ML062080742
| ML062080742 | |
| Person / Time | |
|---|---|
| Site: | Farley, 07200042  |
| Issue date: | 07/27/2006 |
| From: | Scott Shaeffer Reactor Projects Branch 2 |
| To: | Sumner H Southern Nuclear Operating Co |
| References | |
| IR-06-003 | |
| Download: ML062080742 (38) | |
See also: IR 05000348/2006003
Text
July 27, 2006
Southern Nuclear Operating Company, Inc.
ATTN: Mr. H. Lewis Sumner
Vice President - Farley Project
P. O. Box 1295
Birmingham, AL 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000348/2006003, 05000364/2006003, AND 07200042/2006001
Dear Mr. Sumner:
On June 30, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Joseph M. Farley Nuclear Plant, Units 1 and 2. The enclosed integrated inspection report
documents the inspection findings, which were discussed on July 7, 2006, with Mr. Randy
Johnson 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 two inspector identified findings and one self-revealing finding, all of very
low safety significance (Green). These findings were determined to involve violations of NRC
requirements. However, because of the very low safety significance and because they were
entered into your corrective action program, the NRC is treating these findings as non-cited
violations (NCVs), in accordance with Section VI.A of the NRCs Enforcement Policy. If you
contest these NCVs, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the United States 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 Farley
Nuclear Plant.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosures, and your response (if any) will be available electronically for public inspection in the
2
NRC Public Document Room or from the Publicly Available Records (PARS) component the
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/
Scott M. Shaeffer, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Docket Nos. 50-348, 50-364, and 72-42
Enclosure: Inspection Report 05000348/2006003, 05000364/2006003, and
07200042/2006001
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
__ML062080742
OFFICE
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SIGNATURE
PKV /RA/
RCC /RA/
ECM /RA/
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NAME
KVanDoorn
RChou
EMichel
LLake
DATE
July 28, 2006
July 25, 2006
July 25, 2006
July 26, 2006
E-MAIL COPY?
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NO YES
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OFFICE
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SIGNATURE
GBK /RA/
AND /RA/
NJG /RA/
MAS /RA/
NAME
GKuzo
ANielsen
NGriffis
MScott
DATE
July 25, 2006
July 25, 2006
July 25, 2006
July 28, 2006
E-MAIL COPY?
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cc w/encl:
B. D. McKinney, Licensing
Services Manager, B-031
Southern Nuclear Operating
Company, Inc.
Electronic Mail Distribution
J. R. Johnson
General Manager, Farley Plant
Southern Nuclear Operating
Company, Inc.
Electronic Mail Distribution
J. T. Gasser
Executive Vice President
Southern Nuclear Operating
Company, Inc.
Electronic Mail Distribution
Bentina C. Terry
Southern Nuclear Operating Company, Inc.
Bin B-022
P. O. Box 1295
Birmingham, AL 35201-1295
State Health Officer
Alabama Department of Public Health
RSA Tower - Administration
201 Monroe St., Suite 700
P. O. Box 303017
Montgomery, AL 36130-3017
M. Stanford Blanton
Balch and Bingham Law Firm
P. O. Box 306
1710 Sixth Avenue North
Birmingham, AL 35201
William D. Oldfield
Quality Assurance Supervisor
Southern Nuclear Operating Company
Electronic Mail Distribution
Distribution w/encl: (See page 4)
1
Letter to H. Lewis Sumner from Scott Shaeffer dated July 27, 2006
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000348/2006003, 05000364/2006003, AND 07200042/2006001
Distribution w/encl:
R. Martin, NRR
C. Evans
L. Slack, RII EICS
OE Mail
RIDSNRRDIRS
PUBLIC
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.:
50-348, 50-364, 72-42
License Nos.:
Report Nos.:
05000348/2006003, 05000364/2006003, and 07200042/2006001
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Joseph M. Farley Nuclear Plant
Location:
7388 N. State Highway 95
Columbia, AL 36319
Dates:
April 1- June 30, 2006
Inspectors:
C. Patterson, Senior (Sr.) Resident Inspector
J. Baptist, Resident Inspector
K. VanDoorn, Sr. Reactor Inspector (Section 1R08)
R. Chou, Reactor Inspector (Section 1R08)
E. Michel, Reactor Inspector (Sections 1R08, 4OA2)
M. Scott, Reactor Inspector (Section 4OA2)
L. Lake, Reactor Inspector (Section 4OA2)
G. Kuzo, Sr. Health Physicist (Sections 2OS1, 4OA1)
A. Nielsen, Health Physicist (Sections 2OS2, 4OA1)
N. Griffis, Health Physicist (Sections 2PS2, 4OA5)
Approved by:
Scott M. Shaeffer, Chief
Reactor Projects Branch 2
Division of Reactor Projects
SUMMARY OF FINDINGS
IR 05000348/2006003, 05000364/2006003, and 07200042/2006001; 04/01/2006-06/30/2006;
Joseph M. Farley Nuclear Plant, Units 1 & 2; Problem Identification & Resolution, Refueling and
Other Outage Activities.
The report covered a three-month period of inspection by the resident inspectors, five regional-
based Reactor Inspectors, and three Health Physicists. Three Green non-cited violations were
identified. The significance of most findings are indicated by their color (Green, White, Yellow,
or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process
(SDP). Findings for which the SDP does not apply may be Green or be assigned a severity
level after 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
C Green. A Green non-cited violation (NCV) of 10 CFR 50.55a (a) (2) was identified by
the NRC for the licensee failing to comply with the ASME Boiler and Pressure Vessel
Code,Section XI, for Class 2 Components. The licensee failed to meet the ASME
Code requirements for a Unit 2 Charging Safety Injection pump casing replacement,
when they did not obtain a completed NIS-2 form signed by the Authorized Nuclear
Inservice Inspector (ANII).
The finding is more than minor because it affected the mitigating systems cornerstone
objective to assure the reliability of systems that respond to events to prevent
undesirable consequences and was associated with the design control attribute in that
qualification remains questionable. The finding was evaluated as very low risk
significance (Green) because it was a qualification deficiency confirmed not to result in
a loss of operability. This finding has been entered into the licensees corrective
Action Program. (Section 4OA2.2)
C Green. A Green non-cited violation (NCV) of Technical Specification 5.4, Procedures,
was identified by the NRC for failure to follow procedural guidance associated with
removal of debris in containment. The licensee performed a containment inspection;
however, failed to follow adequate procedural guidance to ensure the containment
environment was acceptable for power operations.
This finding is more than minor because it could be reasonably viewed as a precursor
to a significant event involving debris accumulation on the containment sump screens
and a subsequent impairment to suction flow for Emergency Core Cooling System
(ECCS) pumps. Although it impacted the Mitigating System Cornerstone, it did not
result in a loss of function per Inspection Manual Chapter (IMC) Part 9900, Technical
Guidance, Operability Determination Process for Operability and Functional
Assessment, did not represent an actual loss of safety function, and was not
potentially risk significant due to possible external events. This finding was entered
into the licensees corrective action program. (Section 1R20)
2
Cornerstone: Barrier Integrity
C Green. A Green self-revealing non-cited violation (NCV) of 10 CFR 50, Appendix B,
Criterion VII, Control of Purchased Material, Equipment, and Services, was identified
for failure to control adequately contractors during the Unit 1 refueling outage that
resulted in damage to the fuel transfer system. This was a self-revealing violation
when a pillar block weld broke resulting in damage to the transfer cart, rails, basket,
and dummy fuel assembly. The licensee entered the deficiency into their corrective
action program for resolution.
This finding is more than minor because it could be reasonably viewed as a precursor
to a significant event involving damage to a fuel assembly. Although the damage
occurred to a dummy fuel assembly, the stresses applied to the fuel transfer system
occurred during core offload and it is was fortuitous that the failure happened when a
dummy assembly was in the fuel transfer basket. This finding is of very low safety
significance because no damage to a fuel assembly actually occurred. (Section 1R20)
B.
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near Rated Thermal Power (RTP) until April 8, 2006, when the unit was
shutdown to begin a refueling outage. The outage ended on May 24, 2006. The unit operated
at RTP until June 30, 2006 when the unit was shutdown due to an inoperable main steam
isolation valve.
Unit 2 operated at or near 100% RTP for the duration of the reporting period.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04
Equipment Alignment
a.
Inspection Scope
Partial System Walkdowns. The inspectors performed partial walkdowns of the following
three systems to verify they were properly aligned when redundant systems or trains
were out of service. The walkdowns were performed using the criteria in licensee
procedures FNP-0-AP-16, Conduct of Operations - Operations Group, and
FNP-0-SOP-0, General Instructions to Operations Personnel. The walkdowns included
reviewing the Updated Final Safety Analysis Report (UFSAR), plant procedures and
drawings, checks of control room and plant valves, switches, components, electrical
power line-ups, support equipment, and instrumentation.
C 1-2A, 2B, 1C, 2C Emergency Diesel Generators (EDGs) while 1B EDG out of service.
C 1B Motor Driven Auxiliary Feedwater (MDAFW) Pump during 1A MDAFW freeze seal
maintenance activities.
C 2B and 2C Service Water (SW) Pumps while 2A SW pump was out of service.
Complete Walk-down. The inspectors conducted a complete walkdown of the
accessible portions of the Unit 1 Residual Heat Removal system (RHR). The inspectors
used licensee procedure FNP-1-SOP-7.0A, Residual Heat Removal System and
drawings D-175038 and D-175041, to verify adequate system alignment of on-service
equipment. The inspectors also interviewed personnel and reviewed control room logs,
Maintenance Rule (MR) monthly reports, condition reports (CRs), quarterly system
health reports, outstanding work orders, and industry operating experience to verify that
alignment and equipment discrepancies were being identified and appropriately
resolved. Documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified. The inspectors did identify one valve that
was improperly sealed. The licensee immediately corrected the valve status and
captured the issue in CR 2006105758. The position of the valve was correct and the
function of the system was not compromised due to this discrepancy.
1R05
Fire Protection
6
a.
Inspection Scope
Fire Area Tours. The inspectors conducted a walk-down of the seven fire areas listed
below to verify the licensees control of transient combustibles, the operational readiness
of the fire suppression system, and the material condition and status of fire dampers,
doors, and barriers. The requirements were described in licensee procedures
FNP-0-AP-36, Fire Surveillance and Inspection; FNP-0-AP-38, Use of Open Flame;
FNP-0-AP-39, Fire Patrols and Watches; and the associated Fire Zone Data sheets.
C Unit 1 and 2 EDG Building, Switchgear Room Train A, Zone 56A
C Unit 1 and 2 EDG Building, Diesel Generator 2C, Zone 57
C Unit 1 and 2 EDG Building, Diesel Generator 1B, Zone 58
C Unit 1 and 2 EDG Building, Diesel Generator 2B, Zone 59
C Unit 1 and 2 EDG Building, Diesel Generator 1C, Zone 60
C Unit 1 and 2 EDG Building, Diesel Generator 1-2A, Zone 61
C Unit 1 and 2 EDG Building, Switchgear Room Train B, Zone 56C
b.
Findings
No findings of significance were identified.
1R08
Inservice Inspection (ISI) Activities
a.
Inspection Scope
Piping Systems ISI. The inspectors reviewed the implementation of the licensees ISI
program for monitoring degradation of the reactor coolant system boundary and the risk
significant piping system boundaries for Unit 1. The inspectors selected a sample of
American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,
Section XI required examinations and a sample of risk-informed ISI Program
examinations. The inspectors conducted an on-site review of the following
nondestructive examination (NDE) activities to evaluate compliance with Technical
Specifications (TS), ASME Section XI and ASME Section V requirements - 1989 Edition,
and to verify that indications and defects were appropriately evaluated and dispositioned
in accordance with the requirements of ASME Section XI, IWB-3000 acceptance
standards.
Ultrasonic Testing (UT):
- 32"-ALA2-4300-14-RI, Main Steam to Penetration
- 555"-Elliptical Head to Upper Shell, Steam Generator C
The inspectors reviewed the following examination records in addition to the records for
the above observed examinations:
UT:
- 12"-ALA1-4301-6-RI, Pipe to Elbow
- 10"-ALA2-4516-28-RI, Elbow to Pipe
- 3"-ALA2-4540-32-RI, Valve to Pipe
7
- 2"-ALA1-4310-2-RI, Pipe to Elbow
- 2"-ALA1-4307-5-RI, Elbow to Pipe
The inspectors also observed two visual and two UT examinations of the containment
liner along excavated portions of the containment moisture barrier performed in
accordance with ASME Section XI, IWE, 1992 Edition. This examination was perform in
response to previously discovered indications of moisture barrier degradation observed
by an NRC inspector and recorded in the licensees Corrective Action Program.
Qualification and certification records for examiners, inspection equipment, and
consumables along with the applicable NDE procedures for the above ISI examination
activities were reviewed and compared to requirements stated in ASME Section V and
Section XI.
The inspectors performed a review of piping system related problems that were
identified by the licensee and entered into the corrective action program. The inspectors
reviewed these corrective action documents to confirm that the licensee had
appropriately described the scope of the problems and had implemented effective
corrective actions.
Boric Acid Corrosion Control (BACC) ISI. The inspectors reviewed the licensees Boric
Acid Corrosion Control Program (BACCP) to ensure compliance with commitments
made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel
Reactor Pressure Boundary, and Bulletin 2002-01, Reactor Pressure Vessel Head
Degradation and Reactor Coolant Pressure Boundary Integrity. The inspectors
conducted an on-site record review and an independent walk-down of the reactor
building, which was not normally accessible during at-power operations, to evaluate
compliance with licensee BACCP requirements and 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Action, requirements. In particular, the inspectors verified that
licensee visual examinations focused on locations where boric acid leaks can cause
degradation of safety significant components and that degraded or non-conforming
conditions were properly identified in the licensees corrective action system.
The inspectors reviewed the licensees program implementation procedures and a
sample of plant issue reports (corrective action documents) to ensure that leaks were
being identified and addressed at an appropriate threshold. A sample review of
corrosion assessments was also completed for boric acid deposits found on reactor
coolant system piping and other ASME Code Class components to verify that the
minimum design code required section thickness had been maintained for any affected
component(s).
Steam Generator (SG) Tube ISI. The inspectors reviewed activities, plans, pre-outage
degradation assessment and procedures for the inspection and evaluation of the steam
generator Inconel Alloy 690TT tubing for Unit 1 SGs A, B, and C, to determine if the
activities were being conducted in accordance with Technical Specifications and
applicable industry standards. Data gathering, analysis, and evaluation activities were
reviewed. The inspectors reviewed data results for tubes at SG B - R45C59; SG A -
R17C04, R15C10, and R26C10; and SG C - R11C82, R12C82, R10C93, R04C51,
8
R03C56, and R13C81 to verify the adequacy of the licensees primary, secondary, and
resolution analyses. The inspectors observed the licensee perform 100% of the
video/visual inspection for the secondary side on the top of tubesheet area of the steam
generators to determine if foreign materials or loose parts were present and the licensee
was conducting appropriate evaluations. The inspectors also reviewed data operators
and analysts certifications and qualifications, including medical exams.
b.
Findings
No findings of significance were identified.
1R11
Licensed Operator Requalification
a.
Inspection Scope
The inspectors observed portions of the licensed operator training and testing program
to verify implementation of procedures FNP-0-AP-45, Farley Nuclear Plant Training
Program, FNP-0-TCP-17.6, Simulator Training Evaluation Documentation, and FNP-0-
TCP-17.3, Licensed Operator Continuing Training Program. The inspectors observed
scenarios conducted in the licensees simulator for a steam generator tube leak, failed
instrumentation, and load rejection. The inspectors observed high risk operator actions,
overall performance, self-critiques, training feedback, and management oversight to
verify operator performance was evaluated against the performance standards of the
licensees scenario. In addition, the inspectors observed implementation of the
applicable emergency operating procedures listed in the attachment to verify that
licensee expectations in procedures FNP-0-AP-16 and FNP-0-TCP-17.6 were met.
Documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified.
1R12
Maintenance Effectiveness
a.
Inspection Scope
The inspectors reviewed the following two issues to verify implementation of licensee
procedures FNP-0-87, Maintenance Rule (MR) Scoping Manual; NMP-ES-021,
Structural Monitoring Program for the Maintenance Rule; and FNP-0-89, FNP
Maintenance Rule Site Implementation Manual; and compliance with 10CFR50.65. The
inspectors assessed the licensees evaluation of appropriate work practices, common
cause failures, functional failures, maintenance preventable functional failures, repetitive
failures, availability and reliability monitoring, trending and condition monitoring, and
system specialist involvement. The inspectors also interviewed maintenance personnel,
system specialists, the MR coordinator, and operations personnel to assess their
knowledge of the program.
C CR 2006104043, 1C Component Cooling Water (CCW) Pump Did Not Start During
Safety Injection (SI)/Loss of offsite power (LOSP) Test
9
C CR 2006104286, Maintenance preventable functional failure (MPFF) of 1C Service Air
Compressor
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Control
a.
Inspection Scope
The inspectors assessed the licensees planning and control for the following five
planned activities to verify the requirements in licensee procedures FNP-0-ACP-52.3,
Guidelines for Scheduling of On-Line Maintenance; NMP-GM-006, Work Management;
and FNP-0-AP-16, Conduct of Operations - Operations Group; and the MR risk
assessment guidance in 10CFR50.65a(4) were met.
C CR 2006103167, Outage Risk-Entered Orange Prematurely
C CR 2006102858, Instrument Air Compressor Problems
C CR 2006101909, Significant Axial Flux Oscillations During Derate
C CR 2006103262, 1B EDG Air Receiver Blowdown During Orange Shutdown Safety
Assessment
C CR 2006104091, Damage to Unit 1 Fuel Transfer System
b.
Findings
No findings of significance were identified.
1R15
Operability Evaluations
a.
Inspection Scope
The inspectors reviewed the following five operability evaluations to verify they met the
requirements of licensee procedures FNP-0-AP-16, Conduct of Operations and
FNP-0-ACP-9.2, Operability Determination for Technical Adequacy, Consideration of
Degraded Conditions, and Identification of Compensatory Measures. The inspectors
reviewed the evaluations against the design bases, as stated in the UFSAR and
Functional System Descriptions (FSDs) to verify system operability was not affected.
C CR 2006103043, Unit 2 Main Steam Isolation Valves (MSIVs)
C CR 2006104130, OD 6-04, 2B EDG minor leakage Service Water Piping
C CR 2006104389, OD 6-05, 1-2A EDG service water leakage at Expansion joint
C CR 2006104942, Turbine Driven Auxiliary Feedwater (TDAFW) Pump Check Valve
Flow Verification test failure
C CR 2006105386, OD 6-06, 2A MDAFW Pump miniflow line vent piping leak
b.
Findings
No findings of significance were identified.
10
1R17
Permanent Plant Modifications
a.
Inspection Scope
The inspectors reviewed the following plant modification to verify the implementation of
procedure FNP-0-AP-8, Design Modification Control. This included verification that the
design bases, licensing bases, and performance capability or risk significant systems,
structures, and components would not be degraded through the modifications and the
modifications would not place the plant in an unsafe condition. The inspectors also
discussed the modifications with engineering and operations personnel, and reviewed
the related procedures and drawings.
C DCP 03-1-9976, Unit 1 Containment Scaffolding Storage Modification
b.
Findings
No findings of significance were identified.
1R19
Post Maintenance Testing
a.
Inspection Scope
The inspectors reviewed the criteria contained in licensee procedures FNP-0-PMT-0.0,
Post Maintenance Test Program, to verify post-maintenance test procedures and test
activities for the following five systems/components were adequate to verify system
operability and functional capability.
C WO 2061229101, VT-2 check of SW Leak
C FNP-1-STP-45.7, MSIV and Bypass Valves Cold Shutdown Inservice Test
C FNP-0-ETP-3643, Verification of Rod Control System Operability
C FNP-2-STP-22.23, TDAFW Pump Trip and Throttle Valve Mechanism and Indicator
Operability Test
C FNP-0-STP-80.1, Diesel Generator 1-2A Operability Test .
b.
Findings
No findings of significance were identified
1R20
Refueling and Other Outage Activities
a.
Inspection Scope
Refueling Activities. The inspectors reviewed the following activities related to the Unit 1
refueling outage for conformance to licensee procedure FNP-0-UOP-4.0, General
Outage Operations Guideline, and FNP-1-UOP-4.1, Controlling Procedure for Refueling.
Surveillance tests were reviewed to verify results were within the TS required
specification. Shutdown risk, management oversight, procedural compliance, and
operator awareness were evaluated for each of the following activities. Documents
reviewed are listed in the attachment.
11
C Outage Risk Assessment
C Cooldown
C Core offload and reload
C Reactor coolant instrumentation
C Electrical system alignments and bus outages
C Reactor vessel disassembly and assembly activities
C Outage-related surveillance tests
C Containment Closure
C Low Power Physics Testing and Startup Activities
C Clearance Activities
C Decay Heat Removal and Spent Fuel Pool Cooling
b.
Findings
.1
Failure to Follow Established Procedures Regarding Debris in Containment
Introduction: A Green non-cited violation (NCV) for failure to follow procedural guidance
associated with removal of debris in containment was identified by the NRC.
Description: On May 20, 2006, the licensee implemented procedures FNP-1-STP-34.1,
Containment Inspection (Maintenance) and FNP-1- STP-34.0, Containment Inspection
(General). The Containment Inspection (Maintenance) procedure is maintained open
until all work is completed to track work items as they enter and exit containment. The
Containment Inspection (General) procedure was performed by the licensee as a
general walkdown to verify that containment was acceptable. The licensee completed
the Containment Inspection (General) procedure and documented it as satisfactorily
finished. The acceptance criteria is to ensure that there is No loose debris (rags, trash,
clothing, etc.) present in the containment which could be transported to the containment
sump and cause restricted pump suctions during LOCA conditions.
On May 21, 2006, after Unit 1 plant entry into Mode 4, the resident inspectors
performed a containment closeout tour to determine the status of housekeeping and
equipment storage in containment. In general, the inspectors found that the
containment was clean and free of items of a substantial size that may impact the
performance of the containment sumps. However, the inspectors identified an
appreciable quantity of smaller debris consisting of wires, foam insulation, a spray
bottle, mopheads, and a plastic garbage bag that were not removed. A majority of the
articles identified by the inspectors were located on the 105' level and were apparently
left behind from untracked work occurring on May 20, 2006. These items would have
had access to the containment sump screens during loss of coolant accident (LOCA)
conditions; however, could only have partially affected one of the containment sump
screens while the remaining two would have been fully functional.
Based on the amount, characterization, and location of the debris, the inspectors
concluded that in the aggregate, the licensee failed to follow adequate procedural
guidance to ensure the containment environment was acceptable for power operations.
The specific corrective action document initiated by the licensee associated with the
inspector identified issues/discrepancies is CR 2006104914.
12
Analysis: The performance deficiency associated with this issue is the failure to meet
the acceptance criteria of FNP-1- STP-34.0, Containment Inspection (General). The
procedure requires the containment general area to be free of foreign objects unless
they are identified on a tracking log sheet in procedure FNP-1-STP-34.1. The items
found by the inspectors were present at the time of FNP-1- STP-34.0 completion and
inadequate measures were taken to ensure their removal. The inspectors referenced
Inspection Manual Chapter (IMC) 0612 and determined the finding is more than minor
because it could be reasonably viewed as a precursor to a significant event involving
debris accumulation on the containment sump screens and a subsequent impairment to
suction flow for Emergency Core Cooling System (ECCS) pumps. Although the amount
of debris identified would not have a significant impact on the operability of all three
containment sump screens, the continued accumulation of debris could impact long-
term post accident sump functionality. The inspectors further referenced IMC 0609 for
the SDP review and determined the finding was of very low safety significance.
Although it impacted the Mitigating System Cornerstone, it did not result in a loss of
function per Part 9900, Technical Guidance, Operability Determination Process for
Operability and Functional Assessment, did not represent an actual loss of safety
function, and was not potentially risk significant due to possible external events. No
specific cross-cutting areas were identified.
Enforcement: Technical Specification 5.4, Procedures, requires that written procedures
be implemented for those systems referenced in Regulatory Guide 1.33, Revision 2,
Appendix A, February 1978, which would include the Emergency Core Cooling System.
Contrary to the above, on May 20, 2006 the licensee performed FNP-1-STP-34.0,
Containment Inspection (General) and identified satisfactory completion, even though
the acceptance criteria was not met regarding debris in containment. The debris was
not tracked by any mechanism and remained in the general containment areas until
identified by the resident inspectors on May 21, 2006. This resulted in items being left
behind that could have had an adverse impact on post-accident containment sump
recirculation flow. Because this finding is of very low safety significance and because it
was entered into the licensees corrective action program (CR 2006104914), this
violation is being treated as a NCV, consistent with Section VI.A of the NRC
Enforcement Policy: NCV 05000348/200600301, Failure to Follow Established
Procedures Regarding Debris in Containment.
.2
Failure to Control Contractors Results in Fuel Transfer System Damage
Introduction: A Green non-cited violation (NCV) was identified for failure to control
contractors during the refueling outage that resulted in damage to the fuel transfer
system. This was a self-revealing violation when a pillar block weld broke resulting in
damage to the transfer cart, rails, basket, and dummy fuel assembly.
Description: During the Unit 1 refueling outage, contractor support personnel were
asked to respond to a slack upender cable issue in the spent fuel pool area during core
offload. Core offload occurred during April 15-17, 2006. Using a digital control box,
vendor support personnel made several adjustments to the transfer system cable travel
length. The impact of these adjustments was not understood. These adjustments
caused the pivot points of the upender and basket to be misaligned such that one
13
contacted the other when taken to the vertical position. Because of the adjustments, the
upender and transfer carts alignment marks were misaligned by several inches. Over
time stresses were put on the welds until one failed resulting in the damage to the
system. The failure occurred during checks of the system with a dummy fuel assembly.
The dummy assembly was damaged. The Unit 2 dummy assembly was used to check
the system out following repairs.
Analysis: The inspectors referenced Inspection Manual Chapter (IMC) 0612 and
determined the finding is more than minor because it could be reasonably viewed as a
precursor to a significant event involving damage to a fuel assembly. Although the
damage occurred to a dummy fuel assembly the stresses applied to the fuel transfer
system occurred during core offload and it is was fortuitous that the failure happened
when a dummy assembly was in the basket. The inspectors reviewed IMC 0609 for the
SDP review and determined the finding to be of very low safety significance. Although it
impacted the Barrier Integrity Cornerstone, there was no damage to a fuel assembly.
Enforcement: 10 CFR 50, Appendix B, Criterion VII, Control of Purchased Material,
Equipment, and Services, requires that the effectiveness of the control of contractors
shall be assessed at intervals consistent with the importance, complexity, and quantity
of the product or services. Contrary to the above, Unit 1 contractor support personnel
for refueling activities during April 2006, were not adequately assessed to understand
the impact of adjustments the contractor personnel made to the transfer system cable
travel during core offload to prevent damage to the fuel transfer system. Because this
finding is of very low safety significance and because it was entered into the licensees
corrective action program (CR 2006104091), this violation is being treated as a NCV,
consistent with Section VI.A of the NRC Enforcement Policy: NCV
05000348/200600302, Failure to Control Contractors Results in Fuel Transfer System
Damage.
1R22
Surveillance Testing
a.
Inspection Scope
The inspectors reviewed surveillance test procedures and either witnessed the test or
reviewed test records for the following eight surveillance tests to determine if the tests
adequately demonstrated equipment operability and met the TS requirements. The
inspectors reviewed the activities to assess for preconditioning of equipment, procedure
adherence, and valve alignment following completion of the surveillance. The
inspectors reviewed licensee procedures FNP-0-AP-24, Test Control; FNP-0-M-050,
Master List of Surveillance Requirements; and FNP-0-AP-16, Conduct of Operations;
and attended selected briefings to determine if procedure requirements were met.
Surveillance Tests
C FNP-2-SOP-17.0, Appendix 4, MSIV Functional Test
C FNP-1-STP-11.15, Residual Heat Removal (RHR) Heat Exchange Discharge Valve
Mechanical Stop Verification
C FNP-1-STP-40.0, SI with LOSP Test
C FNP-1-STP-22.13, TDAFW Pump Check Valve Flow Verification
14
In-Service Tests (ISTs)
C FNP-1-STP-45.7, MSIV and Bypass Valves Cold Shutdown Inservice Test
Reactor Coolant System (RCS) Leak Detection
C FNP-1-STP-9.0, RCS Leakage
Containment Isolation Valves
C FNP-2-STP-627, Local Leak Rate Testing of Containment Penetrations, Pen 64A
C FNP-2-STP-627, Local Leak Rate Testing of Containment Penetrations, Pen 93
b.
Findings
No findings of significance were identified
1R23
Temporary Plant Modifications
a.
Inspection Scope
The inspectors reviewed the following temporary modification (TM) and associated
10CFR50.59 screening criteria against the system design bases information and
documentation and the licensees temporary modifications procedure FNP-0-AP-8,
Design Modification Control. The inspectors reviewed implementation, configuration
control, post-installation test activities, drawing and procedure updates, and operator
awareness for this TM.
C TM 10613319102, Unit 1 Upender Car Traveler Seismic Clips
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a.
Inspection Scope
The inspectors evaluated one emergency plan drill on June 28, 2006, to verify the
licensee was properly classifying the event, making required notifications, making
protective action recommendations, and conducting self-assessments. The inspectors
used procedure FNP-0-EIP-15.0, Emergency Drills, as the inspection criteria and
observed the drill on June 28 in the Technical Support Center (TSC). The inspectors
reviewed FNP-0-EIP-9.0, Emergency Classification and Actions, and other supporting
procedures to validate the classification of the event made by the licensee. The
inspectors subsequently observed and reviewed the notifications made, communications
between emergency response team members, team work of licensee personnel,
licensee identification of weaknesses and deficiencies, corrective action documentation,
and overall performance.
15
- The June 28 drill consisted of hurricane force winds and partial loss of offsite power
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstones: Occupational Radiation Safety and Public Radiation Safety
2OS1 Access Control To Radiologically Significant Areas (21 Samples)
a.
Inspection Scope
Access Controls Licensee activities for controlling and monitoring worker access to
radiologically significant areas and tasks were evaluated. The inspectors evaluated
changes to and adequacy of procedural guidance; directly observed implementation of
established administrative and physical radiological controls; appraised radiation worker
and health physics technician (HPT) knowledge of and proficiency in implementing
radiation protection activities; and assessed occupational exposures to radiation and
radioactive material.
The inspectors directly observed controls established for workers and HPT staff in
airborne radioactivity area, radiation area, high radiation area (HRA), locked-high
radiation area (LHRA), and very high radiation area (VHRA) locations. Controls and
their implementation for HRA keys and for storage of irradiated material within U1 and
U2 spent fuel pool (SFP) areas were reviewed and discussed in detail. The inspectors
reviewed and evaluated U1 refueling outage tasks including under vessel bare metal
inspection activities; reactor vessel head disassembly; snubber inspection and
maintenance; fuel off-load; valve maintenance and replacement; reactor coolant pump
maintenance; steam generator primary and secondary side maintenance; radioactive
waste (radwaste) handling and storage; and transportation activities. The inspectors
attended pre-job briefings and reviewed radiation work permit (RWP) details to assess
communication of radiological control requirements to workers. Occupational workers
adherence to selected RWPs and HPT proficiency in providing job coverage were
evaluated through direct observations and interviews with licensee staff. Electronic
dosimeter (ED) alarm set points and worker stay times were evaluated against area
radiation survey results and actual dose rates encountered and doses received.
Worker exposure as measured by ED and by licensee evaluations of potential skin
doses resulting from discrete radioactive particle or dispersed skin contamination events
during the 1R20 activities were reviewed and assessed independently. For HRA tasks
involving potentially significant dose rate gradients, e.g., steam generator maintenance
activities, the inspectors evaluated the potential for use of dosimeter multi-badging to
monitor worker exposure.
Postings for access to radiologically controlled areas (RCAs) and physical controls for
the U1 reactor containment and for U1 and U2 reactor auxiliary building (RAB) locations
designated as LHRAs and VHRAs were evaluated during facility tours. The inspectors
independently measured radiation dose rates or directly observed conduct of licensee
16
radiation surveys and results for U1 containment equipment and work locations, U2
drumming/storage room, outside radioactive material storage areas, and selected U1
and U2 RAB locations. All results were compared to current licensee surveys and
assessed against established postings and radiological controls.
Licensee controls for airborne radioactivity areas with the potential for individual worker
internal exposures of greater than 30 millirem (mrem) Committed Effective Dose
Equivalent were evaluated. For selected RWPs identifying potential airborne areas
associated with 1R20 activities, e.g., under vessel maintenance, valve maintenance, U1
seal table equipment maintenance, and reactor vessel stud cleaning activities, the
inspectors evaluated the implementation and effectiveness of administrative and
physical controls including air sampling, barrier integrity, engineering controls, and
postings. Licensee identification and assessment of potential radionuclide intakes by
workers between April 1, 2005 through April 20, 2006, were reviewed and evaluated.
Radiation protection activities were evaluated against Updated Final Safety Analysis
Report (UFSAR), Technical Specification (TS), and 10 Code of Federal Regulations
(CFR) Parts 19 and 20 requirements. Specific assessment criteria included UFSAR
Section 11, Radioactive Waste Management, and Section 12, Radiation Protection;
10 CFR 19.12; 10 CFR 20, Subpart B, Subpart C, Subpart F, Subpart G, Subpart H, and
Subpart J; TS Sections 5.4, Procedures, and 5.7, High Radiation Area Controls; and
approved procedures. Detailed procedural guidance and records reviewed for this
inspection area are listed in Sections 2OS1, 2OS2, 2PS2, 4OA1, and 4OA5 of the report
Attachment.
Problem Identification and Resolution Licensee Corrective Action Program (CAP)
documents associated with access control to radiologically significant areas were
reviewed and assessed. The inspectors evaluated the licensees ability to identify,
characterize, prioritize, and resolve the identified issues in accordance with Nuclear
Management Procedure (NMP)-GM-002, Corrective Action Program, Version (Ver.) 4.
Licensee Condition Report (CR) documents and audits associated with access controls,
personnel monitoring instrumentation, and personnel contamination events were
reviewed. Licensee CAP documents reviewed and evaluated in detail during inspection
of this program area are identified in Section 2OS1 of the report Attachment.
b.
Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (15 Samples)
a.
Inspection Scope
As Low As Reasonably Achievable (ALARA) The inspectors reviewed ALARA program
guidance and its implementation for ongoing refueling outage job tasks. The inspectors
evaluated the accuracy of ALARA work planning and dose budgeting, observed
implementation of ALARA initiatives and radiation controls for selected jobs in-progress,
assessed the effectiveness of source-term reduction efforts, and reviewed historical
dose information.
17
ALARA planning documents and procedural guidance were reviewed and projected
dose estimates were compared to actual dose expenditures for the following high dose
jobs: steam generator (S/G) maintenance, reactor head assembly/disassembly,
scaffolding construction, and work on valves LCV-459/460 near the regenerative heat
exchanger. Differences between budgeted dose and actual exposure received were
discussed with cognizant ALARA staff. Changes to dose budgets relative to changes in
radiation source term and/or job scope were also discussed. The inspectors attended
pre-job briefings and evaluated the communication of ALARA goals, RWP requirements,
and industry lessons-learned to job crew personnel.
The inspectors made direct field or closed-circuit-video observations of outage job tasks
involving S/G maintenance and reactor head disassembly. For the selected tasks, the
inspectors evaluated radiation worker (radworker) and HPT job performance; surveys of
the work areas; appropriateness of RWP requirements; and adequacy of implemented
engineering controls.
Implementation and effectiveness of selected program initiatives with respect to source-
term reduction were evaluated. Chemistry program ALARA initiatives, including crud
burst/cleanup activities, and their effect on U1 containment and auxiliary building dose
rate trends were reviewed. The effectiveness of temporary shielding installed near the
U1 regenerative heat exchanger was assessed through review of pre-shielding versus
post-shielding dose rate data and expected person-rem saved.
Plant exposure history for calendar year (CY) 2002 through CY 2004 (three year rolling
average) was reviewed. The inspectors also reviewed selected monthly dose reports
and daily RWP dose tracking worksheets. In addition, the inspectors examined dose
records of selected declared pregnant workers to evaluate assignment of gestation
dose.
ALARA program activities and their implementation were reviewed against 10 CFR
Part 20, and approved licensee procedures. In addition, licensee performance was
evaluated against guidance contained in Regulatory Guide (RG) 8.8, Information
Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations
will be As Low As Reasonably Achievable and RG 8.13, Instruction Concerning Prenatal
Radiation Exposure. Procedures and records reviewed within this inspection area are
listed in Sections 2OS2 of the report Attachment.
Problem Identification and Resolution The inspectors reviewed selected CRs and a
licensee self-assessment in the area of exposure control. The inspectors evaluated the
licensees ability to identify, characterize, prioritize, and resolve the identified issues in
accordance with NMP-GM-002, Corrective Action Program Ver. 4. Specific CAP
documents reviewed in detail for this inspection area are identified in Section 2OS2 of
the report Attachment.
b.
Findings
No findings of significance were identified.
18
2PS2 Radioactive Material Processing and Transportation (6 Samples)
a.
Inspection Scope
Waste Processing and Characterization Selected liquid and solid radwaste processing
system components were inspected for material condition and for configuration
compliance with the UFSAR and Process Control Program (PCP). Inspected equipment
included the recycle hold-up tanks; supplemental demineralizer system; resin transfer
piping; resin and filter packaging components; and abandoned waste evaporator
equipment. The inspectors discussed component function, equipment operability, and
changes to radwaste storage areas with licensee staff.
The 2004 Annual Radioactive Effluent Release Report and radionuclide
characterizations from January 1, 2005 through Year-to-Date 2006 for each major waste
stream were reviewed and discussed with radioactive waste (radwaste) staff. For
Reactor Coolant System (RCS) filters, SFP filters, and Dry Active Waste (DAW); the
inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of scaling
factors, and examined comparison results between gamma emitting radionuclides
reported in the licensee waste stream characterizations and the vendor laboratory data.
For selected shipments of spent resin and Low Specific Activity (LSA) waste, the
methodology used for waste stream mixing and concentration averaging was evaluated.
The inspectors also reviewed the licensees procedural guidance for monitoring changes
in waste stream isotopic mixtures.
Radwaste processing activities were reviewed for compliance with 10 CFR Part 50.59
and consistency with the licensees current PCP and UFSAR, Chapter 11. Waste
stream characterization analyses and selected shipping records were reviewed against
regulations detailed in 10 CFR Part 20, 10 CFR Part 61, 49 CFR Part 173, and guidance
provided in the Branch Technical Position (BTP) on Waste Classification and Waste
Form. Reviewed documents are listed in Section 2PS2 of the report Attachment.
Transportation The inspectors directly observed preparation activities for a shipment of
contaminated laundry. The inspectors noted package markings and placarding, and
interviewed shipping technicians regarding Department of Transportation (DOT)
regulations. The inspectors observed dose rate surveys of the shipping packages and
compared the results to DOT limits.
Five shipping records were reviewed for consistency with licensee procedures and
compliance with NRC and DOT regulations. The inspectors reviewed emergency
response information, DOT shipping package classification, radiation survey results, and
evaluated whether receiving licensees were authorized to accept the packages. The
licensees procedures for use of Type B shipping casks were compared to
recommended vendor protocols and Certificate of Compliance (CoC) requirements. In
addition, training records for individuals currently qualified to ship radioactive material
were reviewed.
Transportation program implementation was reviewed against regulations detailed in
10 CFR Parts 20 and 71, 49 CFR Parts 172-178; as well as the guidance provided in
NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
19
Documents reviewed during the inspection are listed in Section 2PS2 of the report
Attachment.
Problem Identification and Resolution. The inspectors reviewed and discussed with HP
supervision selected CRs and audits associated with transportation and radioactive
waste processing program activities. The inspectors assessed the licensees ability to
characterize, prioritize, and resolve the identified issues in accordance with licensee
procedure NMP-GM-002, Corrective Action Program, Ver. 4.0.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a.
Inspection Scope
The inspectors sampled licensee submittals for the PIs listed below to verify the
accuracy of the data reported. The PI definitions and the guidance contained in NEI 99-
02, Regulatory Assessment Performance Indicator Guideline, Rev. 2, and licensee
procedure FNP-0-AP-54, Preparation and Review of NRC Performance Indicator Data,
were used to verify procedure and reporting requirements were met.
Mitigating Systems Cornerstone
- Unit 1 and Unit 2 Safety System Functional Failures
The inspectors reviewed samples of raw PI data, Licensee Event Reports (LERs), and
Monthly Operating Reports for the period covering April 2004 through March 2006. The
data reviewed from the LERs and Monthly Operating Reports was compared to
graphical representations from the most recent PI report. The inspectors also examined
a sampling of operations logs and procedures to verify the PI data was appropriately
captured for inclusion into the PI report as well as insuring that the individual PIs were
calculated correctly.
Barrier Integrity Cornerstone
- Unit 1 and Unit 2 Reactor Coolant System Activity
- Unit 1 and Unit 2 Reactor Coolant System Leakage
The inspectors reviewed raw PI data for the period from October 2004 through March
2006 consisting of daily chemistry analysis and daily leak rate logs. The inspectors
reviewed the recent PI report to verify the data was accurately reflected in the report.
Occupational Radiation Safety Cornerstone
20
- Occupational Exposure Control Effectiveness
The inspectors reviewed PI data collected from October 1, through March 30, 2006.
The inspectors assessed CAP records to determine if HRA, VHRA, or unplanned
exposures, which resulted in TS or 10 CFR 20 non-conformances, had occurred during
the review period. In addition, the inspectors reviewed selected personnel
contamination event data, internal dose assessment results, and ED alarms for
cumulative doses and/or dose rates exceeding established setpoints. Documents
reviewed are listed in the Attachment.
Public Radiation Safety Cornerstone
- Radiological Control Effluent Release Occurrences
The inspectors reviewed the PI results for the period of July 1, 2005 through March 31,
2006. For the assessment period, the inspectors reviewed dose totals to the public, out-
of-service (OOS) effluent radiation monitors and selected compensatory sampling data,
and selected CRs related to Radiological Effluent Technical Specific/ Offsite Dose
Calculation Manual issues. The inspectors also reviewed licensee procedural guidance
for collecting and documenting PI data. Documents reviewed are listed in the
Attachment.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1
Daily Review
As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"
and to help identify repetitive equipment failures or specific human performance issues
for follow-up, the inspectors performed a daily screening of items entered into the
licensees corrective action program. This review was accomplished by reviewing daily
hard copy summaries of CRs and by reviewing the licensees electronic CR database.
.2
Annual Sample Review
Work-Around Review
a.
Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
the inspectors performed a detailed review of the work-around lists for Unit 1 and 2
shared and Unit 2 that were in effect on May 7, 2006. The inspectors reviewed the
proposed corrective action and schedule for each item on the work-around list. The
inspectors reviewed the compensatory action and cumulative effects on plant operation.
21
The inspectors verified each item was being dispositioned in accordance with plant
procedure ACP-17, Operator Work-Around.
b.
Findings and Observations
The inspectors found that operator work-arounds were being identified at an appropriate
threshold. Items were entered into the corrective action program or actions taken were
appropriate.
Review of High Head Safety Injection pump replacement issue
a.
Inspection Scope
As a result of NRC awareness that the replacement High Head Safety Injection pump
casing and discharge head had not been stored or controlled in accordance with an
approved 10CFR50 Appendix B Quality Assurance program, inspectors verified
compliance of the replacement activity with applicable portions of the American Society
of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI.
b.
Findings and Observations
Introduction: A Green NCV of 10 CFR 50.55a(a)(2) was identified by the inspectors for
the licensee failing to comply with the ASME Boiler and Pressure Vessel Code, Section
XI, for Class 2 Components. IWA-7520 of the 1989 edition of Section XI requires that as
part of replacement activities, a completed Owners Report for Repairs or
Replacements, Form NIS-2, be maintained; and subsequently submitted in accordance
with the requirements of IWA-6000.
Description: During the scheduled Unit 2 Fall 2005 outage, the licensee replaced
Charging/High Head Safety Injection Pump 2B casing and discharge head with
replacements that had been previously used in a test facility without an approved quality
assurance program. The replacement pump casing and discharge head, originally
designated and N Stamped per ASME Section III for Code Class 2 service, had not
been continuously controlled as a safety-related ASME Code Class 2 part since 1984 in
accordance with the requirements of an approved quality assurance program.
Therefore, the level of quality had to be re-established in order for the pump to be used
as an ASME Section XI replacement. Methods on how to re-establish quality levels
were not addressed in the ASME Code.
After discussions with the NRC staff, (refer to [SNC] letter to the NRC dated December
2, 2005), SNC considered methods to address the question related to control of pump
pressure retaining parts which would allow returning the 2B Charging / High Head
Safety Injection pump to service. After consideration of submitting a relief request, the
use of Generic Letter 89-09, or to disposition the issue within the SNC corrective action
program, SNC determined that the appropriate process to disposition this issue and
return the 2B charging pump to service was for SNC to document an evaluation for the
pump casing and discharge head acceptability within their corrective action program.
Subsequently the pump was replaced using a design change under Job Number E21-
2051013013.
22
As identified in SNCs letter to the NRC dated March 1, 2006, that transmitted Farley
Nuclear Plant Unit 2 Inservice Inspection Summary Report, SNCs evaluation consisted
of a code reconciliation that addressed the use of different codes, evaluated any
differences between design specification requirements and design, and evaluated the
potential effects of the uncontrolled storage of parts. The evaluation included a
conclusion that the parts retained an acceptable level of quality and safety. SNC
considered the Code requirements of the casing and discharge head to be confirmed.
Also identified in the March 1, 2006 letter transmitting the Inservice Inspection Summary
was information that due to the Authorized Nuclear Inservice Inspectors (ANII)
interpretation of Code requirements, the ANII decided not to sign the NIS-2 form for the
Charging / High Head Safety Injection Pump replacement. It was SNCs position that,
based on an ASME Section XI response to SNCs request for a code interpretation, the
replacement parts are fully qualified to be used as permanent ASME Section XI Class 2
replacement parts. The interpretation requested a response to the question Is it a
requirement of IWA-2110 (h) [1989 Edition of ASME Section XI] that the ANII approve
owners corrective action performed under the owners Appendix B program? The reply
from ASME was No.
The inspectors performed an in-house review of available information provided by the
licensee on the replacement pump. The performance of non-destructive inspection,
visual inspection, and reconciliation evaluation details were reviewed by the inspectors,
and satisfactory inservice testing of the replacement pump was observed. The
inspectors reviewed the following documents associated with the Charging / High Head
Safety Injection Pump 2B replacement and did not identify any immediate safety
concerns with the use of the pump, and that except for the ANIIs signature, those
requirements of ASME Section XI, 1989 Edition affecting the structural integrity of the
replacement pump have been essentially met.
- Doc. No. U-418256, ANSI/ASME Code Reconciliation Document
- DOEJ-SS-2052666001-001, Documentation of Engineering Judgment 001 Code
Reconciliation
- DOEJ-SM-2052666001-002, Documentation of Engineering Judgment 002 Storage
Issue for Pump Casing
- DOEJ-SS-2052666001-003, Documentation of Engineering Judgment 003 Design
Deficiencies Evaluation
- RER 2052666001, Design Verification Summary
- NDR-05-003 Attach. 5, Nonconformance Disposition Report
- SNCs response to inspectors questions transmitted in e-mail dated 2/21/06.
However, the inspectors did not consider that the code case interpretation provided by
ASME resolved the lack of signature by the ANII. ASME Section XI, Article IWA-7520 of
ASME Section XI, 1989 Edition also requires a completed Owners Report for Repairs or
Replacements, Form NIS-2, be maintained by the owner, and article IWA-6000 requires
that the NIS-2 form be submitted to regulatory authorities with the Inservice Inspection
Summary Report within 90 days of the completion of the Inservice inspection conducted
during each refueling outage. As identified in SNCs letter to the NRC dated March 1,
2006, that transmitted FNP Unit 2 Inservice Inspection Summary Report, the NIS-2 form
was not completed. The NIS-2 form was not signed by the ANII, and SNC did not fully
23
explain why. Review of the documentation by the inspectors suggested that the ANII
remains concerned with the replacement part being out of control of an Appendix B
program for many years. The NRC holds the signature of the ANII in high regard and
relies on the ANII approval of such actions to assure Code compliance. An incomplete
NIS-2 form, indicates that all requirements of the Code have not been complied with.
Analysis: The performance deficiency was that the licensee did not meet Code by not
having the NIS-2 form completed. The finding is more than minor because it affects the
Mitigating Systems Cornerstone objective to assure the reliability of systems that
respond to events to prevent undesirable consequences and is associated with the
design control attribute in that qualification remains questionable because there is a lack
of assurance that all code requirements have been met.
The inspectors processed the finding using Appendix A of Inspection Manual Chapter 0609. The finding was evaluated using the Phase 1 screening worksheet and because
it represents a qualification deficiency confirmed not to result in a loss of operability, it
was determined that the deficiency is of very low risk significance (Green). This finding
has been entered into the licensees Corrective Action Program as CR 2006104737.
Enforcement: 10CFR50.55a(a)(2) requires Nuclear Class 1, 2 and 3 components meet
the requirements of the Section XI of the ASME Code. The 1989 Edition of Section XI
requires a completed NIS-2 Form signed by an ANII that documents that the owner has
performed examinations and taken corrective measures in accordance with all the
requirements of ASME Section XI. Contrary to the above, for the Unit 2B Charging/High
Head Safety Injection pump replacement, the licensee did not have a completed NIS-2
Form.
Because this finding is of very low safety significance and has been entered into the
licensees corrective action program as CR 2006104737, this violation is being treated
as a non-cited violation (NCV) consistent with Section VI.A of the NRC Enforcement
manual (NCV 05000364/2006003-03, Failure to Meet Pump Code
/Requirements/Details).
.3
Semi-Annual Trend Review
a.
Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
the inspectors performed a review of the licensees CAP and associated documents to
identify trends that could indicate the existence of a more safety significant safety issue.
The inspectors review focused on CRs with corrective action that were not sufficiently
comprehensive to reduce the likelihood or prevent recurrence of the condition. The
review also considered the results of the daily inspector CAP item screening discussed
in Section 4OA2.1, licensee trending efforts, and licensee human performance results.
The inspectors reviewed the licensee quarterly trend reports for November 2005 -
January 2006, and February - April 2006, daily CRs, selected completed CRs,
Maintenance Rule (a)(1) list, equipment health reports, and quality assurance reports to
identify issues not recognized by the licensee. The inspectors compared and contrasted
their results with the results contained in the licensees quarterly trend reports.
Corrective actions associated with a sample of the issues identified in the licensees
trend report were reviewed for adequacy. The inspectors also evaluated the reports
24
against the requirements of the licensees CAP procedures FNP-0-AP-30.0, Corrective
Action Reporting and NMP-GM-002-GL05, Corrective Action Program Trend Coding and
Analysis Guideline, and the requirements of 10 CFR 50, Appendix B.
b.
Findings and Observations
No findings of significance were identified. The inspectors noted that the licensee had
identified adverse trends with quality of physical work and events related to 4160V
breakers. These were the two most significant issues noted by the inspectors during the
semi-annual review period. Quality of physical work is currently being addressed by the
Maintenance department with action item due dates of August 2006. The events
surrounding 4160V breakers will continue to be evaluated into the next quarter (May
2006, June 2006) and have not been analyzed by the licensee as its trend quarter has
not ended. A separate observation was noted by the inspectors. During a review of the
last two quarters of the Farley Key Performance Indicators, there has been little
increase in the Reactivity Management Index (RMI) with a significant number of
reactivity events. The RMI has remained in a Green overall status over the last twelve
months in the presence of twenty-two level 3 and 4 reactivity management events.
When input to the RMI calculation, the severity level 3 and 4 reactivity issues do not
have significant worth. This would require 20-40 incidents over a twelve month period to
change RMI color and require additional licensee attention. The inspectors identified
this to the licensee and the licensee is evaluating a potential change to the RMI
calculation (SNC fleet wide) that could improve the threshold for monitoring these
events to be more useful in recognizing and addressing potential problem areas. The
licensee has addressed this issue in CR 2006105917.
4OA5 Other Activities
.1
Operation of an Independent Spent Fuel Storage Installation (ISFSI)
a.
Inspection Scope
Inspectors reviewed selected ISFSI operations records to verify that the licensee had
properly identified each fuel assembly in the two latest casks placed on the ISFSI pad.
The inspectors also reviewed Technical Specifications to verify that the fuel placed in
these casks met the requirements. The inspectors also reviewed ISFSI document
control practices to verify that the required records were being retained and duplicate
records were being kept at a separate location. The inspectors walked down the ISFSI
pads to assess the material condition of the casks, the installation of security
equipment, and the performance of the monitoring systems.
b.
Findings
No findings of significance were identified.
.2
Independent Spent Fuel Storage Installation (ISFSI) Radiological Controls
a.
Inspection Scope
The inspectors reviewed gamma-ray, neutron, and contamination surveys of the ISFSI
facility. Inspectors also observed routine gamma-ray surveys and compared the results
25
to previous surveys and TS limits. The inspectors evaluated implementation of
radiological controls, including labeling and posting, and discussed controls with an HP
Technician and HP supervisory staff. Environmental monitoring for direct radiation from
the ISFSI was reviewed, and inspectors observed placement of thermoluminescent
dosimeters.
Radiological control activities for ISFSI areas were evaluated against 10 CFR Part 20,
10 CFR Part 72, and Admentment 2 to the Certificate of Compliance No. 1014 TS
details. Documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified.
.3
(Closed) NRC Temporary Instruction (TI) 2515/165: Operational Readiness of Offsite
Power and Impact on Plant Risk
The inspectors reviewed licensee procedures and controls and interviewed operations
and maintenance personnel to verify these documents contained specific attributes
delineated in the TI to ensure the operational readiness of offsite power systems in
accordance with plant Technical Specifications; the design requirements provided in 10 CFR 50, Appendix A, General Design Criterion 17, "Electric Power Systems;" and the
impact of maintenance on plant risk in accordance with 10 CFR 50.65(a)(4),
"Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants." Documents reviewed are listed in the Attachment. Appropriate documentation
of the results of this inspection was provided to NRC headquarters staff for further
analysis, as required by the TI. This completes the Region II inspection TI requirements
for the Joseph M. Farley Nuclear Station.
.4
(Closed) Unresolved Item [URI] 50-348, 364/2004-04-02, Non-conservative Acceptance
Criteria Used For Service Water Pump Testing
The above URI was opened because the licensee was testing the Component Cooling
Water (CCW) heat exchangers just prior to refueling outages after they had been
cleaned. Therefore, they were potentially getting incorrect as-found performance data
on heat exchanger capability.
In August, 2005, the licensee tested the worst case Unit 2 heat exchanger at the end of
an operating cycle, prior to cleaning. The inspectors reviewed the data and discussed
the data collection methods and results with the licensee. Farley corrective action
Condition Report 2004102818 documented the result and the technical justification,
RER C051797501. The licensees conclusion was that the CCW exchangers are fully
capable of performing their function under worst case conditions between their
scheduled once-per-cycle cleanings. With the attendant changes the licensee had
made to their preventive maintenance program regarding cleaning of the river intake
structure and ultimate heat sink pond, the inspectors concluded the licensees response
to CR 2004 102818 was acceptable.
26
4OA6 Meetings, Including Exit
On July 7, 2006, the inspectors presented the inspection results to Mr. Randy Johnson
and the other members of his staff who acknowledged the findings. The inspectors
confirmed that proprietary information was not provided or examined during the
inspection.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
W.L. Bargeron, Assistant General Manager - Operations
W. R. Bayne, Performance Analysis Supervisor
S. H. Chestnut, Engineering Support Manager
P. Harlos, Health Physics Manager
L. Hogg, Security Manager
J. Horn, Training and Emergency Preparedness Manager
J.R. Johnson, Plant General Manager
T. Livingston, Chemistry Manager
B. L. Moore, Maintenance Manager
W. D. Oldfield, Quality Assurance Supervisor
J. Swartzwelder, Work Control Superintendent
R. J. Vanderbye, Emergency Preparedness Coordinator
R. Wells, Operations Manager
T. L. Youngblood, Assistant General Manager - Plant Support
NRC personnel
S. Shaeffer, Division of Reactor Projects, Branch Chief
D. Simpkins, Division of Reactor Projects, Acting Branch Chief
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000348/2006-003-01
Failure to Follow Procedures for Containment Closeout
(Section 1R20)
05000348/2006-003-02
Failure to Control Contractors Results in Fuel Transfer
System Damage (Section 1R20)
05000364/2006-003-03
Failure to Meet Pump Code Requirements/Details (Section
4OA2.2)
Closed
05000348, 364/2515/165
TI
Operational Readiness of Offsite Power and Impact on
Plant Risk (Section 4OA5.3)
05000348, 364/2004-04-02 URI
Non-conservative Acceptance Criteria Used For Service
Water Pump Testing (Section 4OA5.4)
Discussed
None
2
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R04: Equipment Alignment
UFSAR Section 5.5.7
Technical Specification Section 3.5.2
Student Lesson Plan OPS-52102B-40302C Emergency Core Cooling
RHR Function System Description A-181002
Section 1R08: Inservice Inspection Activities
NMP-ES-024-502, PDI Generic Procedure for the Ultrasonic Examination of Ferritic Pipe Welds
(Appendix VIII), Ver 1.0
NMP-ES-024-501, PDI Generic Procedure for the Ultrasonic Examination of Austinetic Pipe
Welds (Appendix VIII), Ver 1.0
FNP-0-NDE-100.34, Nondestructive Examination Procedure
Boric Acid Corrosion Control Program Quarterly Report 4th Quarter - October 1, 2005 -
December 31, 2005
NMP-ES-019, Boric Acid Corrosion Control Program, Ver 1.0
NMP-ES-019-GL01, Boric Acid Corrosion Control Implementation Guideline, Ver 1.0
FNP-0-M-101, Boric Acid Corrosion Control Program, Ver 10
FNP-0-EPT-4496, Corrosion Assessment, Ver 2.0
Westinghouse MRS-SSP-1169-ALA/APR, Rev. 2, Farley Units 1 & 2 In-Service Steam
Generator Eddy Current Analysis Guidelines
Westinghouse MRS-SSP-1052-ALA/APR, Rev. 6, Secondary Side Tubesheet Inspection -
Westinghouse Model 54F Sgs
Westinghouse MRS 2.4.2 GEN-35, Rev. 11, Eddy Current Inspection Of Preservice and
Inservice Heat Exchanger Tubing
MIS-FNP-5-17, Review of Farley 1R19 Steam Generator Secondary Side Loose Parts for Next
Cycle of Operation, Dated November 4, 2004
Steam Generator Degradation Assessment for Farley 1R20 Inservice Inspection for Model 54F
Steam Generator History and Operating Assessment
Eddy Current Analyses Calibration Nos. 19 for SG B tube R45C59; 9 for SG A tubes R17C04;
11 for SG A tubes R15C10 and R26C10; 41 for SG C tubes R11C82, R12C82, and R10C93;
and 43 for SG C tubes R04C51, R03C56, and R13C81
Specific Assessment of Potential Degradation Mechanism Developed for the Upcoming Outage
Data Acquisition and Analysis Personnel Qualification for Level II Data Operators, Level II A
and Level III A Analysts
Farley Unit 1, ALA R20 Steam Generator Inspection Reference Manual
Westinghouse Reference Letter LTR-SGDA-04-326, Evaluation of Loose Parts for Farley Unit 1
Fall 2004 Outage, Dated October 29, 2004
RER C049635101, to initiate, Implement, and Resolve Loose Parts in the Secondary Side of
Steam Generators Due to Flex Gaskets from the Broken Spiral Wound Gaskets Used in Valves
CRs: 2005112010, 2004000259, 2004104601, 2005111036, 2004104465, 2006103677,
2006103737, 2006103680, 2006103746, 2005200863
3
Attachment
Section 1R11: Licensed Operator Requalification
FNP-1-AOP-2.0, Steam Generator Tube Leakage
FNP-1-AOP-4.0, Loss of Reactor Coolant Flow
FNP-1-AOP-17.0, Rapid Load Reduction
FNP-1-AOP-100, Instrumentation Malfunction
FNP-1-UOP-3.1, Power Operation
FNP-0-ESP-0.1, Reactor Trip Response
FNP-0-EIP-9.0, Emergency Classification and Actions
Section 1R20: Refueling and Outage Activities
FNP-0-UOP-4.0, General Outage Operations Guidance
FNP-0-AP-52, Equipment Status Control and Maintenance Authorization
FNP-1-UOP-4.1, Refueling Outage Operation
FNP-0-AP-94, Outage Nuclear Safety
FNP-1-UOP-4.3, Mid-Loop Operations
FNP-0-ACP-47.3, Outage Preparation
FNP-1-STP-35.0, Reactor Coolant System Pressure and Temperature/Pressurizer
Temperature Limits Verification
FNP-1-UOP-2.1, Shutdown of Unit From Minimum Load to Hot Standby
FNP-1-UOP-2.2, Shutdown of Unit From Hot Standby to Cold Shutdown
FNP-1-SOP-1.6, Draining th Reactor Coolant System
FNP-1-SOP-1.3, Reactor Coolant System Filling and Venting-Vacuum Method
FNP-1-STP-18.4, Ctmt Mid-Loop and/or Refueling Integrity Verification and Ctmt Closure
FNP-1-IMP-201.45, Refueling Reactor Coolant System Level Calibration Q1B21FT0416
FNP-1-STP-35.1, Unit Startup Technical Specification Verification
FNP-0-ETP-3643, Verification of Rod Control System Availability
FNP-1-STP-101, Zero Power Reactor Physics Testing
FNP-1-STP-29.6, Calculation of Estimated Critical Condition
Section 2OS1: Access Controls to Radiologically Significant Areas
Procedures, Manuals, and Guidance Documents
Farley Nuclear Plant (FNP) Dosimetry Procedure (DOS)-1, Personnel Monitoring, Version
(Ver) 43.0
FNP-1- FHP - 1.0, Refueling Operations, Ver. 8.0
FNP-0-ACP- 7.0, Foreign Material Exclusion Program, Ver. 16.0
FNP -0-RPC -0.1, Key Control Program and Health Physics Guidance for Control of High
Radiation Areas, Exclusion Areas (Locked High Radiation Areas) and Very High Radiation
Areas, Ver. 9.0
FNP-1-RPC -0.2, Unit 1 Reactor Vessel Maintenance Sump Entry, Ver. 2.0
FNP 0-RPC-4, Refueling Survey, Ver. 18.0
FNP-0-RPC-26, Radiological Surveys and Monitoring, Ver. 32.0
FNP-0-RCP-26, Temporary Change Notice Form (TCNF), Radiological Surveys and Monitoring,
Ver. 32.1
FNP-0-RPC-29, Contamination Guidelines, Ver. 38.0
FNP-0-RPC-29.1, Guidelines for Personnel Decon and Response to Personnel Contamination
4
Attachment
Events, Ver. 5.0
FNP-0-RCP-57, Radioactive and Potentially Radioactive Material Handling, Ver. 28
FNP-0-RPC-114, Operation and Care of Safety and Supplys Model RC 2095, LANCS Model
LI-520Y Series All Clear Air-Fed Hood, and Defense Apparels HSQ-10 Supplied Air Hood
Assembly, Ver. 7.0
FNP-0-RPC-367, Radiological Control Associated with Primary Steam Generator
Channelheads, Ver. 36.0
Records and Data Reviewed
Unit 1 (U1) and Unit 2 (U2) Spent Fuel Pool (SFP) Trash (Non-Fuel Items Inventory), as of
04/20/2006
U1 ICA-3, U1 SFP Inventory Map, as of 04/20/2006
U2 ICA-6, U2 SFP Inventory Map, as of 04/20/2006
U1 SFP Debris Canister Stack Sequence: Long Term and Short Term Storage Items (Data as
of 04/21/2006).
U2 SFP FSDC Stack Sequence: Long Term and Short Term Storage Items as Loaded for
FATF 02-99-021, (Data as of 04/21/2006)
Radiation Work Permit (RWP) 06-0501, Operations: Routine Inspections, Activities and
Training Within Radiation Controlled Areas by Operations Personnel. CAUTION: Do Not use
this RWP for Containment (CTMT) Entries, Rev. 0
RWP 06-0703, Engineering Support (ES): All Work Associated with Fuel Inspections, Fuel
Oxide Measurements, Top Nozzle Replacement, etc., in the Unit 1 (U1) or Unit 2 (U2) Spent
Fuel Pool (SFP). NOTE: Not for New Fuel Receipt. CAUTION: Do Not Use this RWP for
CTMT Entries, Rev. 0
RWP 06-0705, ES: All Work Associated with SFP Inventory Classification in the U1 or U2 SFP.
Caution: Do Not Use this RWP for CTMT Entries, Rev. 0
RWP 06-1444, Maintenance - All; All Work Associated with Repairs, Inspections, and
Observations in the Unit 1 Reactor Vessel Maintenance Sump to Support the 1R20 Outage.
No Entry While Incore Thimbles are Withdrawn or While Driving Detectors, Rev. 0
RWP 06-1453, Maintenance-MM; All Work Associated with Maintenance on Valves Reading
1000 mRem/hr or Greater Contact and not Specifically Covered by Other RWPs in the
Auxiliary Building and Unit 1 CTMT to Support the 1R20 Outage, Rev. 0
RWP 06-1467, Maintenance-MM, FAC,HP,WMS; All Work Associated with Cleanout and
Maintenance of the U1 Reactor Cavity Transfer Canal to Support the 1R20 Outage. (To
Include Decon of the Upender Sump, Blind Flange Work, and all Repairs). Rev. 0
RWP 06-1488, Maintenance-MM, WMS, Wyle; All Work Associated with Routine Snubber Work
in the Auxiliary Building, the Unit 1 Containment, and the Snubber Trailer/Building to Support
the 1R20 Outage, Rev. 0
RWP 06-1504, Operations-OPS, ES, Westinghouse; All work to Support for Refueling Activities
in the U1 Containment and the U1 SFP Room during the 1R20 Outage, Rev.0
RWP 06-1707, Maintenance - All: All Work Associated with Visual Bottom Mounted
Instrumentation (BMI) Penetration Inspection of the U1 Containment Reactor Vessel
Maintenance Sump during the 1R20 Outage. No Entry While Incore Thimbles are Withdrawn
or While Driving Detectors, Rev. 0.
RWP 06-1730, Engineering-ES; Westinghouse, ES, HP; All Work Associated with the
Installation and Removal of Nozzle Dam/Nozzle Covers in the Primary Steam Generators to
Support the 1R20 Outage, Rev. 0
RWP 06-1731, Engineering-ES; Westinghouse, ES, HP; All Work Associated with the Primary
Steam Generator Eddy Current Testing and Repairs to Support the 1R20 Outage, Rev. 0
5
Attachment
Survey Number (No.) 13991, Unit 1 Reactor Cavity (1CB155), Dose Rate Data, 11/11/04
Survey No. 22708, U-1 Auxiliary Building 83 foot elevation (1AB83), Dose Rate and
Contamination Data, 03/19/06
Survey No. 23275, 1AB83, Dose Rate and Contamination Data, 04/09/2006
Survey No. 23317, 1AB83, Dose Rate and Contamination Data, 04/10/2006
Survey No. 23360, 1Auxiliary Building RHR Pump Room (1AB83131) Dose Rate and
Contamination Data, 04/10/06
Survey No. 23403, U1 Containment Sump (1CB) Dose Rate and Contamination Data,
04/11/2006
Survey No. 23371, HP Derived Air Concentration Report, U1 Reactor Sump Air Sample,
04/11/2006
Survey No. 23457, U1 Containment Transfer Canal (1CB155) Dose Rate Data, 04/12/2006
Health Physics Survey (HPS) Data (Airborne) Associated with Reactor Stud Cleaning Activities:
HPS No. 23541, 04/15/2006; No. 23611, 04/15/2006; No. 23701, 04/16-17/2006; No. 23654,
04/16/2006;
Record of Dose Rates During Initial Fuel Assembly Transfer, 04/15/2006
Health Physics Locked High Radiation Area (LHRA) Control Log, Data 04/10-11/2006
Dose Report Data for: RWP 06-1444, Containment Reactor Vessel Maintenance Sump
Activities, 04/11/2006; RWP 06-1453, Maintenance on Valves $1000 mRem/hr, 04/20/2006;
RWP 06-1467, Cleanout and Maintenance of the U1 Rx Cavity Transfer Canal to Support the
1R20 Outage, 04/20/2006; RWP: 06-1707, Containment Reactor Vessel Maintenance
Sump Activities, 04/11/2006; RWP 06-1731, Primary Steam Generator Eddy Current Testing
and Repairs, 04/20/2006; RWP 06-1739, Installation & Removal of Nozzle Dam/Nozzle
Covers in Primary Steam Generator, 04/20/2006
Corrective Action Program (CAP) Documents
Nuclear Management Procedure (NMP) General Management ((GM) - 002, Corrective Action
Program, Ver. 4
Farley Nuclear Plant Health Physics Self Assessment, INPO 03-004 Guide (Radiation
Protection), 12/06-10/2004
Review of Locked High Radiation Area (LHRA) Administrative Controls, 08/09-12/2005
Condition Report (CR) 2005104299, HP Individual Logged-in on the Wrong RWP , 04/28/2005
CR 2005110573, Containment Entry Made on Auxiliary Building RWP, 10/24/2005
CR 2005110695, Inadvertent Sign-In to Wrong RWP, 10/26/2005
CR 2005105623, Maintenance Worker Direct Alarming Dosimeter Not Operating During RCA
Entry, 06/09/2006
CR 2005107891, Loss of Controls Securing Tri-Nuke Equipment Within the U1 Spent Fuel
Pool, 08/05/2005
CR 2005111645, Very High Radiation Area Key Lost, 11/15/2005
Section 2OS2: As Low As Reasonably Achievable (ALARA)
Procedures, Manuals, and Guidance Documents
FNP-0-RCP-7, Coordinated Exposure Reduction Program, Ver. 2.0
FNP-0-RCP-15, Temporary Shielding, Ver. 35.0
FNP-0-RCP-19, Pre and Post Job ALARA Planning for Work in Radiation Controlled Areas of
the Plant, Ver. 19.0
Chemical Degassing/Crud Burst 1RFO20
NMP-GM-002, Corrective Action Program, Ver. 4.
6
Attachment
Records and Data Reviewed
RWP Dose Tracking Records, 4/10/06 - 4/21/06
Plant ALARA Review Committee meeting minutes, 4/6/06
Monthly ALARA Report, March 2006
Declared Pregnant Worker Dosimetry Records, 2003 - 2004
1R20 ALARA Planning Checklists: Steam Generator Eddy Current Testing and Reactor Head
Assembly/Disassembly
RWP 06-1731, Steam Generator Eddy Current Testing, Rev. 0
RWP 06-1461, Reactor Head Assembly/Disassembly, Rev. 0
RWP 06-1481, All Work Inside U1 Regenerative Heat Exchanger Fence, Rev. 1
Survey 23241, U1 Regenerative Heat Exchanger Valves (Pre-Shield), 4/8/06
Survey 23427, U1 Regenerative Heat Exchanger Valves (Post-Shield), 4/11/06
Survey 23779, U1 C Steam Generator Hot Leg, 4/18/06
Crud Burst/Cleanup Survey Point Data, 1R17 - 1R20
CAP Documents
Health Physics Self-Assessment, INPO 03-004 Guide (Radiation Protection), 12/6/04 - 12/10/04
CR 2005111491, Improvement needed in determining the total job scope of scaffold
requirements for an outage, 11/11/05
CR 2005102060, Plant dose is at 161% of budget dose for the first month of the year, 2/18/06
CR 2004104041, Request an engineering evaluation to consider permanent shielding on U1
regenerative heat exchanger, 10/10/04
CR 2006101009, Increased dose to replace 1A incore detector due to new equipment issues,
2/3/06
Section: 2PS2 Radioactive Material Processing and Transportation
Procedures, Manuals, and Guidance Documents
FNP UFSAR, Chapter 11, Radioactive Waste Management
Joseph M. Farley Nuclear Plant Annual Radioactive Effluent Release Report for 2004, 04/25/05
FNP-0-030, Process Control Program, Rev. 15
FNP-0-RCP-801, Disposable Demineralizer System Operation (Hittman), Ver. 16.0
FNP-0-RCP-809, Isotopic Characterization, Scaling Factor Utilization, and Waste Classification
of Radioactive Waste Streams for Offsite Shipments and/or Near Surface Disposal, Ver.16.0
FNP-0-RCP-810, Shipment of Radioactive Waste, Ver. 44.0
FNP-0-RCP-811, Shipment of Radioactive Material, Ver. 34.0
FNP-0-RCP-839, Segregation of Low Level Solid Wastes, Rev. 6
FNP-1-SOP-49.0, Solid Waste Processing System, Ver. 23.0
FNP-1-SOP-50.0, Liquid Waste Processing System, Ver. 52.0
FNP-1-SOP-50.4, Demineralizer Resin Removal and Addition, Ver. 32.0
FNP-0-SOP-50.7, Liquid Waste Processing using the Supplemental Demineralizer System,
Ver. 36
Records and Data
U1 Waste Processing System Piping and Instrumentation Diagram (P&ID) No. D-175042,
Sheets 1, 2, 3,4 and 7
U2 Waste Processing System P&ID No. D-205042, Sheets 1, 2, 3,4 and 7
Radioactive Waste Shipment (RWS) 06-11, Primary Resin, 02/18/06
7
Attachment
RWS 05-24, LWP Mixed Media, 12/02/05
RWS 05-11, DAW Sealands, 07/15/05
RWS 05-05, Primary Resin, 04/21/05
RWS 05-02, Filter HIC, 02/23/05
10 CFR Part 61 Radioactive Waste Stream Analysis Reports, SFP filters, RCS filters, and
DAW, 2004 and 2005
10 CFR Part 61 Radioactive Waste Stream Analysis Report, Filter Smears, HIC C in shipment
RWS 05-02
10 CFR Part 61 In-House to Industrial Laboratory Comparison and Data Set Validation for SFP
filters, RCS filters and DAW, 2005
CoC no. 9208, Model No.10-142 Shipping Package, Rev. 15
Hazardous Material/Waste Handler Training Certificates for shipping staff
CAP Documents
CR 2004106550, FNP QA discovered a weight discrepancy in Radwaste Shipping
paperwork, 12/02/04.
CR 2005100836, QA Audit F-CRW-2004, Comment # 2: Potential misclassification of
radioactive waste shipments, 01/21/05
CR 2005108566, Challenges in cleaning sludge from Floor Drain Tank, 08/26/05
CR 2005111241, When loading DAW found 2 bags of trash with incorrect dose rates marked
on the bags, 11/06/05
CR 2006101693, Written to document contamination issues at the OSGSF fenced RCA
associated with preparations to ship 2 degraded sealand containers, 02/23/06
CR2006102506, Active leaks in the roof of the LLRWB, 03/21/06
CR2006102504, Fire Protection System has leak at the area center of the northern most aisle
in main bay of LLRWB, 03/21/06
Health Physics Self Assessment, December 06, 2004 to December 10, 2004
QA Surveillance # 2004-04, Primary Spent Resin Cask Shipment, 02/17/04
Section: 4OA1 Performance Indicator Verification
Procedures
FNP-0-AP-54, Preparation and Reporting of NRC Performance Indicator Data and NRC
Operating Data, Ver. 6.0
Records and Data Reviewed
Access Control Alarms, Cumulative Dose and Dose Rate Data; October 1, 2005 through
April 18, 2006
Special Report 2005-001-00, Inoperable Radiation Monitor R-29B
OOS effluent monitor logs, 6/1/05 - 3/6/06
Gaseous effluent release permits 50387.027.057.G and 60071.017.010.G
Liquid effluent release permits 50874.021.162.L and 60208.021.023.L
CAP Documents
CR 2005105520, 2B turbine building sump auto sampler failed to collect sample, 6/7/05
CR 2005112468, R-29B inoperable for greater than 7 days, 12/7/05
CR 2006100073, Isolation valve RCV-23B failed to close on high radiation signal from steam
generator blow down monitor R23B during surveillance test, 1/5/06
8
Attachment
Section 4OA2: Identification and Resolution of Problems
April 12, 2006 SNC Letter to Farley Engineering Manager, File: RER C051797501: Log: PS-
06-0117
Section 4OA5: Other Activities
A-173444, Power Quality Guide
BPO-1, Bulk Power Operations
FNP-1-SOP-28.1, Turbine Generator Operations
FNP-0-SOP-38.0, Diesel Generators
FNP-0-ACP-4, Switchyard control
FNP-1-STP-27.1, A.C. Source verification
FNP-0-ACP-52.3, On-Line Risk Assessment
FNP-1-AOP-5.2, Degraded Grid
FNP-2-ECP-0.0, Loss of All AC Power
ISFSI Radiological Controls Procedures, Manuals, and Guidance Documents
FNP-0-STP-822, Hi-Storm Overpack Surface Dose Rates, Ver. 2.0
Certificate of Compliance (CoC) No. 1014 for the Holtec International Hi-Storm 100 Cask
System, Amendment No. 2
Holtec Report No. HI-2053364, Hi-Storm CoC Radiation Protection Program Dose Rate Limits
for SNC, 03/01/05
Surveys, Data, and Records Reviewed
ISFSI Pad Survey # 18151, 08/25/05
ISFSI Pad Survey # 20156, 11/05/05
ISFSI Perimeter TLD Exposure Report - 2nd quarter 2005, 01/27/06
CR 2005110575, Discrepancy in neutron dose measured with DAD versus TLD for worker
involved in the spent fuel dry cask storage campaign, 10/24/05