
Carolina Power and Light Company
ATTN: Mr. C. J. Gannon, Jr.
Vice President - Harris Plant
Shearon Harris Nuclear Power Plant
P. O. Box 165, Mail Code: Zone 1
New Hill, North Carolina 27562-0165
SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT - NRC INTEGRATED
INSPECTION REPORT 05000400/2006003
Dear Mr. Gannon:
On June 30, 2006, the US Nuclear Regulatory Commission (NRC) completed an inspection at
your Shearon Harris reactor facility. The enclosed integrated inspection report documents the
inspection findings, which were discussed on July 20, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the 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 one finding concerning the performance of inadequate maintenance on
the train A chiller of the essential services chilled water (ESCW) system, resulting in that
component being inoperable for a period of time in excess of that permitted by your Technical
Specifications. This finding was determined to involve a violation of NRC requirements and has
potential safety significance greater than very low safety significance. The finding did not
present an immediate safety concern in that a fully redundant train B of the ESCW system,
with its associated chiller, remained operable or available during the A trains period of
inoperability. Additionally, the train A chiller has been returned to service and the condition of
concern no longer exists. The NRC will inform you of its final determination of the significance
of the condition and any associated enforcement action.
In addition, the report documents one self-revealing finding of very low safety significance
(Green). This finding was determined to involve a violation of NRC requirements. However,
because of its very low safety significance and because it has been entered into your corrective
action program, the NRC is treating this issue as a non-cited violation, in accordance with
Section VI.A.1 of the NRCs Enforcement Policy. If you deny this non-cited violation, you should
provide a response with the basis for your denial, within 30 days of the date of this inspection
report to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC
20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of
Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001;
and the NRC Resident Inspector at the Shearon Harris facility.
CP&L 2
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) components 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/
Paul E. Fredrickson, Chief
Reactor Projects Branch 4
Division of Reactor Projects
Docket No.: 50-400
License No.: NPF-63
Enclosure: NRC Inspection Report 05000400/2006003
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
_ML062090477
OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE RXM /RA/ PBO /RA/ SJV /RA/ SJV for /RA/ BWM1 /RA/ SJV for /RA/ MSL for /RA/
NAME RMusser POBryan SVias SJV BCrowley BMiller LLake JRivera-Ortiz
DATE 07/28/2006 07/28/2006 07/28/2006 07/18/2006 07/18/2006 07/18/2006 07/18/2006
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE MSL for /RA/ SON for /RA/ SON for /RA/ SON for /RA/
NAME MScott WLoo RHamilton JKreh
DATE 07/18/2006 07/20/2006 07/20/2006 07/20/2006
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
CP&L 3
cc w/encl: Beverly Hall, Acting Director
Chris L. Burton, Manager Division of Radiation Protection
Performance Evaluation and N. C. Department of Environmental
Regulatory Affairs CPB 9 Commerce & Natural Resources
Carolina Power & Light Company Electronic Mail Distribution
Electronic Mail Distribution
Public Service Commission
Robert J. Duncan II State of South Carolina
Director of Site Operations P. O. Box 11649
Carolina Power & Light Company Columbia, SC 29211
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution Chairman of the North Carolina
Utilities Commission
Eric McCartney c/o Sam Watson, Staff Attorney
Plant General Manager--Harris Plant Electronic Mail Distribution
Progress Energy Carolinas, Inc.
Shearon Harris Nuclear Power Plant Robert P. Gruber
Electronic Mail Distribution Executive Director
Public Staff NCUC
J. Wayne Gurganious 4326 Mail Service Center
Training Manager-Harris Plant Raleigh, NC 27699-4326
Progress Energy Carolinas, Inc.
Harris Energy & Environmental Center Herb Council, Chair
Electronic Mail Distribution Board of County Commissioners
of Wake County
Christos Kamilaris, Manager P. O. Box 550
Support Services Raleigh, NC 27602
Carolina Power & Light Company
Shearon Harris Nuclear Power Plant Tommy Emerson, Chair
Electronic Mail Distribution Board of County Commissioners
of Chatham County
David H. Corlett, Supervisor Electronic Mail Distribution
Licensing/Regulatory Programs
Carolina Power & Light Company Distribution w/encl: (See page 4)
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
David T. Conley
Associate General Counsel - Legal
Department
Progress Energy Service Company, LLC
Electronic Mail Distribution
John H. O'Neill, Jr.
Shaw, Pittman, Potts & Trowbridge
2300 N. Street, NW
Washington, DC 20037-1128
CP&L 4
Report to C.J. Gannon from Paul E. Fredrickson dated July 28, 2006
SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT - NRC INTEGRATED
INSPECTION REPORT 05000400/2006003
Distribution w/encl:
C. Patel, NRR
L. Slack, RII EICS
RIDSNRRDIRS
PUBLIC
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No: 50-400
License No: NPF-63
Report No: 05000400/2006003
Licensee: Carolina Power and Light Company
Facility: Shearon Harris Nuclear Power Plant, Unit 1
Location: 5413 Shearon Harris Road
New Hill, NC 27562
Dates: April 1 through June 30, 2006
Inspectors: R. Musser, Senior Resident Inspector
P. OBryan, Resident Inspector
S. Vias, Senior Reactor Inspector, (Section 1R08)
B. Crowley, Senior Reactor Inspector, (Sections 1R08, 4OA5)
B. Miller, Reactor Inspector, (Section 1R08)
L. Lake, Reactor Inspector, (Sections 1R08, 4OA5)
J. Rivera-Ortiz, Reactor Inspector, (Sections 1R08, 1R12,
4OA5)
M. Scott, Senior Reactor Inspector, (Section 1R12)
W. Loo, Senior Health Physicist (Sections 2OS1, 2OS2, 2PS2)
R. Hamilton, Senior Health Physicist (Sections 2PS2, 4OA1)
J. Kreh, Emergency Preparedness Inspector (Section 2OS1)
Approved by: P. Fredrickson, Chief
Reactor Projects Branch 4
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000400/2006-003; 04/01/2006 - 06/30/2006; Shearon Harris Nuclear Power Plant, Unit 1;
Operability Evaluations.
The report covered a three-month period of inspection by resident inspectors, and announced
inspections by two regional senior health physics inspectors, one regional health physics
inspector, one regional emergency preparedness inspector, three regional senior reactor
inspectors, and three regional reactor inspectors. One Green non-cited violation (NCV), and
one AV with a potential safety significance greater than Green, were identified. The
significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP 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 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. A Green self-revealing NCV of Technical Specification (TS) 6.8.1 was
identified for the failure to follow procedures while performing maintenance on a
service water valve which supports the train A essential services chilled water
(ESCW) system chiller. This deficiency led to the valve actuator disconnecting
from the valve, and rendered the train A ESCW system chiller inoperable. The
licensee entered this failure to follow procedure into the Corrective Action
Program (CAP).
This finding is more than minor because it affected the reliability objective of the
equipment performance attribute under the Mitigating Systems Cornerstone in
that it affected the mitigating availability of the train A ESCW chiller. This
finding was determined to be of very low safety significance (Green) because it
did not represent a loss of system safety function, the single train of the ESCW
system affected did not lose functionality for greater than the TS allowed outage
time, and the finding was not potentially risk-significant due to external events.
This finding is associated with the cross-cutting area of human performance
because maintenance personnel improperly executed plant procedures.
(Section 1R15)
maintain adequate procedures for the performance of maintenance on the
ESCW system chillers. Specifically, procedures lacked sufficient details to
perform maintenance on the chillers pre-rotational vane actuator. This
deficiency led to the train A ESCW system chiller being incapable of starting
and inoperable for a period of time greater than allowed by the TS.
This issue is more than minor because it affected the reliability objective of the
equipment performance attribute under the Mitigating Systems Cornerstone in
Enclosure
3
that it affected the mitigating availability of the train AESCW chiller. The finding
was determined to have potential safety significance greater than very low
because of the resultant reduced functional capability of the ESCW system to
mitigate events, and the length of time the condition existed. This significance of
this AV will remain indeterminate pending completion of the significance
determination process. A contributing cause of this issue is associated with the
cross-cutting area of human performance, in that the maintenance organization
did not generate specific, written procedures to perform ESCW maintenance.
(Section 1R15)
B. Licensee Identified Violations
None.
Enclosure
REPORT DETAILS
Summary of Plant Status
The unit began the inspection period at rated thermal power and operated at full power until
April 8, when the unit was removed from service for the commencement of Refueling Outage
13 (RFO13). The unit was returned to service on May 16, and commenced power ascension.
On May 18, the unit, at 29 percent power, was shutdown to repair a hydrogen leak on the main
generator. On May 19, the unit was returned to service, and achieved rated thermal power on
May 21. The unit operated at or near rated thermal power for remainder of the inspection
period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
When a tornado warning was issued for the site on June 11, the inspectors reviewed
actions taken by the licensee in accordance with Procedure AP-300, Severe Weather
Response, to ensure that the adverse weather conditions would neither initiate a plant
event nor prevent any structure, system or component (SSC) from performing its design
function.
After the licensee completed preparations for seasonal high temperature, the inspectors
walked down the emergency diesel generators and the high head safety injection
system. These systems were selected because their safety related functions could be
affected by adverse weather. The inspectors reviewed documents listed in the
attachment, observed plant conditions, and evaluated those conditions using criteria
documented in Procedure AP-301, Adverse Weather.
The inspectors reviewed the following action requests (ARs) associated with this area,
to verify that the licensee had identified and implemented appropriate corrective actions:
- 163469, Emergency Service Water Pump Start Due to Elevated Temperatures
- 166213, AOP-006 Entry Due to Increased Generator Temperatures
b. Findings
No findings of significance were identified.
Enclosure
5
1R04 Equipment Alignment
a. Inspection Scope
The inspectors performed the following three partial system walkdowns, while the
indicated SSC were out-of-service (OOS) for maintenance and testing:
- The A residual heat removal system with the B residual heat removal system
OOS on April 14
- The A emergency service water system with the B emergency service water
system OOS on April 15
- The condensate storage tank including AFW suction piping and valves on June
19.
To evaluate the operability of the selected trains or systems under these conditions, the
inspectors reviewed valve and power alignments by comparing observed positions of
valves, switches, and electrical power breakers to the procedures and drawings listed in
the Attachment.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
For the 20 areas identified below, the inspectors reviewed the licensees control of
transient combustible material and ignition sources, fire detection and suppression
capabilities, fire barriers, and any related compensatory measures, to verify that those
items were consistent with FSAR Section 9.5.1, Fire Protection System, and FSAR
Appendix 9.5.A, Fire Hazards Analysis. The inspectors walked down accessible portions
of each area and reviewed results from related surveillance tests, to verify that
conditions in these areas were consistent with descriptions of the applicable FSAR
sections. Documents reviewed are listed in the Attachment.
- 236', 261', and 286' levels of the fuel handling building including areas 5-F-FPP,
5-F-CHF, and 5-F-BAL (3 areas)
- All levels of the reactor containment building, fire area 1-C (1 area)
- B emergency diesel generator building including areas 1-D-1-DGB-RM, 1-D-3-
DGB-ES, 1-D-DTB, 1-D-1-DGB-ASU, 1-D-1-DGB-ER, and 1-D-3-DGB-HVR (6
areas).
- All levels of the turbine building including areas 1-G-286, 1-G-314, 1-G-240, and
1-G-261 (4 areas).
- The 261' level of the reactor auxiliary building including areas 1-A-4-COMB, 1-A-
4-COME, and 1-A-4-COMI (3 areas)
Enclosure
6
- The 305' level of the reactor auxiliary building including areas 12-A-6-HV7, 12-A-
6-CHF1, and 12-A-6-CHF2 (3 areas)
Also, to evaluate the readiness of the licensees personnel to prevent and fight fires, the
inspectors observed fire brigade performance during an unannounced fire drill in the
emergency diesel generator building on June 7.
The inspectors reviewed AR 196748196748 Fire Brigade Members Not Meeting Expectations
to verify that the licensee had identified and implemented appropriate corrective actions.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors walked down the turbine building 240' and 263' elevations containing
risk-significant SSCs which are below flood levels or otherwise susceptible to flooding
from postulated pipe breaks, to verify that the area configuration, features, and
equipment functions were consistent with the descriptions and assumptions used in
FSAR Section 3.6A.6, Flooding Analysis, and in the supporting basis documents listed
in the Attachment. The inspectors reviewed the operator actions credited in the analysis
to verify that the desired results could be achieved using the plant procedures listed in
the Attachment.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Annual Review
The inspectors visually inspected the A component cooling water heat exchanger, and
reviewed the results of Procedure EPT-163, Generic Letter 89-13 Inspections (Raw
Water Systems and Local Area Air Handler Inspection and Documentation), to verify
that any potential heat exchanger deficiencies which could degrade heat exchanger
performance were identified and properly addressed by the licensee. The inspectors
also verified that the frequency of inspection was sufficient to detect degradation prior to
loss of heat removal capability below design basis values. Documents reviewed are
listed in the Attachment.
Enclosure
7
b. Findings
No findings of significance were identified.
1R08 Inservice Inspection (ISI) Activities
.1 Piping Systems ISI
a. Inspection Scope
On April 17-28, the inspectors reviewed the implementation of the ISI program for
monitoring degradation of the reactor coolant system (RCS) boundary and the risk
significant piping system boundaries. The inspectors selected a sample of American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI
required examinations for review.
The inspectors conducted an on-site review of nondestructive examination (NDE)
activities to evaluate compliance with the TS and the applicable editions of ASME
Section V and XI (1989 Edition/No Addenda for examinations credited to the second 10-
year ISI interval), and to verify that indications and defects (if present) were
appropriately evaluated and dispositioned in accordance with the requirements of ASME
Section XI, IWB-3000 or IWC-3000 acceptance standards.
Specifically, the inspectors directly observed the electronic data collection associated
with NDE activities described below and reviewed their corresponding NDE procedures,
NDE reports, and equipment certification records, and personnel qualifications records.
- Automated ultrasonic (UT) examination data for reactor pressure vessel (RPV)
welds RVN0Z0-Outlet Nozzle to Shell @ 145deg, STHW-RV-04 - lower shell to
lower head circumferential weld, and RVNOZCI-N-05SE- inlet nozzle to shell
weld (Class 1).
- Automated eddy current (ET) examination of the RPV bottom head penetrations
- 3, #32, and #34 (Class 1).
- Remote visual examination of RPV internal surfaces including the bottom core
support lug and key-way located at 90deg, the surface of the lower head
including the bottom mounted instrument connections (BMIs), and the outlet
nozzle located at 145deg (Class 1).
The inspectors reviewed a report for the automated UT examination of RPV weld
RVN0Z0, outlet nozzle to shell @ 145deg (Class 1). The review was conducted to verify
that the evaluation and disposition of recordable indications was in accordance with the
applicable version of ASME Section XI, IWB-3000.
The inspectors reviewed a sample of welding activities performed since the beginning of
the last refueling outage for ASME Class 2 piping. The inspectors reviewed welding
procedures, welder performance qualification records, and NDE records associated with
weld AH-3-S2-FW18, 4-inch diameter butt weld, service water to containment fan
Enclosure
8
coolers, ASME Class 2. The review was conducted to verify that the Section XI
Repair/Replacement requirements were met for the weld and its subsequent repair.
b. Findings
No findings of significance were identified.
.2 Boric Acid Corrosion Control (BACC) Program
a. Inspection Scope
On April 17-28, the inspectors reviewed the licensees BACC activities to verify that
licensee commitments made in response to NRC Generic Letter 88-05 Boric Acid
Corrosion of Carbon Steel Reactor Pressure Boundary and to applicable industry
guidance documents, had been implemented. Specifically, the inspectors performed an
on-site record review of procedures and the results of the licensees mode 3
containment walkdown inspection after RFO 13 was completed. The inspectors also
conducted an independent walkdown of the reactor building to evaluate compliance with
licensee BACC program requirements and to verify that degraded or non-conforming
conditions, such as boric acid leaks identified during the mode 3 containment walkdown,
were properly identified and corrected in accordance with the CAP.
The inspectors reviewed a sample of engineering evaluations completed for evidence of
boric acid found on systems containing borated water to verify that the minimum design
code required section thickness had been maintained for the affected components.
Specifically, the inspectors reviewed the following evaluations:
- AR 190572190572 Brown boric acid at valve packing (multiple valves on SI
b. Findings
No findings of significance were identified.
.3 Steam Generator Tube ISI
a. Inspection Scope
From April 24-28, the inspectors reviewed the Unit 1 steam generator (SG) tube eddy
current testing (ECT) examination activities to ensure compliance with the TS,
applicable industry operating experience and technical guidance documents, and 1989
Edition with no addenda ASME Code Section XI requirements.
The inspectors reviewed licensee SG inspection activities to ensure that ECT
inspections conducted during RFO-13 conformed to the Steam Generator Integrity
Program. The inspectors reviewed the SG examination scope, ECT acquisition
Enclosure
9
procedures, examination technique specification sheets (ETSS), ECT analysis
guidelines, the current SG specific assessment of potential degradation mechanisms,
SG Operational Assessment and Condition Monitoring documents from the previous
Unit 1 outage, and the current SG tube plugging and stabilization procedures. The
inspectors reviewed documentation to ensure that the ECT probes and equipment
configurations used were qualified to detect the expected types of SG tube degradation
in accordance with Appendix H, Performance Demonstration for Eddy Current
Examination of EPRI Pressurized Water Reactor Steam Generator Examination
Guidelines: Revision 6." Additionally, the inspectors reviewed the qualification and
certification records for the ECT standards, SG tube plugs, SG tube stabilizers, and
ECT data analysis and resolution analysis personnel.
The secondary side water chemistry and loose parts monitoring programs were
reviewed to ensure they were consistent with applicable industry guidance documents.
The inspectors independently reviewed the licensees secondary side visual examination
results and associated evaluations for loose parts that are not retrievable and will remain
in the SGs during the next operating cycle. The inspectors observed ECT acquisition,
resolution analysis, tube stabilization, and tube plugging activities.
b. Findings
No findings of significance were identified.
.4 Identification and Resolution of Problems
The inspectors performed a review of ISI related problems, BACC and SG ISI, that were
identified by the licensee and entered into the CAP as nuclear condition reports (NCR)
documents. The inspectors reviewed the NCRs to confirm that the licensee had
appropriately described the scope of the problem and had initiated corrective actions.
The inspectors performed this review to verify compliance with 10CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, requirements. The corrective action
documents reviewed by the inspectors are listed in the Attachment.
1R11 Licensed Operator Requalification
a. Inspection Scope
On June 13, the inspectors observed licensed-operator performance during
requalification simulator training, to verify that operator performance was consistent with
expected operator performance, as described in the training exercise guide. This
training tested the operators ability to place the plant in a safe condition after a station
blackout. The inspectors focused on clarity and formality of communication, the use of
procedures, alarm response, control board manipulations, group dynamics and
supervisory oversight. The inspectors observed the post-exercise critique to verify that
the licensee had identified deficiencies and discrepancies that occurred during the
simulator training.
Enclosure
10
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
.1 Routine Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed two degraded conditions listed below to verify the licensees
handling of these conditions in accordance with 10CFR50, Appendix B, Criterion XVI,
Corrective Action, and 10CFR50.65, Maintenance Rule. Documents reviewed are listed
in the Attachment.
- Local leak rate test (LLRT) failures during RFO13
- Functional failures of containment isolation valves (Target Rock)
The inspectors focused on the following attributes:
- Appropriate work practices,
- Identifying and addressing common cause failures,
- Scoping in accordance with 10 CFR 50.65(b),
- Characterizing reliability issues (performance),
- Charging unavailability (performance),
- Trending key parameters (condition monitoring),
- 10 CFR 50.65(a)(1) or (a)(2) classification and reclassification, and
- Appropriateness of performance criteria for SSCs/functions classified (a)(2)
and/or appropriateness and adequacy of goals and corrective actions for
SSCs/functions classified (a)(1).
The inspectors reviewed the following ARs associated with this area to verify that the
licensee had identified and implemented appropriate corrective actions:
- 190624, Maintenance Rule Reclassifications
- 196750, Target Rock Position Indication a1 Goal Exceeded
b. Findings
No findings of significance were identified.
.2 Periodic Evaluation (Triennial)
a. Inspection Scope
On June 26-30, the inspectors reviewed the licensees Maintenance Rule (MR) periodic
assessment, Maintenance Rule Cycle 12, Periodic a(3) Assessment, to assess the
Enclosure
11
effectiveness of the assessment and verify that it was issued in accordance with the
requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants. The inspectors review included an evaluation
of: periodic assessment timeliness, balancing of reliability and unavailability, (a)(1)
activities, (a)(2) activities, and use of industry operating experience for the 18 month-
period covered by the assessment. The inspectors reviewed selected MR activities
covered by the assessment period for the following MR a(1) status component and
attendant systems: 6.9 KV AC distribution, ESCW system, 250 VDC distribution,
containment isolation, and normal service water. Additionally, the inspectors conducted
a plant walkdown to assess the condition of risk significant plant structures within the
scope of the MR to verify that condition monitoring was adequately performed.
The inspectors reviewed selected plant work order data, self assessments, system
health reports, reliability and unavailability monitoring status documents, significant
adverse condition investigation reports, MR system scoping documents, and attendant
MR expert panel meeting minutes. The inspectors also discussed and reviewed
relevant corrective action reports, and discussed issues with system engineers and
licensee management. In addition, the inspectors attended a Plant Nuclear Safety
Committee meeting to assess the management approval for the transition of 250 VDC
distribution and main feedwater systems from MR status a(1) to a(2). Specific
documents reviewed are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The inspectors reviewed the licensees risk assessments and the risk management
actions for the plant configurations associated with the four activities listed below. The
inspectors verified that the licensee had performed adequate risk assessments, and
implemented appropriate risk management actions when required by 10CFR50.65(a)(4).
For emergent work, the inspectors also verified that any increase in risk was promptly
assessed, and that appropriate risk management actions were promptly implemented.
- Week of May 15 including reactor startup for low power physics testing, reactor
shutdown, and various other maintenance activities.
- Severe thunderstorm warning on May 26.
- Emergent repairs to the ESCW chilled water system chiller, WC-2A on June 2.
- Tornado warning on June 11.
b. Findings
No findings of significance were identified.
Enclosure
12
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed four operability determinations addressed in the ARs listed
below. The inspectors assessed the accuracy of the evaluations, the use and control of
any necessary compensatory measures, and compliance with the TS. The inspectors
verified that the operability determinations were made as specified by Procedure OPS-
NGGC-1305, "Operability Determinations." The inspectors compared the justifications
made in the determination to the requirements from the TS, the FSAR, and associated
design-basis documents, to verify that operability was properly justified and the subject
component or system remained available, such that no unrecognized increase in risk
occurred:
Engaged
- 196258, A-SA Chiller Failed to Start
- 196857, Heat Load Calculation Does Not Bound Fuel Shipped to HNP
b. Findings
.1 Failure to perform maintenance on valve 1SW-1055 in accordance with maintenance
procedure
Introduction. A Green self-revealing NCV of TS 6.8.1, Written Procedures was
identified for a failure to follow a procedure used to adjust the linkage between a service
water control valve and its actuator, which ultimately resulted in rendering one ESCW
train inoperable.
Description. While in mode 6 during RFO13, preventive maintenance was conducted on
service water control valve 1SW-1055 in late April, with the work completed on April 30.
This valve utilizes an electric-hydraulic actuator to regulate service water flow to the train
A ESCW chiller condenser. The maintenance included valve and actuator separation
and valve disassembly as documented in work order #816771. Following valve and
actuator reassembly, while adjusting the valve stroke, using Procedure PIC-I058,
Calibration of a ITT Milliampere Hydramotor Actuator Model NH-92 & 94, Revision 11,
adequate thread engagement was not maintained on a fastener that connects the
service water discharge valve actuator to the valve body. This procedure compliance
error ultimately resulted in the valve becoming separated from its actuator and the train
A ESCW chiller being inoperable for approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (TS allowed outage time
is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
The train A ESCW chiller is required by TS 3.7.13, Essential Services Chilled Water
System, to be operable in modes 1 through 4. On May 10, the plant entered mode 4
and on May 16, the plant entered mode 1. The valve actuator for 1SW-1055 separated
from the valve body of 1SW-1055 on May 16 and was repaired later the same day. The
train A ESCW chiller functioned correctly for approximately 8.5 days of cumulative
runtime prior to valve 1SW-1055 becoming separated.
Enclosure
13
Analysis. On April 30, Procedure PIC-I058 was not followed during adjustment of the
linkage between service water control valve 1SW-1055 and its actuator, contributing to
adequate thread engagement not being maintained between the valve and its actuator.
This procedure adherence problem ultimately resulted in train A ESCW chiller being
inoperable for approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This issue is a performance deficiency
associated with the Mitigating Systems Cornerstone. The finding is greater than minor
since it affects the Mitigating Systems Cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e. core damage), and is associated with the cornerstone
attribute of equipment performance.
The risk significance of this issue was evaluated using NRC Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-
Power Situations because the train A ESCW chiller is only required to be operable in
modes 1 through 4, after plant heat up has started and residual heat removal has been
secured. This finding was determined to be of very low safety significance (Green)
because the finding did not represent a loss of system safety function, the ESCW train
affected did not lose functionality for greater than the TS allowed outage time, and the
finding was not potentially risk-significant due to external events. This finding is
associated with the cross-cutting area of human performance because maintenance
personnel improperly executed plant procedures.
Enforcement. TS 6.8.1 states, in part, that procedures shall be established,
implemented, and maintained covering activities referenced in appendix A of Regulatory
Guide 1.33, revision 2. Regulatory Guide 1.33, revision 2, part 9 states that
maintenance that can affect the performance of safety-related equipment should be
properly planned and performed in accordance with written procedures, documented
instructions, or drawings appropriate to the circumstances. Procedure PIC-I058,
Revision 11 provides instructions for adjusting the valve stroke for valve 1SW-1055
during maintenance. Contrary to above, on April 30, the adjustment to valve 1SW-1055
was not adequately conducted, in that the valve actuator was not properly secured to
the valve body, resulting in the valve actuator separating from the valve body on May
16. However, because this violation is of the very low safety significance, the issue was
entered into the CAP (AR 194627194627, and the deficient condition was promptly corrected,
this finding is being treated as an NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy: NCV 05000400/2006003-01, Failure to Follow Procedure During
Service Water Control Valve Preventive Maintenance.
.2 Inadequate maintenance procedure for replacement of train A ESCW chiller pre-
rotational vane actuator
Introduction. Subsequent to the ESCW chiller maintenance activity (with related NCV)
conducted in late April, and discussed in Section 1R15.b.1, additional, but different,
maintenance was conducted on the same chiller in early May. For this maintenance
work, a self-revealing AV was identified for a failure to maintain adequate procedures
that specified the required torque of a threaded fastener which holds the chiller
Enclosure
14
compressor pre-rotational vane shaft to the pre-rotational vane actuator linkage arms.
Because of this lack of guidance, the chiller maintenance was completed with the chiller
compressor threaded fastener under-torqued, which led to the fastener becoming loose
during chiller operation, ultimately resulting in the chiller failing to start on June 1.
Description. While in mode 6 during RFO13, maintenance was conducted on the train
A ESCW system chiller in early May, with the work completed on May 4. This
maintenance included replacing the chiller compressor and pre-rotational vane actuator
as documented in work order #664677. The pre-rotational vane shaft is integral to the
compressor and connected to the pre-rotational vane actuator via linkage arms. The
linkage arms are connected to the pre-rotational vane shaft with a threaded fastener.
During normal operation when the chiller is shut down, the pre-rotational vanes move to
the minimum load position and linkage arm rotation causes a limit switch to make-up.
This limit switch must be made-up to satisfy a chiller electrical start interlock.
After the compressor and pre-rotational vane actuator was replaced, the threaded
fastener connecting the pre-rotational vane shaft to the linkage arms was not
adequately tightened. Since the threaded fastener wasnt properly tightened, it
loosened during subsequent chiller operation. Eventually, the fastener became too
loose and the linkage arms and pre-rotational vane shaft lost synchronization. The train
A ESCW chiller was successfully started on May 25 and run for approximately one
hour. The train A ESCW chiller then failed to start on the next start attempt on June 1.
Since the pre-rotational vanes do not move when the chiller is shutdown, inspectors
determined that the pre-rotational vane shaft and linkage arms lost synchronization
during chiller operation on May 25. At that time, the linkage arms were not correctly
positioned to make-up the chiller start interlock, and therefore the chiller was inoperable
starting at 9:16 p.m. on May 25.
The original equipment manufacturer of the train A ESCW chiller is the York
International Division of Borg-Warner Corporation. York maintenance procedures
specify that 75 foot-pounds of torque be used to tighten the threaded fastener which
connects the pre-rotational vanes to the actuator linkage arms. Despite this guidance,
the licensee did not develop adequate maintenance procedures to ensure that the
threaded fastener was properly tightened and therefore licensee maintenance personnel
did not apply the proper torque to the fastener when performing the maintenance.
Analysis. For the train A ESCW chiller compressor replacement in RFO13,
maintenance procedures were inadequate in that they did not specifically direct
maintenance personnel to tighten the pre-rotational vane shaft to linkage arm fastener
to 75 ft-lb, which is a performance deficiency associated with the Mitigating Systems
Cornerstone. This issue is more than minor because it affected the reliability objective of
the equipment performance attribute under the Mitigating Systems Cornerstone in that it
affected the mitigating availability of the train A ESCW chiller. The finding was
determined to have potential safety significance greater than very low because of the
resultant reduced functional capability of the ESCW system to mitigate events, and
because the degradation existed for approximately seven days. The finding did not
present an immediate safety concern in that the fully redundant train B of ESCW
Enclosure
15
remained operable or available during the train A period of inoperability. A contributing
cause of this issue is associated with the cross-cutting area of human performance, in
that the maintenance organization did not generate specific, written procedures to
perform ESCW maintenance.
Enforcement. TS 6.8.1 states, in part, that procedures shall be established,
implemented, and maintained covering activities referenced in appendix A of Regulatory
Guide 1.33, revision 2. Regulatory Guide 1.33, revision 2, part 9 states that
maintenance that can affect the performance of safety-related equipment should be
properly planned and performed in accordance with written procedures, documented
instructions, or drawings appropriate to the circumstances. Contrary to this,
maintenance procedures did not adequately address torque requirements for the train
A ESCW chiller pre-rotational vane shaft to actuator linkage fastener, leading to the
train A ESCW chiller being inoperable for a period of time in excess of that permitted
by the TS. Pending determination of safety significance, this finding is identified as AV:
AV05000400/2006003-02, Failure to Maintain Adequate Procedures Such That a
Required Torque Was Not Provided for a Threaded Fastener on an ESCW System
Chiller.
1R19 Post Maintenance Testing
a. Inspection Scope
For the five post-maintenance tests listed below, the inspectors witnessed the test
and/or reviewed the test data, to verify that test results adequately demonstrated
restoration of the affected safety function(s) described in the FSAR and the TS. The
tests included the following:
- CM-E0020, Replace Oil Filled Electrolytic Capacitors and/or Ferro-Resonant
Transformer Assembly and Tune Westinghouse 7.5 KVA Static Inverters, for
testing following maintenance on vital inverter II on April 21.
- OST-1073, 1B-SB Emergency Diesel Generator Monthly Operability Test,
Modes 1 - 6, for post-outage testing on April 22.
- OST-1040, Essential Services Chilled Water Systems Operability Quarterly
Interval Modes 1 - 6, after the B train P-4 pump was rebuilt on April 24.
pump replacement on May 2.
- OST-1214, Emergency Service Water System Operability Train A Quarterly
Interval Modes 1 - 4, after the A emergency service water pump was replaced
on May 3.
b. Findings
No findings of significance were identified.
Enclosure
16
1R20 Refueling and Outage Activities
.1 Review of Outage Plan
a. Inspection Scope
Prior to the RFO-13, the inspectors reviewed the licensees outage risk control plan to
verify that the licensee had performed adequate risk assessments, and had
implemented appropriate risk management strategies when required by
b. Findings
No findings of significance were identified.
.2 Monitoring of Shutdown Activities
a. Inspection Scope
The inspectors observed portions of the cooldown process to verify that TS cooldown
restrictions were followed.
b. Findings
No findings of significance were identified.
.3 Licensee Control of Outage Activities
a. Inspection Scope
The inspectors observed the items or activities described below, to verify that the
licensee maintained defense-in-depth commensurate with the outage risk control plan
for key safety functions and applicable TS requirements when taking equipment OOS.
The inspectors reviewed the licensees responses to emergent work and unexpected
conditions to verify that resulting configuration changes were controlled in accordance
with the outage risk control plan, and to verify that control room operators were kept
cognizant of plant configuration.
- clearance activities
- RCS instrumentation
- electrical power
- spent fuel pool cooling
- inventory control
- reactivity control
- containment closure
Enclosure
17
b. Findings
No findings of significance were identified.
.4 Reduced Inventory Conditions
a. Inspection Scope
The inspectors reviewed the licensees commitments from Generic Letter 88-17, and
confirmed by sampling that those commitments were still in place and adequate.
Periodically during the reduced inventory conditions, the inspectors reviewed system
lineups to verify that the configuration of the plant systems were in accordance with
those commitments.
b. Findings
No findings of significance were identified.
.5 Refueling Activities
a. Inspection Scope
The inspectors observed fuel handling operations (removal, inspection, and insertion)
and other ongoing activities, to verify that those operations and activities were being
performed in accordance with the TS and approved procedures. Also, the inspectors
observed refueling activities to verify that the location of the fuel assemblies was
tracked, including new fuel, from core offload through core reload.
b. Findings
No findings of significance were identified.
.6 Monitoring of Heatup and Startup Activities
a. Inspection Scope
Prior to mode changes and on a sampling basis, the inspectors reviewed system lineups
and/or control board indications to verify that TS, license conditions, and other
requirements, commitments, and administrative procedure prerequisites for mode
changes were met prior to changing modes or plant configurations. Also, the inspectors
periodically reviewed RCS boundary leakage data, and observed the setting of
containment integrity, to verify that the RCS and containment boundaries were in place
and had integrity when necessary. Prior to reactor startup, the inspectors walked down
containment to verify that debris has not been left which could affect performance of the
containment sumps. The inspectors reviewed reactor physics testing results to verify
that core operating limit parameters were consistent with the design.
Enclosure
18
b. Findings
No findings of significance were identified.
.7 Identification and Resolution of Problems
a. Inspection Scope
Periodically, the inspectors reviewed the items that had been entered into the CAP, to
verify that the licensee had identified problems related to outage activities at an
appropriate threshold and had entered them into the CAP. For the significant problems
documented in the CAP and listed below, the inspectors reviewed the results of the
licensees investigations, to verify that the licensee had determined the root cause and
implemented appropriate corrective actions, as required by 10CFR50, Appendix B,
Criterion XVI, Corrective Action.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
For the six surveillance tests identified below, the inspectors witnessed testing and/or
reviewed test data, to verify that the SSCs involved in these tests satisfied the
requirements described in the TS and the FSAR, and that the tests demonstrated that
the SSCs were capable of performing their intended safety functions.
- OST-1824, 1B-SB Emergency Diesel Generator Operability Test 18 Month
Interval Modes 1 through 6 and Defueled on April 8 .
- OST-1813, Remote Shutdown System Operability 18 Month Interval Modes 5, 6
or Defueled, on April 11.
- MST-E0013, 1E Battery Performance Test on April 18.
- * EST-213, ASME System Pressure Test for Fuel Oil Piping on May 2.
- OST-1826, Safety Injection: ESF Response Time, Train B 18 Month Interval on
a Staggered Test Basis Mode 5-6, on May 8.
- ** EST-212, Type C Local Leak Rate Tests
- This procedure included inservice testing requirements.
- This procedure included testing of a large containment isolation valve.
b. Findings
No findings of significance were identified.
Enclosure
19
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed the temporary modification described in Engineering Change
64090R1, to verify that the modification did not affect the safety functions of important
safety systems, and to verify that the modification satisfied the requirements of
10CFR50, Appendix B, Criterion III, Design Control.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed two operations simulator examinations conducted on June 20
and June 27, to verify licensees self-assessment of classification, notification, and
protective action recommendation development in accordance with 10CFR50, Appendix
E.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstones: Occupational Radiation Safety and Public Radiation Safety
2OS1 Access Controls To Radiologically Significant Areas
a. Inspection Scope
Access Controls The inspectors evaluated licensee activities for monitoring and
controlling worker access to radiologically significant areas, focusing on those activities
associated with RFO13. The inspection included direct observation of administrative
and physical controls, appraisal of the knowledge and proficiency of radiation workers
and health physics technicians (HPTs) in implementing radiological controls, and review
of the adequacy of procedural guidance and its implementation.
The inspectors reviewed licensee procedures regarding access control to radiologically
significant areas. Selected procedural details for posting, surveying, and access control
to airborne radioactivity, radiation area, high radiation area (HRA), locked high radiation
area (LHRA), and very high radiation area (VHRA) locations were reviewed and
Enclosure
20
discussed with cognizant licensee representatives. The inspectors reviewed
administrative guidance documents and procedures for control of non-fuel radioactive
material stored in the spent fuel pools, and evaluated several radiation work permits
(RWPs) used for work in radiologically significant areas associated with RFO-13. The
selected RWPs were assessed for adequacy of access controls and specified electronic
dosimeter (ED) alarm setpoints against expected work area dose rates and work
conditions. Access control procedures for posted LHRA and VHRA locations were
reviewed and discussed with selected Radiation Protection (RP) management,
supervision, and technicians.
During facility tours, the inspectors evaluated selected radiological postings, barricades,
and surveys associated with radioactive material storage areas and radiologically
significant areas within the reactor containment building, reactor auxiliary building, waste
processing building, and fuel handling building. The inspectors conducted independent
dose-rate measurements at various building locations and compared those results to
licensee radiation survey map data. The surveyed locations included the lower refuel
cavity, the personnel airlock area, the train B RHR pump valve chamber, and the seal
table room. The inspectors independently assessed implementation of LHRA controls,
and evaluated the adequacy of the licensees LHRA and VHRA key controls through
procedural reviews and supervisory interviews.
During the inspection, the proficiency and knowledge of the radiation workers and RP
staff in communicating and applying radiological controls for selected tasks were
evaluated. The inspectors attended briefings for work activities associated with RWPs
3438 and 3596 (reactor head/core activities and SG sludge lance activities).
Radiological worker and HPT training/skill levels, procedural adherence, and
implementation of RWP-specified access controls, including those associated with
changing radiological conditions, were observed and evaluated by the inspectors during
selected job site reviews and tours within the radiological control area. In addition, the
inspectors interviewed selected management personnel regarding radiological controls
associated with RFO-13 activities.
RP activities were evaluated against Updated Final Safety Analysis Report (UFSAR)
Section 12, Radiation Protection; TS 6.12, High Radiation Area; 10 CFR 19.12; 10 CFR Part 20, Subparts B, C, F, G, H, and J; and approved procedures. The procedures and
records reviewed are listed in the Attachment.
Problem Identification and Resolution CAP NCR documents associated with access
control to radiologically significant areas, radiation worker performance, and HPT
proficiency were reviewed and assessed. The NCRs for this program area, listed in the
Attachment, were reviewed and evaluated in detail. The inspectors assessed the
licensees ability to identify, characterize, prioritize, and resolve the identified issues in
accordance with approved CAP procedures.
The inspectors completed 21 of the required 21 samples for Inspection Procedure (IP) 71121.01. All samples have now been completed for this IP.
Enclosure
21
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls
a. Inspection Scope
As Low As Reasonably Achievable (ALARA) Implementation of the licensee's ALARA
program during RFO-13 was observed and evaluated by the inspectors. The inspectors
reviewed ALARA planning, dose estimates, and prescribed ALARA controls for the five
outage work tasks expected to incur the maximum collective exposures. Reviewed
activities included installation of temporary/emergent lead shielding for the reactor head
stand on the refueling floor in containment, installation and removal of insulation,
decontamination activities, reactor headwork activities, incore instrument work and other
various work activities associated with RFO-13. Also, incorporation of planning,
established work controls, expected dose rates and dose expenditure into the ALARA
pre-job briefings and RWPs for those activities were reviewed. The inspectors also
independently verified that selected job site dose rates were consistent with the dose
rates recorded on pre-job survey maps for containment and auxiliary building work areas
and equipment. The inspectors made direct field or closed-circuit-video observations of
work activities associated with the reactor head lift and pulling of the thimbles in the seal
table room, and evaluated the licensees use of engineering controls, low dose waiting
areas, and on-the-job supervision for selected activities that were conducted in the
reactor containment building.
Selected elements of the licensee's source term reduction and control program were
examined to evaluate the effectiveness of the program in supporting implementation of
the ALARA program goals. Reviewed areas included primary chemistry shutdown
controls, radiation field monitoring and trending, and temporary/emergent shielding.
Trends in individual and collective personnel exposures at the facility were reviewed.
Records of year-to-date individual radiation exposures sorted by work groups were
examined for significant variations of exposures among workers. Exposure tracking
during RFO-13, and records of exposures to declared pregnant workers incurred from
November 2004 through March 2006 as well as associated guidance for controlling such
exposures, were also reviewed. Trends in the plants three-year rolling average
collective exposure history, outage, non-outage and total annual doses were reviewed
and discussed with licensee representatives.
The licensee's ALARA program implementation and practices were evaluated for
consistency with UFSAR Chapter 12, Sections 1-5, Radiation Protection; 10 CFR
Part 20 requirements; Regulatory Guide 8.29, Instruction Concerning Risks from
Occupational Radiation Exposure, February 1996; and licensee procedures.
Documents reviewed are listed in the Attachment.
Problem Identification and Resolution The inspectors reviewed NCR documents and
audits listed in the Attachment that are related to the ALARA program. The inspectors
Enclosure
22
assessed the licensees ability to identify, characterize, prioritize, and resolve the
identified issues in accordance with CAP-NGGC-0200, Corrective Action Program, Rev.
16.
The inspectors completed 15 of the required 15 samples for IP 71121.02. All samples
have now been completed for this IP.
b. Findings
No findings of significance were identified.
2PS2 Radioactive Material Processing and Transportation
a. Inspection Scope
Waste Processing and Characterization The inspectors evaluated licensee methods for
processing and characterizing radioactive waste (radwaste). Inspection activities
included direct observation of processing equipment for solid and liquid radwaste and
evaluation of waste stream characterization data.
Solid and liquid radwaste equipment was inspected for material condition, configuration
compliance with the UFSAR, and consistency with Process Control Program (PCP)
requirements. Inspected equipment included liquid radwaste hold-up tanks; resin
transfer piping; abandoned waste evaporators; remote operating equipment for
packaging filters, and elements of the modular fluidized transfer demineralization
system. The inspectors discussed system changes, component function, and
equipment operability with licensee staff. In addition, procedural guidance for resin
transfer was evaluated and compared with current equipment configuration.
Licensee radionuclide characterizations for selected waste streams were reviewed and
discussed with radwaste staff. For primary resin, radwaste filters, and dry active waste
the inspectors evaluated analyses for hard-to-detect nuclides and appropriate use of
scaling factors. Comparison results between licensee waste stream characterization
data and outside laboratory data were reviewed for 2005 and 2006. For selected
shipment records, waste classification calculations were independently performed and
the methodology used for resin waste stream mixing and concentration averaging was
evaluated. The inspectors also interviewed radwaste staff and reviewed procedural
guidance to evaluate the licensees program for monitoring changing operational
parameters.
Radwaste processing activities were reviewed for consistency with the licensees PCP,
Rev. 8; and UFSAR, Chapter 11, Amendment 52. Waste stream characterization
analyses were reviewed against regulations detailed in 10 CFR Part 61.55 and guidance
provided in the Branch Technical Position on Waste Classification and Waste Form,
1983. Reviewed documents are listed in the Attachment.
Enclosure
23
Transportation The inspectors evaluated the licensees activities related to
transportation of radioactive material. The evaluation included direct observation of
shipment preparation activities and review of shipping related documents.
The inspectors directly observed transportation activities including the shipment of
several containers of vendor tools and equipment associated with split pin inspection,
pressurizer safety relief valve, and snubber. The inspectors observed placarding of the
shipment vehicles and marking and labeling of the shipment packages. The inspectors
observed technicians performing radiation and contamination surveys on packages and
vehicles.
As part of the document review, the inspectors evaluated five shipping records for
consistency with licensee procedures and compliance with NRC and DOT regulations.
In addition, training records for individuals currently qualified to ship radioactive material
were checked for completeness and the training curriculum provided to these workers
was evaluated. Documents reviewed are listed in the Attachment.
Transportation program implementation was reviewed against regulations detailed in
10 CFR Parts 20 and 71, 49 CFR Parts 170-189; as well as the guidance provided in
NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
Problem Identification and Resolution Selected NCR documents associated with
radwaste processing and transportation were reviewed and discussed with cognizant
licensee representatives. The inspectors assessed the licensees ability to characterize,
prioritize, and resolve the identified issues in accordance with Procedure CAP-NGGC-
0200, Corrective Action Program, Rev. 16. Reviewed documents are listed in the
Attachment.
The inspectors completed 6 of the required 6 samples for IP 71122.02. All samples
have now been completed for this IP.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (PI) Verification
a. Inspection Scope
The inspectors sampled licensee data from the period January 2005 through April 2006
for the performance indicators (PIs) listed below. To verify the accuracy of the reported
PI data, PI definitions and guidance contained in NEI 99-02, Regulatory Assessment
Indicator Guideline, Rev. 4, were used to verify the basis for each data element.
Enclosure
24
Cornerstone: Occupational Radiation Safety
- Occupational Exposure Control Effectiveness PI
For the reviewed period, the inspectors assessed CAP records to determine whether
HRA, VHRA, or unintended radiation exposures, resulting in TS or 10 CFR 20 non-
conformances, had occurred. In addition, the inspectors reviewed selected personnel
contamination event data, internal dose assessment results, and ED alarms associated
with dose rates exceeding 1 rem/hr and cumulative dose rates exceeding established
set-points from January 2005 through April 2006. Reviewed documents relative to this
PI are listed in the Attachment.
Cornerstone: Public Radiation Safety
- Radiological Effluent Technical Specification/Offsite Dose Calculation Manual
Radiological Effluent Occurrences PI
The inspectors reviewed OOS effluent monitor logs and six effluent release permits. The
inspectors reviewed documents listed in the Attachment.
The inspectors completed 2 of the required 2 samples for IP 71151.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Routine Review of ARs
To aid in the identification of repetitive equipment failures or specific human
performance issues for followup, the inspectors performed frequent screenings of items
entered into the CAP. The review was accomplished by reviewing daily AR reports.
.2 Annual Sample Review
a. Inspection Scope
The inspectors selected AR 141462141462for detailed review. This AR was associated with
degraded performance of the M-7 containment penetration. The inspectors reviewed
this report to verify that the licensee had identified the full extent of the issue, performed
an appropriate evaluation, and specified and prioritized appropriate corrective actions.
The inspectors evaluated the report against the requirements of the CAP as delineated
in corporate Procedure CAP-NGGC-0200, Corrective Action Program, and 10 CFR 50,
Appendix B.
Enclosure
25
b. Findings
No findings of significance were identified.
.3 Semi-Annual Trend Review
c. Inspection Scope
The inspectors performed a review of the CAP and associated documents to identify
trends that could indicate the existence of a more significant safety issue. The
inspectors review was focused on repetitive equipment issues, but also considered the
results of inspector CAP item screenings, licensee trending efforts, and licensee human
performance results. The inspectors review nominally considered the six-month period
of January through June, although some examples expanded beyond those dates when
the scope of the trend warranted. The review also included issues documented outside
the normal CAP in system health reports, self assessment reports, and Maintenance
Rule assessments. The specific items reviewed are listed in the Attachment. The
inspectors compared and contrasted their results with the results contained in the
licensees latest semi-annual trend reports. The inspectors also evaluated the
licensees trend reports against the requirements of the CAP as specified in CAP-
NGGC-0200, Corrective Action Program.
b. Assessment and Observations
There were no findings of significance identified. The inspectors observed that the
licensee performed adequate trending reviews. The licensee routinely reviewed cause
codes, involved organizations, key words, and system links to identify potential trends in
the CAP data. The inspectors compared the licensees process results with the results
of the inspectors daily screening and did not identify any discrepancies or potential
trends in the CAP data that the licensee had failed to identify.
4OA5 Other Activities
.1 (Closed) NRC Temporary Instruction 2515/150, Reactor Pressure Vessel Head and
Head Penetration Nozzles (NRC Order EA-03-009) (Unit 1)
a. Inspection Scope
From April 17-21 the inspectors reviewed the licensees activities relative to the NDE of
the reactor pressure vessel head (RPVH) nozzles and the visual examination to identify
potential boric acid leaks from pressure-retaining components above the RPVH in
response to NRC Bulletins 2001-01, 2002-01, 2002-02, and NRC Order EA-03-009
Modifying Licenses dated February 20, 2004 (NRC Order).
The inspectors review of the NDE of RPVH nozzles included: a) review of NDE
procedures, b) assessment of NDE personnel training and qualification, c) review of
NDE equipment certification, and d) observation and assessment of UT and ET
examinations. The inspectors also held interviews with contractor representatives
(Wesdyne) and other licensee personnel involved in the RPVH examination.
Enclosure
26
The activities were reviewed to verify licensee compliance with the regulatory
requirements of the NRC Order and gather information to help the NRC staff identify
possible further regulatory positions and generic communications.
Specifically, the inspectors reviewed a sample of the results from the volumetric UT and
surface ET examinations of RPVH nozzles as follows:
- Observed a portion of in-process UT/ET scanning of RPVH nozzle Nos. 3, 32,
and 24
- Reviewed the UT/ET reports and electronic data for RPVH nozzle Nos. 4, 6, 18,
23, 38, 43, and 51
- Reviewed the UT/ET reports for RPVH nozzle Nos. 19 and 20
- Reviewed the results of the UT examination performed to assess for leakage into
the annulus between the RPVH penetration nozzle and the RPVH low-alloy steel
(interference fit zone) for all penetration numbers listed above
- Reviewed training and qualification records, including qualification and
certification procedures, for NDE personnel who performed the above volumetric
and surface examinations
- Reviewed certification and calibration records for NDE equipment used to
perform the above volumetric and surface examinations.
- Reviewed Wesdyne examinations procedures used to perform the above
volumetric and surface examinations
The inspectors also reviewed the procedures and the results for visual examinations
performed to identify potential boric acid leaks from pressure-retaining components
above the RPVH.
b. Observations and Findings
1) Verification that the examinations were performed by qualified and
knowledgeable personnel.
The inspectors found that volumetric and surface NDEs were performed in accordance
with approved and demonstrated procedures with trained and qualified examination
personnel. All examiners were qualified in accordance with the ASME Code and had
significant experience, including experience examining RPVHs. In addition to
qualification to Code requirements, UT and ET personnel had additional training on
RPVH examination.
2) Verification that the examinations were performed in accordance with approved
and demonstrated procedures.
Enclosure
27
The Harris Unit 1 RPVH has 52 control rod drive mechanism (CRDM) nozzles with
thermal sleeves, 4 instrument column nozzles, 8 spare penetration nozzles, 1 reactor
vessel level indicator system (RVLIS) nozzle, and one vent nozzle, for a total of 66
nozzles.
All penetration nozzles, except the vent nozzle, were examined by remote automated
UT and ET examination from the inside diameter (ID) surface in accordance with
Wesdyne approved and demonstrated Procedures WDI-SSP-1013, WDI-SSP-1016,
WDI-SSP-1014, WDI-SSP-1017, and WDI-SSP-1025. The nozzles were examined
using a blade UT/ET probe (for nozzles with thermal sleeve only) and a rotating UT/ET
probe (for nozzles without thermal sleeves). Each type of probe was mounted in a
single examination module and scanning was performed axially (vertical up and vertical
down). For nozzles with thermal sleeves, the examination employed the time of flight
diffraction (TOFD) technique using a blade probe containing one set of 44 degree/6
MHz/L-wave transducers (oriented vertically), and a 0 degree/2.5 MHz/L-wave pulse-
echo transducer for assessment of leakage into the interference fit zone. For nozzles
without thermal sleeves (open housing), except the vent nozzle, the examination
employed the TOFD technique using two sets of 55 degree/5 MHz/L-wave transducers
(one set oriented vertically for circumferential flaws and the other oriented horizontally
for axial flaws), and a 0 degree/2.25 MHz/L-wave pulse-echo transducer for assessment
of leakage into the interference fit zone.
The vent nozzle was examined by manual ET examinations in accordance with
Wesdyne approved and demonstrated Procedures WDI-STD-101 and WDI-STD-114.
The vent nozzle inside surface was ET examined with a multi plus-point coils array
probe combined with a bobbin probe. The surface of the vent nozzles J-groove weld
was ET examined using a multi plus-point coils array probe to assess leakage through
the J-groove weld.
The inspectors reviewed Wesdyne procedures and observed in-process examinations to
verify that activities were performed in accordance with approved and demonstrated
procedures. The inspectors found that Wesdyne examination procedures were
demonstrated to examine and detect flaws in the RPVH nozzles, as documented in the
Electric Power Research Institute (EPRI), Materials Reliability Program (MRP)
documents listed in the Attachment. Approved acceptance criteria and critical
parameters for RPVH leakage were applied in accordance with these demonstrated
procedures.
3) Verification that the licensee was able to identify, disposition, and resolve
deficiencies.
All indications of cracks or interference fit zone leakage were required to be reported for
further examination and disposition. Based on observation of the examination process,
the inspectors considered deficiencies would be appropriately identified, dispositioned,
and resolved.
UT indications associated to the geometry of the examined volume were identified at
several J-groove welds. All indications did not exhibit crack characteristics and were
dispositioned as metallurgical/geometrical indications (not service related).
Enclosure
28
4) Verification that the licensee was capable of identifying the primary water stress
corrosion cracking (PWSCC) and/or RPVH corrosion phenomenon described in
the NRC Order.
The NDE techniques employed for the examination of RPVH nozzles had been
previously demonstrated under the EPRI MRP/Inspection Demonstration Program as
capable of detecting PWSCC type manufactured cracks as well as cracks from actual
samples from another site. Based on the demonstration, observation of in-process
examinations, and review of NDE data, the inspectors determined that the licensee was
capable of identifying PWSCC and/or corrosion as required by the NRC Order.
5) Evaluation of the RPVH condition (e.g. debris, insulation, dirt, boron from other
sources, physical layout, viewing obstructions).
No bare metal visual examinations required by the NRC Order were scheduled for RFO-
13. A 100% bare metal visual examination was performed during May 2003 refueling
outage and the NRC inspection activities for this visual examination were documented in
NRC Integrated Inspection Report 2003-003.
6) Evaluation of the licensees ability to identify and characterize small boron
deposits, as described in NRC Bulletin 2001-01.
No bare metal visual examinations required by the NRC Order were scheduled for RFO-
13. A 100% bare metal visual examination was performed during May 2003 refueling
outage and the NRC inspection activities for this visual examination were documented in
NRC Integrated Inspection Report 2003-003.
7) Evaluation of the extent of material deficiencies (i.e., cracks, corrosion, etc.) that
required repair.
No examples of RPVH leakage, material deficiencies, or flaws requiring repair were
identified during the NDEs. As indicated above, UT indications were identified at several
J-groove welds and they were dispositioned as metallurgical/geometric indications (not
service related).
8) Evaluation of any significant impediments to effectively perform each
examination method (e.g., centering rings, insulation, thermal sleeves, nozzle
distortion, etc.)
The RPVH nozzle examination volume extended from a minimum of 2-in above the
highest point of the J-groove weld to the maximum coverage possible below the lowest
point of the J-groove weld, which resulted to be more than 1-inch for all nozzles.
Westinghouse Letter LTR-PAFM-06-27, Shearon Harris Upper Head Penetration Hoop
Stress Distribution Below the Weld, dated April 20, documents that the stress levels in
all the RPVH nozzle surfaces located at more than 1-inch below the J-groove weld are
less than 20 ksi tension.
Enclosure
29
The inspectors concluded that the examination coverage requirement of the NRC Order
was met for all RPVH penetration nozzles. No significant items that could impede the
examination processes were noted during observation of the NDEs.
9) Evaluation of the basis for the temperatures used in the susceptibility ranking
calculation.
During the inspection activities documented in NRC Integrated Inspection Report 50-
400/2003-003, the inspectors reviewed the susceptibility ranking calculation and the
basis for the RPVH temperatures used in the calculation, as documented in calculation
HNP-M/MECH-1091, Effective Degradation Years for the Reactor Vessel Head. The
basis for the RPVH temperature used in the calculation was supported by
correspondence from Westinghouse to Progress Energy (Letter PGN-03-40, dated May
28, 2003).
10) Verification that the methods used for disposition of NDE identified flaws were
consistent with NRC flaw evaluation guidance.
No indications considered to be flaws were found during the RPVH examination. As
indicated above, UT indications were identified at several J-groove welds and they were
dispositioned as metallurgical/geometric indications (not service related).
11) Evaluation of the existing procedures to identify potential boric acid leaks from
pressure-retaining components above the RPVH and the licensees followup
actions for indications of boric acid leaks.
The inspectors reviewed Procedure OPT-1519, which was implemented to conduct
inspection activities required by the NRC Order to identify potential boric acid leaks from
pressure retaining components above the RPVH. The licensee also implemented
Procedure EST-227 in conjunction with Procedure OPT-1519. Procedure EST-227
requires visual examinations in areas above the RPVH to meet the Class 1 system
pressure test requirements of ASME Code,Section XI, Rules for Inservice Inspection of
Nuclear Power Plant Components. The licensee generated Condition Report 191710
on April19, 2006 to implement enhancements for Procedure EST-227, in order to clearly
specify that the visual examination requirements established in this procedure are also
intended to meet the NRC Order. The inspectors found that the implementation of the
procedures mentioned above met the requirements of the NRC Order.
The inspectors reviewed the visual examination results for RFO-13 and held discussions
with licensee personnel to confirm followup actions taken for any evidence of boric acid
leaks above the RPVH. The inspectors found that no indications of boric acid leaks
from pressure retaining components above the RPVH were identified.
4OA6 Meetings
On July 20, the resident inspectors presented the inspection results to Mr. Gannon and
other members of his staff. The inspectors confirmed that proprietary information was
not provided or examined during the inspection.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
D. Alexander, Superintendent, Environmental and Chemistry
A. Barginere, Superintendent, Security
D. Corlett, Supervisor - Licensing/Regulatory Programs
R. Downey, MR Coordinator
R. Duncan, Director - Site Operations
P. Fulford, Manger, Nuclear Assessment
C. Gannon, Vice President Harris Plant
B. Gause, Health Physics Supervisor, Radiation Protection
W. Gurganious, Training Manager
K. Henderson, Maintenance Manager
J. Jankens, Lead Specialist, Radwaste
C. Kamiliaris, Manager - Support Services
S. Larson, ISI Coordinator
E. McCartney, Plant General Manager
T. Natale, Manager - Outage and Scheduling
S. OConnor, Manager - Engineering
T. Pilo, Supervisor - Emergency Preparedness
K. Rogers, Lead Specialist, ALARA
G. Simmons, Superintendent - Radiation Control
E. Wills, Operations Manager
Contractor Personnel
C. Holmes, Manager of RPVH examination team, Wesdyne
F. Bonitz, Level III Examiner, Wesdyne
W. Holasak, Level III Examiner, Wesdyne
NRC personnel
P. Fredrickson, Chief, Reactor Projects Branch 4
Attachment
A-2
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000400/2006003-02 AV Failure to Maintain Adequate Procedures Such
That a Required Torque Was Not Provided for a
Threaded Fastener on an ESCW System Chiller
(Section 1R15.b.2)
Opened and Closed
05000400/2006003-01 NCV Failure to Follow Procedure During Service Water
Control Valve Preventive Maintenance (Section
1R15.b.1)
Closed
05000400, 2515/150 TI Reactor Pressure Vessel Head and Head
Penetration Nozzles (NRC Order EA-03-009)
(Section 4OA5)
Attachment
A-3
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
AP-300, Severe Weather Response
AP-301, Adverse Weather
FSAR Section 9.2.1, Service Water System
FSAR Section 8.3.1 AC Power Systems
FSAR Section 6.3, Emergency Core Cooling Systems
Section 1R04: Equipment Alignment
Partial System Walkdown
Residual heat removal system
Procedure OP-111, Residual Heat Removal System
Drawing 2165-S-1324, Simplified Flow Diagram Residual Heat Removal System
Emergency service water system
Procedure OP-139, Service Water System
Drawing 2165-S-0547 and 2165-S-0548, Simplified Flow Diagram Circulating and Service
Water Systems, sheets 1 and 2
Condensate storage tank
Drawing 2165-S-0545, Simplified Flow Diagram Condensate and Air Evacuation Systems Unit
1"
Section 1R05: Fire Protection
FPP-012-03-FHB, Fuel Handling Building Fire Pre-Plan
FPP-012-01-CNMT, Containment Building Fire Pre-Plan
Harris Nuclear Plant Fire Drill Planning Guide
Section 1R06: Flood Protection Measures
FSAR Sections
2.4.10, Flooding Protection Requirements
3.6A.6, Flooding Analysis
Calculations
Appendix I to the HNP Probabilistic Safety Assessment, Internal Flooding Analysis
Attachment
A-4
Procedures
AOP-022, Loss of Service Water
Section 1R07: Heat Sink Performance
Procedures
PLP-620, Service Water Program (Generic Letter 89-13)
EPT-163, Generic Letter 89-13 Inspections (Raw Water Systems and Local Area Air Handler
Inspection and Documentation)
Section 1R08: Inservice Inspection Activities
Procedures
SG-SGDA-04-38, Steam Generator Condition Monitoring Assessment of May 2004 Inspection
Results and Operational Assessment for Operating Cycle 12 & 13, August 2004
EGR-NGGC-0208, Steam Generator Integrity Program, Rev. 0
STD-100-228, Steam Generator Tube Stabilization Device Installation, Rev. 04
SH-SG-002, Rolled Mechanical Tube Plugging and Stabilizer Installation Using the Advanced
Roll Tool (ART) for Steam Generators with 0.6875" OD x 0.040" Wall Tubes, Rev. 00
EST-216, Steam Generator Degradation Assessment for RFO 13, Rev. 15
MNT-NGGC-0007, Foreign Material Exclusion Program, Rev. 6
MRS-SSP-1927-CQL, Westinghouse Model Delta 75 Steam Generator, Secondary Side
Tubesheet Inspection, Harris
MRS-TRC-1726, Eddy Current Steam Generator Examination Site Qualified Techniques
Validation Report, Rev. 0
EGR-NGGC-0207, Boric Acid Corrosion Control, Rev. 1
OPT-1519, Containment Visual Inspection for Boron and Evaluation of Containment Sump In-
Leakage Every Refueling Outage Shutdown, Mode 3, Rev. 8
PDI-ISI-254, Rev.7 Remote Inservice Inspection of Reactor Vessel Welds
PDI-ISI-254-NZ,Rev.0 Remote Inservice Inspection of Reactor Vessel Nozzle to Shell Welds
PDI-ISI-254-SE, Rev.2 Remote Inservice Examination of Reactor Vessel Nozzle to safe End ,
Nozzle to Pipe, and Safe End to Pipe Welds
WCAL-002, Rev.7 Pulser/Receiver Linearity Procedure
WDI-STD-088, Rev. 3 Underwater Remote Visual Examination of Reactor vessel Internals
EGR-NGGC-0207, Rev. 1 Boric Acid Control
OPT-1519, Rev. 8, Containment Visual Inspection for Boron and Evaluation of Containment
Sump In leakage every Refueling Outage Shutdown Mode 3
WDP9.2, Rev. 7 Qualification and Certification of personnel in Non-destructive Examination
SSI-A-005, Rev.22 Qualification and Certification of Nondestructive Examination Personnel
SSI-A-013, Rev3 Qualification and certification of Ultrasonic Examination personnel For Section
XI PSI/ISI Inspections
Self-Assessments
Harris SG Program Self Assessment, Number: 113074
Boric Acid Corrosion Control Program, Number 113075
Attachment
A-5
Action Request (AR)
00125127 Increased Radiation Monitor Readings o REM-3534 (due to SG Loose Parts
April 2004)
00191921 Loose Parts Retrieval in SG-A (Secondary Side)
00191964 Loose Parts Retrieval in SG-B (Secondary Side)
00191965 Loose Parts Retrieval in SG-C (Secondary Side)
00192210 Brown Boric Acid Leakage on FE-985 (Flow element on SI line to RCS Loop 1
Hot Leg)
00190581 Brown Boric Acid Leakage on FE-973 (Flow element on SI line to RCS Loop 3
Cold Leg)
00191004 Flange Downstream of 1SI-353 has Brown Boric Acid Buildup
Section 1R12: Maintenance Effectiveness
NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear
Power Plants
ADM-NGGC-0101, Maintenance Rule Program
HPP-780, Radiation Monitoring System Operating Manual
SD-118, Radiation Monitoring
OP-118, Radiation Monitoring System
OWP-RM-15, Radiation, Effluent, and Explosive Gas Monitoring
ISI-801, Inservice Inspection Program (ISI), Inservice Testing of Valves
EST-212, Type C Local Leak Rate Tests
Assessments
Maintenance Rule Cycle 12, Periodic a(3) Assessment (Assessment Period: June 30, 03 to
November 17, 04 - AR 147581147581
Self Assessment Report 111536, Work Management Process Assessment
Self Assessment Report 146488, HNP Implementation of Corrective Action Program and
Operating Experience Program, September 12 to 16, 2005
Nuclear Assessment Section Report, H-ES-05-01, Harris Engineering Functional Area
Assessment, May 13, 2005
Maintenance Rule Expert Panel Meeting Minutes
Meeting No. 03-03, 3/26/03
Meeting No. 03-05, 6/17/03
Meeting No. 04-03, 2/26/04
Meeting No. 04-15, 12/9/04
Meeting No. 05-03, 2/10/05
Meeting No. 05-07, 4/28/05
Meeting No. 05-12, 9/1/05
Meeting No. 06-04, 5/31/06
Attachment
A-6
Maintenance Rule - System Scoping Documents
System 4085: Essential Service Chilled Water
System 5165: 6.9 KV AC Distribution
System 9001: Containment Isolation Valves
System 3050: Main Feedwater System
System 5230: 250 VDC Electrical Distribution
System 4060: Normal Service Water
System 5095: Emergency Diesel Generators
Corrective Action Documents
Significant Adverse Condition Investigation Reports for
Action Request (AR) 38438
AR 159131159131AR 172596
AR 190137190137AR 105539
AR 119086119086AR 79228
AR 132488132488AR 140449
AR 160899160899AR 099026-01 [evaluation only]
AR 198973198973AR 198948
AR 198754198754Other Documents
Corrective Maintenance Work Orders completed after June 1, 2005 for:
System 4085: Essential Service Chilled Water
System 5165: 6.9 KV AC Distribution
System 5095: Emergency Diesel Generators
System 5100: Diesel Fuel Oil
System 5110: Diesel Jacket Water System
System 5095: Emergency Diesel Generators
System 5112: Diesel Starting Air
System Health Reports for:
System 5165: 6.9 KV AC Distribution
System 5230: 250 VDC Electrical Distribution
System 4060: Normal Service Water
System 9001: Containment Isolation Valves
System 4085: Essential Service Chilled Water
Attachment
A-7
Work Orders 00721491-01 6.9 KV Bus UV Circuit Time Delay Pickup Relay (TYPICAL)
00649357-01 UV Relay for 1A-SA
Section 1R13: Maintenance Risk Assessments and Emergent Work Evaluation
OMP-003, Outage Shutdown Risk Management.
WCM-001, On-line Maintenance.
Section 1R15: Operability Evaluations
OPS-NGGC-1305, Operability Determinations
Section 1R20: Refueling and Outage Activities
FHP-020, Refueling Operations
FHP-014, Fuel and Insert Shuffle Sequence
EST-923, Initial Criticality and Low Power Physics Testing
GP-004, Reactor Startup
Section 1R23: Temporary Plant Modifications
DBD-201, Emergency Diesel Generator System.
Drawing 7-G-0509, Protective Mats For Class 1 Yard Duct Runs Plan & Misc Details.
FSAR section 3.5, Missile Protection.
Section 2OS1: Access Controls to Radiologically Significant Areas
Procedures, Manuals, and Guidance Documents
AP-504, Administrative Controls for Locked and Very High Radiation Areas, Revisions (Rev.) 22
AP-535, Performing Work in Radiological Control Areas, Rev. 19
CAP-NGGC-0200, Corrective Action Program, Rev. 16
HPP-600, Radiation Work Permits, Rev. 20
HPP-625, Performance of Radiological Surveys, Rev. 21
HPP-800, Handling Radioactive Material, Rev. 46
HPS-NGGC-0003, Radiological Posting, Labeling and Surveys, Rev. 8
HPS-NGGC-0014, Radiation Work Permits, Rev. 3
HPS-NGGC-0016, Access Control, Rev. 2
PLP-511, Radiation Control and Protection Program, Rev. 18
Radiation Work Permit (RWP) Documents
RWP 3408, Health Physics Routines
RWP 3438, Reactor Head/Core Activities
RWP 3441, Seal Table Activities
RWP 3455, Nozzle Dam/Nozzle Cover
RWP 3596, S/G Sludge Lance Activities
Attachment
A-8
RWP 3601, Reactor Head Volumetric Exam
RWP 3602, Reactor Vessel 10-Year ISI Activities
RWP 3696, Radiography
Licensee Records and Data
AP-555, Rev. 3, 04-23-06, Attachment 4 - Radiography Checklist, RAB 261 - Lower Filter
House Valve Gallery
AP-555, Rev. 3, 04-24-06, Attachment 4 - Radiography Checklist, RCB 236 - BD 65
HRA Briefing, 04/23/06, Decon RHR Motor A
Radiological Survey Record No. 0413-005, Repair UT Tool and Reinstall, RCB 286', Elevated
Reactor Head Stand Area, 04-13-06
Radiological Survey Record No. 0413-010, Post Shielding for work areas, RCB 286', Elevated
Reactor Head Stand Area, 04-13-06
Radiological Survey Record No. 0424-004, Pre-Radiography Survey, RCB 236, 04-24-06
Nuclear Condition Report (NCR) and Quality Assurance (QA) Documents
AR 141414141414 Personnel error resulting in breach of LHRA boundary, 10/23/2004
AR 156706156706 Entry to HRA on wrong RWP, 04/18/2005
AR 160694160694 Weakness 1 from Self-Assessment on LHRA controls, 06/07/2005
AR 160695160695 Weakness 2 from Self-Assessment on LHRA controls, 06/07/2005
AR 160697160697 Weakness 3 from Self-Assessment on LHRA controls, 06/07/2005
AR 162828162828 ED dose rate alarm, 07/05/2005
AR 181822181822 Unposted HRA, 01/21/2006
AR 190133190133 Inadequate HRA boundary, 04/05/2006
AR 190896190896 Individual failed to verify correct RWP task, 04/12/2006
AR 190958190958 Valid ED dose rate alarm, 04/12/2006
AR 190981190981 Reactor containment building 221' LHRA above in-core sump, 04/12/2006
AR 191123191123 In-core thimble retraction in excess of 21' 8", 04/14/2006
NAS Report File No. H-RP-05-01, Harris Radiation Protection Assessment, 11/02/2005
Self-Assessment Report No. 152347, Locked High Radiation Area Controls, conducted
05/09-18/2005
Section 2OS2: As Low As Reasonably Achievable
Procedures, Manuals, and Guidance Documents
ADM-NGGC-0105, ALARA Planning, Rev. 7
AP-110, Pre-Job/Post-Job Briefings, Rev. 18
AP-530, ALARA, Rev. 8
AP-535, Performing Work in Radiation Control Areas, Rev. 18
AP-555, Radiography, Rev. 3
CAP-NGGC-0200, Corrective Action Program, Rev. 16
DOS-NGGC-0002, Dosimetry Issuance, Rev. 22
DOS-NGGC-0004, Administrative Dose Limits, Rev. 8
HPP-600, Radiation Work Permits, Rev. 20
HPS-NGGC-0003, Radiological Posting, Labeling And Surveys, Rev. 8
Attachment
A-9
HPS-NGGC-0014, Radiation Work Permits, Rev. 3
MNT-NGGC-003, Radiation Shielding Use, Rev. 8
PLP-511, Radiation Control and Protection Program, Rev. 18
Licensee Records and Data
2006 ALARA Budget Considerations
2006 Normal Operations Monthly Dose Goals
2006-2008 HNP Business Plan Initiative for ALARA Program - Radiation Exposure Reduction
Active ALARA Reviews, 04/11/06
ALARA Committee Meeting Minutes, March 2005 - February 2006
Comparison of Dose (REM) for Repetitive Outage Tasks from RFO-1 to RFO-12
Dose Budgets for 2005
Dose Projection Basis for 2005
Evaluated Risk Assessment, Manual Eddy Current, RWP # 3441, 04/10/06
Harris Source Term Management Document
HNP Five Year Dose Reduction Plan 2004 - 2008
HNP Historical Dose 1992 to 2005
HNP Refueling Outage 13, Radiological Status Report, 04/26/06 and 04/27/06
RFO-13 Projections vs. Actuals (04/25/06)
RFO-13 Refueling and Scaffold ALARA Plans
Steam Generator Channel Head Radiation Levels for RFO-1 to RFO-11
TEDE ALARA Evaluation, Seal Table Maintenance Activities, RWP # 3441, 04/07/06
Temporary Shielding Requests (TSR)
TSR 06-024, Revs. 1 and 2, Generic shielding for Rx Head while it is on the head stand
TSR 06-035, Shield under head to remove LHRA
ALARA Work Packages (AWP)
AWP 06-063, Rev. 1, Seal Table Maintenance Activities
AWP 06-068, Shielding Activities
AWP 06-074, Sludge Lance, Foreign Object Search and Retrieval (FOSAR), and Secondary
Side Steam Generator Inspections
AWP 06-076, Rev. 2, Install Remove Nozzle Dams/Nozzle Covers
AWP 06-078, Rev. 1, Replace A RHR Pump and Motor
Radiation Protection Pre-Job Brief Packages
AP-110, Rev. 18, 04-09-06 (2200), Seal Table Room Activities RO-13
AP-110, Rev. 18, 04-10-06 (1330), Seal Table Eddy Current
AP-110, Rev. 18, 04-22-06 (0830), Remove/Install Steam Generator Nozzle Dams/Covers,
RWP 3455, AWP 06-076, Rev. 2
AP-110, Rev. 18, 04-23-06 (1930), Radiography, RWP 3696
AP-110, Rev. 18, 04-23-06 (2030), Decon A RHR Flange and Bolts, RWP 3457, AWP 06-078,
Rev. 1
Attachment
A-10
AP-110, Rev. 18, 04-24-06 (0500), Radiography on BD-65 RCB, RWP 3696
AR 190504190504 RCS sample activity causes higher than expected dose, 04/08/06
AR 190754190754 RCS shutdown crud burst and dose rates, 04/10/06
AR 190806190806 Detail not provided to ALARA for planning on incore ECT, 04/11/06
AR 190978190978 Reactor head shielding package not per design, 04/12/06
AR 191168191168 Contributors to elevated dose rates, 04/14/06
AR 191290191290 RFO-13 dose status, 04/17/06
AR 191927191927 ALARA plan estimated exceeded, 04/21/06
Section 2PS2: Radioactive Material Processing and Transportation
Procedures, Manuals, and Guidance Documents
CAP-NGGC-0200, Corrective Action Program, Rev. 7
HPP-880, Spent Nuclear Fuel Shipping and Receipt, Rev. 27
HPS-NGGC-0001, Radioactive Material Receipt and Shipping Procedure, Rev. 22
HPS-NGGC-0002, Vendor Cask Utilization Procedure, Rev. 13
PLP-300, Process Control Program, Rev. 9
Records and Data
2005 Annual Radioactive Effluent Release Report
Radman Database Report, Change 47 (10 CFR 61.55 analysis data)
Radioactive Materials Receipt Log 2005 and 2006 (Year-To-Date)
Radioactive Materials Shipment Logbook 2005 and 2006 (Year-To-Date)
Radwaste Shipment: 05-004, 21 filter drums in 21-300 cask to Duratek
Radwaste Shipment: 05-013, Boron-10 samples (4 gallon overpack)
Radwaste Shipment: 06-002, Fuel handling tool to Westinghouse
Radwaste Shipment: 06-010, 20 ft sealand container of used Orex protective clothing to ETI
Radwaste Shipment: 06-025, Spent resin in 8-120 cask to Studsvik
Radwaste Shipment: 06-033, Pressurizer relief valve and snubbers
AR 00141185, Inadequate Radioactive Material Labeling
AR 00159463, Ambiguously Labeled Drum Containing Radwaste Material
AR 00169189, Rad waste Shipping Documentation Inattention to Detail
AR 00169354, Transposition Error on Shipment Documentation
Nuclear Assessment H-RP-05-01, Harris Radiation Protection Assessment
Section 4OA1: Performance Indicator Verification
Records
2005 Annual Radioactive Effluent Release Report
Attachment
A-11
AR Searches for High Radiation, HRA, LHRA, and Postings.
CAP-NGGC-0200, Corrective Action Program, Rev. 16
REG-NGGC-0009, NRC Performance Indicators And Monthly Operating Report Data, Rev. 5
Searches for ED alarms >100 mr above setpoint, dose rate alarms >1000mr/hr.
Shearon Harris Nuclear Power Plant, Off-Site Dose Calculation Manual (ODCM), Rev. 16
Section 4OA2: Identification and Resolution of Problems
CAP-NGGC-0200, Corrective Action Program.
HNP-Site Trend Report - Fourth Quarter, 2005 and First Quarter, 2006.
Section 4OA5: Other Activities
WDI-STD-101, RVHI Vent, Eddy Current, Plus-Point Coil, Revision 5
WDI-STD-114, RVHI Vent Tube, Eddy Current, Plus-Point Coil, Bobbin Coil, Revision 4
WDI-SSP-1013, Procedure for Detection and Sizing of Cracks in PWR Reactor Vessel
Closure Head Penetrations using UT Techniques, Revision 2
WDI-SSP-1014, Procedure for the Eddy Current Inspection of Reactor Vessel Head
Penetration Nozzles with Thermal Sleeves using Blade Probes, Revision 2
WDI-SSP-1016, Guidelines for Analyzing Data from PWR Reactor Vessel Head Penetrations
using MASERA and MASERA_TOFD, Revision 1
WDI-SSP-1017, Eddy Current Analysis Guidelines for RPV Penetrations, Revision 1
WDI-SSP-1025, Procedure for the Eddy Current Inspection of Open Reactor Vessel Head
Penetration Nozzles, Revision 3
EST-227, ASME Section XI Class 1 System Pressure Test, Revision 7
OPT-1519, Containment Visual Inspection for Boron and Evaluation of Containment Sump In-
leakage Every Refueling Outage Shutdown Mode 3, Revision 8
WDP-9.2, Qualification and Certification of Personnel in Nondestructive Testing, Revision 7
ANATEC-08, Certification of NDT Personnel, Revision 17
ML-QAP-9.1, Certification of NDE Personnel (ET), Revision 7
SSI-A-005, Qualification and Certification of Nondestructive Examination Personnel, Revision
22
HC-00, Qualification and Certification of Non-destructive Testing Personnel, Revision 14
Engineering Documents
Summary of Demonstration Results: Material Reliability Program (MRP), Demonstration of
Tectaton Equipment and Procedures for the Inspection of Alloy 600 Control Rod Drive
Mechanism (CRDM) Head Penetrations, Dated March 15, 2005
IR-2006-108, Summary of Demonstration Results: Material Reliability Program (MRP),
Demonstration of Tectaton Equipment and Procedures for the Inspection of Alloy 600/690
Control Rod Drive Mechanism (CRDM) Head Penetrations, Dated March 2006
Westinghouse Letter LTR-PAFM-06-27, Shearon Harris Upper Head Penetration Hoop Stress
Distribution Below the Weld, dated April 20, 2006
Attachment
A-12
Corrective Action Documents
Condition Report 00191710*
Condition Report 00191703*
Condition Report 00191704*
- Condition Reports generated as a result of this inspection
Records
Personnel Certification Records for all Wesdyne Examiners
Framatome Equipment Certification Records for the following NDE Equipment
UT/ET Blade Probes: 2889, 2963, 2919, 3002, and 3003
UT Probe - Circumferential: 3493
UT Rotating Probes: 11239-11240, 11241-11242, and 11894
Calibration Reports: UCR-1/5-2, UCR-2/5-2, UCR-3/5-2, UCR-4/5-2, UCR-5/5-2
Attachment