ML052130486
| ML052130486 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/29/2005 |
| From: | Cahill S Reactor Projects Region 2 Branch 6 |
| To: | Singer K Tennessee Valley Authority |
| References | |
| IR-05-003 | |
| Download: ML052130486 (49) | |
See also: IR 05000259/2005003
Text
July 29, 2005
Tennessee Valley Authority
ATTN.: Mr. K. W. Singer
Chief Nuclear Officer and
Executive Vice President
6A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000259/2005003, 05000260/2005003, AND 05000296/2005003
Dear Mr. Singer:
On June 30, 2005, the US Nuclear Regulatory Commission (NRC) completed an inspection at
your operating Browns Ferry Unit 2 and 3 reactor facilities. The enclosed integrated quarterly
inspection report documents the inspection results, which were discussed on July 7, 2005, with
Mr. B. Aukland, and other members of your staff. Additionally, the enclosed report also
documents some inspection of Unit 1 that was performed per our letter to you on December 29,
2004, regarding transitioning Unit 1 into the Reactor Oversight Program (ROP). In that letter
we indicated that the NRC had determined that the ROP cornerstones of Occupational
Radiation Safety, Public Radiation Safety, Emergency Preparedness, and Physical Protection
would be incorporated into the routine ROP baseline inspection program effective January 1,
2005. Remaining results from our inspection of your Unit 1 Recovery Project continue to be
documented in a separate Unit 1 integrated inspection report.
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 NRC-identified finding and three self-revealing findings of very low
safety significance (Green) which were determined to involve violations of NRC requirements.
However, because of the very low safety significance and because the findings were entered
into your corrective action program, the NRC is treating the findings as non-cited violations
(NCVs) consistent with Section VI.A of the NRC Enforcement Policy
The report details a violation of 10 CFR 50, Appendix R, requirements involving circuit analysis
issues for which the NRC is exercising enforcement discretion and reactor oversight process
discretion (i.e., not subjecting the violation to the significance determination process (SDP)).
The basis for the enforcement discretion is NRC Enforcement Manual Section 8.1.7.1, Fire
Induced Circuit Failures. One of the conditions for applying discretion is that the circuit
vulnerabilities be corrected within a reasonable time frame.
TVA 2
NRC Inspection Manual Chapter 0305, Operating Reactor Assessment Program, Section
06.06.2, Violations in Specified Areas of Interest Qualifying for Enforcement Discretion, states
that violations related to certain circuit issues which are eligible for enforcement discretion shall
also be eligible for reactor oversight process discretion.
If you contest any non-cited violation in the enclosed report, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, 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 Senior
Resident Inspector at the Browns Ferry Nuclear Plant.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of 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/
Stephen J. Cahill, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos. 50-259, 50-260, 50-296
License Nos. DPR- 33, DPR-52, DPR-68
Enclosure: Inspection Report 05000259/2005003, 05000260/2005003 AND
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
TVA 3
cc w/encl: State Health Officer
Ashok S. Bhatnagar Alabama Dept. of Public Health
Senior Vice President RSA Tower - Administration
Nuclear Operations Suite 1552
Tennessee Valley Authority P. O. Box 303017
Electronic Mail Distribution Montgomery, AL 36130-3017
Larry S. Bryant, General Manager Chairman
Nuclear Engineering Limestone County Commission
Tennessee Valley Authority 310 West Washington Street
Electronic Mail Distribution Athens, AL 35611
Brian OGrady Jon R. Rupert, Vice President
Site Vice President Browns Ferry Unit 1 Restart
Browns Ferry Nuclear Plant Browns Ferry Nuclear Plant
Tennessee Valley Authority Tennessee Valley Authority
Electronic Mail Distribution P. O. Box 2000
Decatur, AL 35609
Robert J. Beecken, Vice President
Nuclear Operations Robert G. Jones, Restart Manager
Tennessee Valley Authority Browns Ferry Unit 1 Restart
Electronic Mail Distribution Browns Ferry Nuclear Plant
Tennessee Valley Authority
General Counsel P. O. Box 2000
Tennessee Valley Authority Decatur, AL 35609
Electronic Mail Distribution
Distribution w/encl: (See page 4)
John C. Fornicola, Manager
Nuclear Assurance and Licensing
Tennessee Valley Authority
Electronic Mail Distribution
Bruce M. Aukland, Plant Manager
Browns Ferry Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Glenn W. Morris, Manager
Corporate Nuclear Licensing
and Industry Affairs
Tennessee Valley Authority
Electronic Mail Distribution
William D. Crouch, Manager
Licensing and Industry Affairs
Browns Ferry Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
_________________________
OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS
SIGNATURE TMR RLM per email RPC TMR for HJG for HJG HJG for
NAME TRoss RMonk RCarrion EChristnot RHamilton HGepford DCarter
DATE 07/28/2005 07/28/2005 07/26/2005 07/28/2005 07/28/2005 07/26/2005 07/28/2005
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
PUBLIC DOCUMENT YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE LRM per email JLK MAS MSL for
NAME LMiller JKreh MScott AVargas
DATE 07/26/2005 07/26/2005 07/26/2005 07/28/2005
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
U.S. NUCLEAR REGULATORY COMMISSION REGION II
Docket Nos: 50-259, 50-260, 50-296
License Nos: DPR-33, DPR-52, DPR-68
Report Nos: 05000259/2005-003, 05000260/2005-003,
Licensee: Tennessee Valley Authority (TVA)
Facility: Browns Ferry Nuclear Plant, Units 1, 2 & 3
Location: Corner of Shaw and Nuclear Plant Roads
Athens, AL 35611
Dates: April 1, 2005 - June 30, 2005
Inspectors: T. Ross, Senior Resident Inspector
R. Monk, Resident Inspector
E. Christnot, Resident Inspector
R. Hamilton, Senior Health Physicist (Sections 2PS2,
4OA1)
H. Gepford, Health Physicist (Sections 2OS2)
D. Carter, Health Physicist, Region IV (Section 2OS1)
R. Carrion, Project Engineer (Section 2OS2)
L. Miller, Senior Emergency Preparedness Inspector
(Sections 1EP1,1EP4, 4OA1)
J. Kreh, Emergency Preparedness Inspector (Sections
1EP1, 4OA1)
M. Scott, Senior Reactor Inspector (Section 1R12.2)
A. Vargas, Reactor Inspector (Section 1R08)
Approved by: Stephen J. Cahill, Chief
Reactor Project Branch 6
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000259/2005-003, 05000260/2005-003, 05000296/2005-003; 04/01/2005 - 06/30/2005;
Browns Ferry Nuclear Plant, Units 1, 2, and 3; Maintenance Effectiveness, Access Control To
Radiologically Significant Areas.
The report covered a three-month period of routine inspection by resident inspectors and
announced inspections by regional reactor inspectors, emergency preparedness inspectors,
health physicists, and a project engineer. One NRC-identified finding and three self-revealing
findings of very low safety significance (Green) which also involved violations of regulatory
requirements were identified. The significance of issues is indicated by the color assigned
(Green, White, Yellow, Red) using the Significance Determination Process in Inspection Manual
Chapter 0609, Significance Determination Process (SDP). The NRCs program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Barrier Integrity
- Green. The inspectors identified a non-cited violation of 10CFR50.65(a)(1) in
which the licensee has failed to implement timely and effective corrective
actions to preclude multiple, repetitive failures of containment isolation valves in
the Unit 2 and 3 Residual Heat Removal (RHR) Keep Fill System. These failures
ultimately resulted in the failure of two containment isolation valves
simultaneously for the same penetration, which created an open pathway from
containment and a consequential loss of the maintenance rule safety function.
Licensee monitoring and corrective actions per 10 CFR 50.65(a)(1) were
ineffective at ensuring that containment isolation valves in the RHR Keep Fill
System were capable of performing their intended safety function.
The finding is greater than minor because if left uncorrected it would become a
more significant safety concern and because it affected the Containment
Isolation SSC Reliability objective of the SSC and Barrier Performance attribute
under the Barrier Integrity Cornerstone. The finding was assessed using the
SDP, Manual Chapter 0609, Appendix H, Table 4.1. This assessment
determined the finding to be of very low safety significance because, in the case
of the most consequential containment isolation valve failures, the associated
pathway was a small (i.e., 2-inch) line and would not have significantly
contributed to Large Early Release Frequency (LERF). This finding had
cross-cutting aspects associated with Problem Identification and Resolution.
(Section 1R12)
Cornerstone: Occupational Radiation Safety
- Green. The inspectors reviewed a self-revealing, non-cited violation of Technical
Specification (TS) 5.7.2 resulting from the licensees failure to properly control a
high radiation area with dose rates greater than 1.0 rem per hour at 30
Enclosure
2
centimeters (cm) from the source. Specifically, on February 5, 2004, an operator
entered the waste backwash transfer pump room on the 546-foot elevation of the
radwaste building and received a electronic dosimeter dose rate alarm. A survey
of the area identified dose rates of 10,000 millirem (mrem) per hour on contact
and 1500 mrem per hour at 30 cm on a section of pipe. The area was
immediately controlled as a locked high radiation area.
The finding is greater than minor because it is associated with the Occupational
Radiation Safety cornerstone attribute of exposure control and it affected the
associated cornerstone objective to ensure the adequate protection of worker
health and safety from exposure to radiation. Using the Occupational Radiation
Safety Significance Determination Process, the finding was determined to be of
very low safety significance because it did not involve: (1) As Low As
Reasonably Achievable (ALARA) planning and controls, (2) an overexposure, (3)
a substantial potential for overexposure, or (4) an impaired ability to assess
dose. (Section 2OS1)
- Green. The inspectors reviewed a self-revealing, non-cited violation of TS 5.7.1
resulting when operations personnel failed to inform radiation protection
personnel of the operation of the waste backwash transfer pump which caused
an increase in dose rates to high radiation area levels. Specifically, on
November 24, 2004, a radwaste operator received an electronic dosimeter dose
rate alarm when he entered the waste surge and collector pump room on the
546-foot elevation of the radwaste building. The operator entered an area with
dose rates of 159 mrem per hour and received a dose of 5 mrem from the entry.
A survey of the area showed contact dose rates with overhead piping were as
high as 2500 mrem per hour, with general area dose rates of 300 mrem per
hour.
The finding is greater than minor because it is associated with the Occupational
Radiation Safety cornerstone attribute of exposure control and it affected the
associated cornerstone objective to ensure the adequate protection of worker
health and safety from exposure to radiation. Using the Occupational Radiation
Safety Significance Determination Process, the finding was determined to be of
very low safety significance because it did not involve: (1) ALARA planning and
controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4)
an impaired ability to assess dose. The cause of this finding had cross-cutting
aspects associated with human performance. (Section 2OS1)
- Green. The inspectors reviewed two examples of a self-revealing, non-cited
violation of TS 5.4.1 for the failure of workers to comply with radiation work
permit (RWP) requirements. The first example occurred on March 22, 2004,
when an operator entered a posted high radiation area on an RWP that did not
allow entry into high radiation areas. The operator received a electronic
dosimeter dose rate alarm. Radiation dose rates in the area were 600 mrem per
hour on contact and 300 mrem per hour at 30 cm from the radiation source. The
second example occurred on October 4, 2004, when a craft worker entered an
Enclosure
3
area in the overhead, greater than 6 feet, of the Unit 1 reactor building 593-foot
elevation without contacting radiation protection personnel as required by the
RWP. The worker did not review the planned work with radiation protection
personnel prior to entry and did not monitor electronic dosimetry prior to reaching
the dose alarm setpoint. A survey of the overhead area indicated dose rates of
200 mrem per hour on contact, 60 mrem per hour at 30 cm, and 25 mrem per
hour general area from overhead piping.
The finding is greater than minor because it was associated with the
Occupational Radiation Safety cornerstone attribute of program and process and
it affected the associated cornerstone objective to ensure adequate protection of
worker health and safety from exposure to radiation. Using the Occupational
Radiation Safety Significance Determination Process, the finding was
determined to be of very low safety significance because it did not involve
(1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential
for overexposure, or (4) an impaired ability to assess dose. In addition, this
finding had cross-cutting aspects associated with human performance when
personnel failed to follow radiation work permit instructions. (Section 2OS1)
B. Licensee Identified Findings
One violation of very low safety significance was identified by the licensee and has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation is listed in
Section 4OA7 of this report.
Enclosure
Report Details
Summary of Plant Status
Unit 1 was defueled and in a recovery status.
Unit 2 began the report period in the midst of a refueling outage (U2C13). The unit was
restarted on April 16 and achieved full power on April 22, 2005. Unit 2 operated at essentially
full power for the rest of the report period, expect for two significant downpowers. On June 12,
Operations executed a rapid power reduction of Unit 2 to 35% power due to the loss of the 2C
Main Transformer cooling system. Temporary repairs were effected, and the unit was returned
to full power on the same day. On June 18, Unit 2 power was reduced to 50% for two days to
effect permanent repairs to the 2C Main Transformer cooling system and main bus duct cooling
fan.
Unit 3 operated at or near full power for the entire report period expect for planned downpowers
of limited duration to exercise control rods and/or implement a sequence exchange.
3. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
Prior to and during the onset of hot weather conditions, the inspectors reviewed the
licensees implementation of 0-GOI-200-3, Hot Weather Inspection, including applicable
checklists - Attachment #1, Hot Weather Prep Annual Checklist; Attachment #2, Hot
Weather Operational Checklist; Attachment #3, Hot Weather Daily Log (Outside); and
Attachment #4, Hot Weather Daily Log (Inside). The inspectors also reviewed the Hot
Weather Discrepancy Log (PA-104); and discussed implementation of 0-GOI-200-3 with
responsible Operations personnel and management. Furthermore, the inspectors
conducted walkdowns of two potentially affected systems, the risk significant systems of
Residual Heat Removal Service Water (RHRSW) and Emergency Equipment Cooling
Water (EECW). The inspectors also reviewed and discussed (with Operations), the
procedural and TS requirements for operation and continued availability of the ultimate
heat sink (i.e., river).
b. Findings
No findings of significance were identified.
Enclosure
2
1R04 Equipment Alignment
.1 Partial Walkdown
a. Inspection Scope
Partial System Walkdown The inspectors performed partial walkdowns of the three
safety systems listed below to verify redundant or diverse train operability, as required
by the plant TSs. In some cases, the system was selected because it would have been
considered an unacceptable combination from a Probabilistic Safety Assessment (PSA)
perspective for the equipment to be inoperable while another train or system was out of
service. The inspectors verified that selected breaker, valve position, and support
equipment were in the correct position for system operation. Also, the walkdown was
done to identify any discrepancies that could impact the function of the system and lead
to increased risk. The inspectors observations of equipment and component alignment
for the partial walkdowns were compared to the licensee alignment procedures.
- Unit 2 Division II Residual Heat Removal (RHR) per operating instruction (OI) 2-
OI-74 checklist attachments and dwg. 2-47E811
- Unit 3 Core Spray Division I per 2-OI-75 checklist attachments and dwg. 3-
47E814
- Unit 3 High Pressure Coolant Injection (HPCI) per 3-OI-73 checklist attachments
and dwg. 3-47E812
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
Walkdowns. The inspectors reviewed licensee procedures, Standard Program and
Process (SPP)-10.10, Control of Transient Combustibles, and SPP-10.9, Control of Fire
Protection Impairments, and conducted a walkdown of the ten fire areas/zones and
impairments listed below. Selected fire areas/zones were examined in order to verify
licensee control of transient combustibles and ignition sources; the material condition of
fire protection equipment and fire barriers; operational lineup; and/or operational
condition of selected components. Also, the inspectors verified that selected fire
protection impairments were identified and controlled in accordance with procedure
SPP-10.9. In addition, the inspectors reviewed the Site Fire Hazards Analysis (FHA),
Volume 1 and 2 and applicable Pre-fire Plan drawings to verify that the necessary fire
fighting equipment, such as fire extinguishers, hose stations, ladders, and
communications equipment, were in place.
Enclosure
3
- Unit 1 Control Building, elevation 593' (Fire Area 16)
- Unit 1 Control Building, elevation 593' (Fire Area 17)
- Fire Pumps
- Compensatory Measures for Unit 1,2,3 Control Bay Impairments
- Compensatory Measures for Unit 2 Reactor Building Impairments
- Compensatory Measures for Unit 3 Reactor Building Impairments
- Compensatory Measures for Unit 1 Reactor Building Impairments (that
potentially affected Unit 2)
- Intake Pumping Station/RHRSW Pump Rooms (Fire Area 25)
- Control Building elevation 617' (Fire Area 16)
- Control Building elevation 606', including Cable Spreading Rooms (Fire Area 16)
b. Findings
No findings of significance were identified.
1R08 Inservice Inspection (ISI) Activities
a. Inspection Scope
The inspectors observed in-process ISI work activities on Unit 2 during the refueling
outage, reviewed ISI procedures, and reviewed selected ISI records associated with
risk-significant structures, systems, and components. The observations and records
were compared to the requirements specified in the TSs and the ASME Boiler and
Pressure Vessel Code, 1995 Edition 1996 Addenda, to verify compliance and to ensure
that examination results were appropriately evaluated and dispositioned.
The inspectors observed and reviewed non-destructive examination (NDE) activities.
Specific NDE activities were:
Direct Observation
Ultrasonic examination (UT):
Weld # RWC-2-001-G002, Stainless Steel Valve to Carbon Steel Elbow
Weld # RCRD-2-50, Stainless Steel Valve to Carbon Steel Elbow
Record Review
Ultrasonic Examination (UT):
High Pressure System, elbow to valve Weld # HPCI-2-004-009
Residual Heat Removal System Weld Attachment # 2-47B452S0239-1A
Radiographic Examination (RT):
High Pressure System, elbow to valve Weld # HPCI-2-004-009
Enclosure
4
Qualification and certification records for examiners, equipment and consumables, and
NDE procedures for the above ISI examination activities were also reviewed.
The inspectors also reviewed corrective action items such as Problem Evaluation
Reports (PERs) and Routine Work Orders (WOs), associated with the ISI program to
determine if problems were being identified at appropriate thresholds and if adequate
corrective actions were being taken.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
a. Inspection Scope
The inspectors observed operator crew performances during conduct of Simulator
Exercise Guide, 173S225, RHR Room Cooler Inoperable, HPCI Inadvertent Start With
Failure To Isolate, Fuel Failure, Loss Of Offsite Power and RCIC Controller Failure; and
Simulator Evaluation Guide OPL 177.060, RWCU Isolation, Loss of RPS Bus, MSL
Pressure Transmitter Failure, and ATWS with MSIV Closure, to verify that performance
was in accordance with licensee procedures and regulatory requirements.
The inspectors specifically evaluated the following attributes related to the operating
crews performance:
- Clarity and formality of communication
- Ability to take timely action to safely control the unit
- Prioritization, interpretation, and verification of alarms
- Correct use and implementation of Emergency Operating Instructions
- Timely and appropriate Emergency Action Level declarations per Emergency
Plan Implementing Procedures
- Control board operation and manipulation, including high-risk operator actions
- Oversight and direction provided by operations supervision, including ability to
identify and implement appropriate TS actions, regulatory reporting
requirements, and emergency plan actions and notifications
The inspectors attended the post-exercise critiques to assess their effectiveness, and
verify that the licensee-identified issues were comparable to issues identified by the
inspectors.
b. Findings
No findings of significance were identified
Enclosure
5
1R12 Maintenance Effectiveness
.1 Routine Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the two systems listed below with regard to some or all of the
following attributes: (1) work practices; (2) identifying and addressing common cause
failures; (3) Scoping in accordance with 10 CFR 50.65(b) of the maintenance rule (MR);
(4) characterizing reliability issues for performance; (5) trending key parameters for
condition monitoring; (6) charging unavailability for performance; (7) classification and
re-classification in accordance with 10 CFR 50.65(a)(1) or (a)(2); and (8)
appropriateness of performance criteria for Systems, Structures and Components
(SSCs)/functions classified as (a)(2) and/or appropriateness and adequacy of goals and
corrective actions for SSCs/functions classified as (a)(1). The inspectors also compared
the licensees performance against site procedure SPP-6.6, Maintenance Rule
Performance Indicator Monitoring, Trending and Reporting; Technical Instruction
0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting;
0-TI-362, Inservice Testing of Pumps and Valves; and SPP 3.1, Corrective Action
Program. The inspectors also reviewed applicable work orders, engineering evaluations
and system testing to verify that regulatory and procedural requirements were met.
- Unit 2 High Pressure Core Injection
- Unit 2 & 3 RHR Keep Fill Check Valves (Containment Isolation Valves)
b. Findings
Introduction: A Green non-cited violation (NCV) of 10 CFR 50.65(a)(1) was identified by
the inspectors for the licensees failure to implement timely and effective corrective
actions to preclude multiple, repetitive failures of containment isolation valves in the
Unit 2 and 3 Residual Heat Removal (RHR) Keep Fill System (e.g., 3-CKV-074-792,
2-CKV-074-802 and 2-CKV-074-803). These failures ultimately resulted in the failure of
two containment isolation valves simultaneously for the same penetration which created
an open pathway from containment and a consequential loss of the maintenance rule
safety function. Licensee monitoring and corrective actions per 10CFR50.65(a)(1) were
ineffective at ensuring containment isolation valves in the RHR Keep Fill System were
capable of performing their intended function.
Description: The inspectors reviewed PER, work order (WO) records, and Effectiveness
Reviews related to leakage of several RHR Keep Fill Check Valves which are also
Containment Isolation Valves tested in accordance with 10 CFR 50 Appendix J. The
inspectors noted that repetitive local leakrate test (LLRT) failures had occurred since
1998. The second of two consecutive failures of Unit 3 RHR Keep Fill check valve 3-
CKV-074-792 occurred in 1998 and was determined by the licensee to have exceeded
its performance criteria as described in 0-TI-346, Maintenance Rule Performance
Indicator Monitoring, Trending, and Reporting - 10 CFR 50.65. This valve was
Enclosure
6
subsequently placed in Maintenance Rule Category a(1). Analysis and corrective
actions were documented in PER 98-013383 (PER 36374). The licensee had
concluded that the check valves were being fouled by entrained sediment. The
corrective actions initially included cleaning of the Pressure Suppression Chamber
(PSC) head tank and flushing the check valve. WO 98-015733-000 was written to
document this action for Unit 3 and WO 99-001253-000 was written to document the
action on Unit 2. However, prior to the implementation of this corrective action, Unit 2
RHR Keep Fill check valves 2-CKV-074-802 and 2-CKV-074-803 failed their next
regularly scheduled LLRT and exceeded their performance criteria and were also placed
in Maintenance Rule Category a(1) in June of 1999. The existing corrective actions for
Unit 3 check valve 3-CKV-074-792 already planned were considered adequate and
applied to the Unit 2 check valves.
In August and September of 1999, the Unit 3 PSC head tank was cleaned and RHR
Keep Fill check valves 3-CKV-074-792 and 3-CKV-074-804 were flushed. In April of
2000, 3-CKV-074-792 passed its LLRT. However, 3-CKV-074-804 failed. The licensee
determined that the corrective action to flush and clean the PSC head tank and piping
was ineffective and decided not to clean and flush the Unit 2 RHR Keep Fill System.
Another corrective action was added to PER 98-013383 in November of 2000 to develop
a design change for each unit to install crud traps in the system.
In March of 2001, 2-CKV-074-802 and 2-CKV-074-803 passed their regularly scheduled
LLRT prior to any corrective action being done. In March of 2002, 3-CKV-074-792
passed its regularly scheduled LLRT with no corrective action other than the flush and
PSC head tank cleaning which had been deemed ineffective. Afterward, during the
same refueling outage, DCN 50830 was implemented to install crud traps in Unit 3.
Similarly, in February of 2003, during the Unit 2 refueling outage, check valves
2-CKV-074-802 and 2-CKV-074-803 again passed their regularly scheduled LLRT prior
to installation of the crud traps. In March 2003, DCN 50894 installed the crud traps in
Unit 2.
In April of 2003, the licensee determined that all corrective actions had been completed
and that there had been two consecutive successful LLRTs on each valve and hence
Maintenance Rule a(1) actions and monitoring were considered complete. All valves
were returned to Maintenance Rule Category a(2). In actuality, the successful LLRTs
occurred prior to implementation of the corrective actions in each case.
Subsequent to returning these valves to a(2) status, 3-CKV-074-792 failed its next
regularly scheduled LLRT by a large margin, along with two other Unit 3 check valves in
March of 2004. In September of 2004, 2-CKV-074-802 and 2-CKV-074-803,
simultaneously failed thus creating a pathway from the Suppression Pool to the PCS
head tank, which overflowed (PER 68246).
Enclosure
7
The licensee has since concluded that the completed corrective actions were ineffective
and that additional corrective actions are need to ensure reliability of the RHR Keep Fill
check valves and have placed all Keep Fill check valves for both RHR and Core Spray
on each unit (16 valves) into Maintenance Rule Category a (1) as of June 2005.
Analysis: The inspectors determined that the licensees failure to develop effective
corrective actions and to adequately monitor those corrective actions for effectiveness
as required by 10 CFR 50.65a(1) in order to prevent recurring failures of the Unit 2 and
Unit 3 RHR Keep Fill check valves was more than minor because if left uncorrected it
could become a more significant safety concern and it affected the Containment
Isolation SSC Reliability objective of the SSC and Barrier Performance attribute under
the Barrier Integrity Cornerstone. The inspectors assessed the finding using the SDP,
Manual Chapter 0609, Appendix H, Table 4.1, and determined the finding to be of very
low safety significance. The finding was of low safety significance because for the most
consequential failure event that resulted in an open pathway from containment, the
containment isolation valves were in a small (i.e., 2-inch) line to the PSC head tank and
would not significantly contribute to Large Early Release Frequency (LERF). This
finding also had cross-cutting aspects, as described above, associated with Problem
Identification and Resolution.
Enforcement: 10 CFR 50.65, Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants, Paragraph (a)(1), states, in part, that the licensee
...shall monitor the performance or condition of structures, systems, or components,
against licensee-established goals, in a manner sufficient to provide reasonable
assurance that such structures, systems, and components...are capable of fulfilling their
intended functions. Contrary to the above, the licensee goals and corrective actions
were untimely and ineffective in assuring the RHR Keep Fill check valves would be
capable of performing their containment isolation function. The failure to effectively
monitor and correct conditions adverse to quality of the RHR Keep Fill check valves
under Category a(1), is being treated as an NCV, consistent with Section VI.A.1 of the
NRC Enforcement Policy and is identified as NCV 05000260, 296/2005003-01,
Untimely and Ineffective Corrective Actions To Ensure RHR Keep Fill Containment
Isolation Valves Fulfill Their Safety Function Per 10 CFR 50.65 (a)(1). This issue is in
the licensees Corrective Action Program as PER 85130.
.2 Periodic Evaluation (Biennial)
a. Inspection Scope
The inspectors reviewed the licensees Maintenance Rule (MR) periodic assessment,
Maintenance Rule 4th Periodic Assessment Report, April 2002 to March 2004, dated
June 30, 2004, while on site during the week of April 18, 2005. The report was issued to
satisfy paragraph (a)(3) of 10 CFR 50.65, and covered the period as indicated for
Units 2 and 3. The inspection was to determine the effectiveness of the assessment
and that it was issued in accordance with the time requirement of the MR and included
evaluation of: balancing reliability and unavailability, (a)(1) activities, (a)(2) activities,
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8
and use of industry operating experience. To verify compliance with 10 CFR 50.65, the
inspectors reviewed selected MR activities completed during the assessment period for
the following maintenance rule systems: Reactor Fuels, Intermediate Range Monitoring,
Main Steam Relief Valves, Inverters, Drywell Head, and Repetitive Unit Scrams
(Unplanned Capacity Loss). In addition, the inspectors reviewed the most recent
structural inspection report and inspected select plant structures. Specific procedures
and documents reviewed are listed in the attachment to this report.
During the inspection, the inspectors reviewed selected plant WO data, assessments,
modifications, and the site guidance implementing procedure; discussed and reviewed
relevant PERs; reviewed generic operations event data, and probabilistic risk reports;
and discussed issues with system engineers. Operational event information was
evaluated by the inspectors in its use in MR functions. The inspectors selected WOs
and other corrective action documents on systems recently removed from
10 CFR 50.65 a(1) status and those in a(2) status to assess the justification for their
status. The documents were compared to the sites MR program criteria, and the MR
a(1) evaluations and rule-related data bases.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
k. Inspection Scope
For the five risk and emergent work assessments listed below, the inspectors reviewed
licensee actions taken to plan and control the work activities to effectively manage and
minimize risk. The inspectors verified that risk assessments were being performed as
required by 10 CFR 50.65(a)(4). The inspectors reviewed: licensee procedure
SPP-6.1, Work Order Process Initiation; SPP-7.1, Work Control Process; and 0-TI-367,
BFN Dual Unit Maintenance, to verify that procedure steps and required actions were
met. Also, the inspectors evaluated the adequacy of the licensees risk assessments
and the implementation of compensatory measures. The reviews completed included
the following:
- Risk Avoidance during 2A RHR 14-Day Extended TSAS
- Emergent Work Related to Leak on 2A RHR Hx
- Replacement of Main Control Board Switch 3-HS-85-48
- Unit 1 Refueling Zone Supply Damper (Secondary Containment Boundary)
b. Findings
No findings of significance were identified.
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9
1R15 Operability Evaluations
Routine Baseline Review
a. Inspection Scope
The inspectors reviewed the five operability/functional evaluations listed below to verify
technical adequacy and ensure that the licensee had adequately assessed TS
operability. The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR)
to verify that the system or component remained available to perform its intended
function. In addition, where applicable, the inspectors reviewed licensee procedure
SPP-3.1, Corrective Action Program, Appendix D, Guidelines For Degraded/Non-
conforming Condition Evaluation and Resolution of Degraded/Non-conforming
Conditions, to ensure that the licensees evaluation met procedure requirements.
Furthermore, where applicable, inspectors reviewed implemented compensatory
measures to verify they worked as stated and that the measures were adequately
controlled. The inspectors also reviewed PERs daily to verify that the licensee was
identifying and correcting any deficiencies associated with operability evaluations.
- Non-conservative Unit 2 Core Spray System Discharge Valve Torque Switch
Settings (PER 80624)
- Continued Unit 2 HPCI Operation with Unresolved Long-Term Equipment
Performance Issue with 2-LCV-73-8 (PER 80689)
- Compensatory Measures For Inoperable Unit 2/3 Offgas System Check Valves
as Part of Main Steam Ruggedness Boundary for Alternate Source Term
(PER 80690)
- Foreign Material Left In Unit 2 Drywell Following Closeout (PERs 80909 and
80856)
- Seismic, Structural Integrity, and Leak Tightness Evaluation of Unit 1 Refueling
Zone Supply Dampers and Temporary Blank with Minimum Bolting (Technical
Evaluation and Calculation # CDQ1-064-2005-0136)
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications
a. Annual Review
Inspection Scope
The inspectors reviewed Engineering Document Change (EDC) 64175 and Drawing
0-69D160-05 for seal weld activities on the 2A RHR Heat Exchanger floating head. As
part of this inspection, inspectors selectively reviewed licensee procedures 0-TI-405,
Plant Modifications and Design Change Control, and SPP-9.3, Plant Modifications and
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Engineering Change Control. The inspectors also reviewed the associated
10 CFR 50.59 screening against the system design bases documentation to verify that
the modifications had not affected system operability/availability. Furthermore, the
inspectors reviewed selected ongoing and completed work activities to verify that
installation was consistent with the design control documents.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing (PMT)
a. Inspection Scope
The inspectors witnessed and/or reviewed documentation of post-maintenance test
(PMT) activities of the seven risk significant SSCs listed below were adequate to verify
system operability and functional capability following completion of associated work. For
each of the PMTs, some or all of the following aspects were inspected: (1) Effect of
testing on the plant recognized and addressed by control room and/or engineering
personnel; (2) Testing consistent with maintenance performed; (3) Acceptance criteria
demonstrated operational readiness consistent with design and licensing basis
documents such as TS, UFSAR, and others; (4) Range, accuracy and calibration of test
equipment; (5) Step by step compliance with test procedures, and applicable
prerequisites satisfied; (6) Control of installed jumpers or lifted leads; (7) Removal of
test equipment; and, (8) Restoration of SSCs to operable status. The inspectors also
verified PMT activities were consistent with applicable procedural requirements of SPP-
6.3, Post-Maintenance Testing, and MMDP-1, Maintenance Management System.
Furthermore, the inspectors also reviewed problems associated with PMTs that were
identified and entered into the corrective action program.
- Unit 2 #1 Turbine Stop Valve Limit Switch (WO 05-712536-001 and 2-SR-
3.3.1.1.14(8II), Turbine Stop Valve Closure-RPS Trip (Channel B1/B2) Logic
System Functional Test)
- Unit 2 RCIC Pump (WO 03-009820 and 2-SR-3.5.3.3)
- Unit 3 #1 Turbine Stop Valve (WO 05-713332-000)
- Unit 3A Control Room A/C (WO 04-713508-001)
- Fire Pump A Head Replacement (WO 05-714630-000 and 0-SI-4.11.B.1.f(2),
Electric Fire Pump Capability Test)
- Unit 1 Refueling Zone Supply Dampers (Secondary Containment Boundary)
Maintenance of RHR Heat Exchangers)
b. Findings
No findings of significance were identified.
Enclosure
11
1R20 Refueling and Outage Activities
Unit 2 Cycle 13 (U2C13) Scheduled Refueling Outage
a. Inspection Scope
Ongoing Outage Activities
The inspectors continued to examine critical U2C13 outage activities to verify that they
were conducted in accordance with TSs, applicable procedures, and the licensees
outage risk assessment and management plans. Some of the more significant
inspection activities accomplished by the inspectors were as follows:
- Walked down of selected safety-related equipment clearance orders (i.e.,
2-TO-2005-0001, Sections 2-85-069 and 2-63-001)
- Verified operability of reactor coolant system (RCS) pressure, level, and
temperature instruments during various modes of operation
- Verified electrical systems availability and alignment
- Monitored important control room plant parameters
- Verified shutdown cooling, alternate decay heat removal, and spent fuel pool
cooling systems operation
- Evaluated implementation of reactivity controls
- Examined foreign material exclusion (FME) controls put in place around the
spent fuel pool and reactor cavity.
Refueling Activities and Containment Closeout
The inspectors witnessed selected fuel handling operations during the Unit 2 reactor
core fuel shuffle being performed on the refuel floor according to TS and applicable
operating procedures, such as 2-GOI-100-3B, Refueling Operations. The inspectors
also witnessed and examined the video verification of the final reactor core.
Furthermore, the inspectors performed detailed closeout inspections of the Unit 2
drywell and suppression chamber prior to plant startup and reviewed licensee
implementation of 2-GOI-200-2, Drywell Closeout.
Heatup, Mode Transition, Reactor Startup, and Power Ascension Activities
The inspectors examined selected TSs, license conditions, license commitments and
verified that administrative prerequisites were being met prior to Unit 2 mode changes.
The inspectors also reviewed measured RCS identified and unidentified leakage tests,
and verified that containment integrity was properly established. The results of low
power physics testing were discussed with Reactor Engineering and Operations
personnel to ensure that the core operating limit parameters were consistent with the
design. The inspectors witnessed portions of the reactor startup, heatup, and power
ascension in accordance with 2-GOI-100-1A, Unit Startup and Power Operation.
Enclosure
12
Correction Action Program
The inspectors reviewed PERs generated during U2C13 to verify that initiation
thresholds, priorities, mode holds, and significance levels were assigned as required.
Resolution and implementation of corrective actions of several PERs were also reviewed
for completeness.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors either witnessed portions of surveillance tests or reviewed test data for
the seven risk-significant SSCs listed below to verify that the tests met TS surveillance
requirements, UFSAR commitments, and in-service testing (IST) and licensee
procedure requirements. The inspectors review was to confirm that the testing
effectively demonstrated that the SSCs were operationally capable of performing their
intended safety functions and fulfilled the intent of the associated surveillance
requirement. Applicable IST data was compared against the requirements of licensee
procedures 0-TI-362, Inservice Testing of Pumps and Valves; 0-TI-230, Vibration
Monitoring and Diagnostics.
- Unit 2 Standby Liquid Control System Functional Test per 2-SR-3.1.7.7
- 1-2A EDG Monthly Surveillance Test per 0-SR-3.8.1.1(A)
- Unit 2 Turbine Control Valve Fast Closure Test per 2-SR-3.3.1.1.8(9)
- Unit 3 HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at
Rated Reactor Pressure per 3-SR-3.5.1.7 *
- Unit 2 RCIC Comprehensive Test per 2-SR-3.5.3.3
- Unit 3 Drywell Floor Drain Sump Flow Integrator Calibration per 3-SR-3.4.5.3 **
- Unit 3 Diesel Generator 24 Hour Run per 3-SR-3.8.1.7(3A)
- This procedure included inservice testing requirements.
- This procedure included testing of a containment isolation valve.
b. Findings
No findings of significance were identified
Enclosure
13
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed licensee procedures 0-TI-405, Plant Modifications and Design
Change Control; 0-TI-410, Design Change Control; SPP-9.5, Temporary Alterations;
and the temporary modification listed below to ensure that procedure and regulatory
requirements were met. The inspectors reviewed the associated 10 CFR 50.59
screening against the system design bases documentation to verify that the
modifications had not affected system operability/availability. The inspectors reviewed
selected completed work activities and walked down portions of the systems to verify
that installation was consistent with the modification documents and Temporary
Alteration Control Form (TACF).
- TACF 1-05-001-231, Temporary Recorder on Breaker 1-BKR-231-0001A/3D
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP1 Exercise Evaluation
a. Inspection Scope
Prior to the inspection activity, an in-office review was conducted of the exercise
objectives and scenario submitted to the NRC to determine if the exercise would test
major elements of the emergency plan as required by 10 CFR 50.47(b)(14).
The onsite inspection consisted of the following review and assessment of the licensees
biennial exercise conducted June 8, 2005:
- The adequacy of the licensees performance in the biennial exercise was
reviewed and assessed regarding the implementation of the risk-significant
planning standards (RSPS) in 10 CFR 50.47(b)(4), (5), (9), and (10), which are
emergency classification, offsite notification, radiological assessment, and
protective action recommendations, respectively.
- The overall adequacy of the licensees emergency response facilities with regard
to NUREG-0696, Functional Criteria for Emergency Response Facilities and
Emergency Plan commitments. The facilities assessed were the Control Room
simulator, Technical Support Center (TSC), Operations Support Center (OSC)
and Central Emergency Control Center.
- Other performance areas besides the RSPS, such as the emergency response
organizations (ERO) recognition of abnormal plant conditions, command and
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14
control, intra- and inter-facility communications, prioritization of mitigation
activities, utilization of repair and field monitoring teams, interface with offsite
agencies, and the overall implementation of the emergency plan and its
implementing procedures.
- Past performance issues from NRC inspection reports and Federal Emergency
Management Agency (FEMA) exercise reports to determine effectiveness of
corrective actions as demonstrated during this exercise to ensure compliance
with 10 CFR 50.47(b)(14).
- The post-exercise critique to evaluate the licensees self-assessment of its ERO
performance during the exercise and to ensure compliance with 10 CFR 50
Appendix E.IV.F.2.g.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
b. Findings
No findings of significance were identified.
1EP4 Emergency Action Level (EAL) and Emergency Plan Changes
a. Inspection Scope
The inspectors reviewed of revisions to the emergency plan, implementing procedures
and EAL changes to determine if changes had decreased the effectiveness of the plan.
The inspectors also evaluated the associated 10 CFR 50.54(q) reviews associated with
non-administrative emergency plan changes, implementing procedures changes, and
EAL changes. Revision 75 covered the period of July 27, 2004 to March 22, 2005.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 04, Emergency Action Level and Emergency Plan Changes. The
applicable planning standards, 10 CFR 50.47(b)(4), and its related 10 CFR 50
Appendix E requirements, were used as reference criteria. The criteria contained in
NUMARC/NESP-007, Methodology for Development of Emergency Action Levels,
Revision 2, and Regulatory Guide 1.101 were also used as references.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
b. Findings
No findings of significance were identified.
Enclosure
15
1EP6 Drill Evaluation
a. Inspection Scope
On April 27, 2005, the inspectors observed a quarterly emergency preparedness drill of
the licensees emergency response organization for personnel in the control room (i.e.,
simulator), TSC and the OSC. During this drill the inspectors assessed operator
performance to determine if emergency classification, notification, and protective action
recommendations were made in accordance with emergency preparedness procedures.
The inspectors evaluated the adequacy of the post-drill critiques conducted in the TSC
and the simulator.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control To Radiologically Significant Areas
a. Inspection Scope
Licensee activities for monitoring workers and controlling access to radiologically
significant areas were inspected. The inspectors evaluated procedural guidance and
directly observed implementation of administrative and physical controls; appraised
radiation worker and technician knowledge of, and proficiency in implementing, radiation
protection program activities; and assessed worker exposures to radiation and
radioactive material.
Radiological postings and material labeling were directly observed during tours of the
Unit 1 (U1), Unit 2 (U2), and Unit 3 (U3) turbine and reactor buildings and radwaste
processing areas. The inspectors conducted independent surveys in these areas to
verify posted radiation levels and to compare with current licensee survey records.
During plant tours, control of Locked High Radiation Area (LHRA) keys and the physical
status of LHRA doors were examined. In addition, the inspectors observed radiological
controls for non-fuel items stored in the spent fuel pools. The inspectors also reviewed
selected Radiological Control (Radcon) procedures and radiation work permits (RWPs),
and discussed current access control program implementation with Radcon supervisors.
During the inspection, radiological controls for work activities in High Radiation Areas
(HRAs) were observed and discussed. The inspectors attended a pre-job briefing for
work that involved entries into posted locked high radiation areas and directly observed
the work activities involved. The inspectors observed workers adherence to RWP
guidance and Health Physics Technician (HPT) proficiency in providing job coverage.
Controls for limiting exposure to airborne radioactive material were reviewed and
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operation of ventilation units and positioning of air samplers were also observed. The
inspectors evaluated electronic dosimeter alarm setpoints for consistency with
radiological conditions in and around the drywell. In addition, the inspectors interviewed
workers in the U1, U2, and U3 reactor buildings to assess knowledge of RWP
requirements.
The inspectors evaluated worker exposures through review of data associated with
discrete radioactive particle and dispersed skin contamination events. Controls used for
monitoring extremity dose and the placement of dosimetry when work involved
significant dose gradients were reviewed.
Radcon program activities were evaluated against 10 CFR Part 20; TS Sections 5.4,
Procedures, and 5.7, HRA; Regulatory Guide 8.38, Control of Access to High and Very
High Radiation Areas in Nuclear Power Plants; and approved licensee procedures.
Licensee guidance documents, records, and data reviewed are listed in the report
Attachment.
Problem Identification and Resolution: PERs and one audit associated with radiological
controls, personnel monitoring, and exposure assessments were reviewed and
discussed with Radcon supervisors. The inspectors assessed the licensees ability to
identify, characterize, prioritize, and resolve the identified issues in accordance with
licensee procedure SPP-3.1, Corrective Action Program, Revision 7. Specific
documents reviewed are listed in the report Attachment.
b. Findings
(1) Introduction: A Green self-revealing NCV of TS 5.7.2 was identified for the failure to
properly control a high radiation area with dose rates greater than 1.0 rem per hour at
30 centimeters (cm) from the source.
Description: On February 5, 2004, an operator entered the waste backwash transfer
pump room on the 546-foot elevation of the radwaste building and received an
electronic dosimeter dose rate alarm. The individual immediately exited the area and
reported to radiation protection. The operator received a dose of one millirem (mrem)
during the entry. Radiation protection performed a survey of the area and identified
dose rates of 10,000 mrem per hour on contact and 1500 mrem per hour at 30 cm on a
section of pipe. The area was immediately controlled as a locked high radiation area
and a guard was stationed to keep personnel out of the area.
The inspectors determined from a review of log entries and personnel statements that
on February 4, 2004, the Unit 2 spent fuel pool demineralizer was removed from service
to be backwashed and pre-coated. The backwash water and resin were pumped into
the waste backwash receiving tank. The operator would normally pump this tank to a
condensate waste phase separator as needed. On this day, the tank was pumped to
the B condensate waste phase separator at approximately 3:15 p.m. On February 5,
2004, at approximately 4:50 a.m., the operator reported receiving the dose rate alarm
when entering the waste backwash transfer pump room. At approximately 4:55 a.m.,
Enclosure
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the room was controlled as a locked high radiation area. Throughout the morning
several attempts were made to flush the source of the radioactivity. On February 6,
2004, at 5:00 a.m., radiation protection reported that dose rates in the area had
decreased to 70 mrem per hour on contact and 20 mrem per hour at 30 cm from the
original source location. Surveys of adjacent piping showed no increase in radiation
levels. From a review of previous surveys between May 2003 and the present, the
inspectors determined that dose rates in the waste backwash transfer pump room and
other radwaste building rooms have routinely fluctuated based on system operations,
although dose rates approaching 1000 mrem per hour at 30 cm had not been previously
identified. The inspectors concluded the licensee failed to implement timely radiation
surveys after plant evolutions which altered dose rates in the waste backwash transfer
pump room and other radwaste building rooms, such that any change in posting
requirement would be recognized. Following the February 5 event, the licensees
corrective action document stated that operations would revise procedures to require a
notification of radiation protection when the waste backwash transfer pumps were used
for a transfer.
Analysis: The failure to correctly control a high radiation area is a performance
deficiency in that the licensee failed to perform radiation surveys following a plant
evolution which changed the posting requirements to a high radiation area with dose
rates greater than 1.0 rem per hour at 30 cm from the source. The finding is greater
than minor because it is associated with the cornerstone attribute of exposure control
and adversely affected the cornerstone objective because the failure to control a high
radiation area does not ensure adequate protection of worker health and safety from
exposure to radiation. Because the finding involved the potential for workers to receive
significant, unplanned, unintended doses as a result of conditions contrary to TS
requirements, the inspector used the Occupational Radiation Safety Significance
Determination Process described in Manual Chapter 0609, Appendix C, to analyze the
significance of the finding. The inspectors determined that the finding was of very low
significance because: (1) it was not an As Low As Reasonably Achievable (ALARA)
finding, (2) it was not an overexposure, (3) it did have a substantial potential for
overexposure, and (4) it did not compromise the ability to assess dose.
Enforcement: TS 5.7.2 requires that each entryway to a High Radiation Area with
dose rates greater than 1.0 rem per hour at 30 cm from the radiation source, but less
than 500 rads per hour at 1 meter from the radiation source, shall be provided with a
locked door or gate or be continuously guarded to prevent unauthorized entry. The
licensee violated this requirement by failing to control the waste backwash transfer
pump room as a high radiation area with dose rates greater than 1.0 rem per hour at
30 cm from the source. The finding was documented in the licensees corrective
action program as PER 44288. Because this violation was of very low safety
significance and was entered into the licensees corrective action program, it is being
treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV
50-259, 260, and 296/200503-02, Failure to Control a High Radiation Area with Dose
Rates Greater than 1.0 Rem Per Hour.
Enclosure
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(2) Introduction: A Green self-revealing NCV of TS 5.7.1 was identified for the failure to
conspicuously post, barricade, and control access to high radiation areas with dose
rates not exceeding 1.0 rem per hour at 30 cm from the source. The licensee also
identified an additional example of a violation of TS 5.7.1 which is described in Section
4OA7 of the report.
Description: On November 24, 2004, a radwaste operator received an electronic
dosimeter dose rate alarm when entering the waste surge and collector pump room on
the 546-foot elevation of the radwaste building. The operator immediately exited the
area and reported to radiation protection. The operator entered an area with dose rates
of 159 mrem per hour and received a dose of 5 mrem from the entry. A survey of the
area showed elevated dose rates in several areas of the 546-foot elevation, including
the waste backwash transfer pump room. Contact dose rates with overhead piping
were as high as 2500 mrem per hour, with general area dose rates of 300 mrem per
hour. The licensee corrective action document stated that the possible cause of the
elevated radiation levels was the transfer of water from several recent spent fuel pool
cooling demineralizer backwashes. The operators failed to notify radiation protection
personnel of the operation of the waste backwash transfer pumps as required by
procedure 0-OI-77D, Backwash Receivers and Phase Separators System. This
contributed to radiation surveys not being performed following the plant evolution which
may have caused elevated dose rates and the change in the posting requirements. The
licensee posted and controlled the affected areas as high radiation areas. The affected
piping was flushed on November 26, 2004, and radiation levels returned to normal
levels.
Analysis: The failure to identify changes in dose rates following plant evolutions and
make corresponding changes to properly control a high radiation area is a performance
deficiency. The finding is greater than minor because it is associated with the
cornerstone attribute of exposure control and affected the cornerstone objective
because it resulted in unplanned or unintended radiation dose. Because the finding
involved the potential for workers to receive significant, unplanned, unintended doses as
a result of conditions contrary to TS requirements, the inspector used the Occupational
Radiation Safety Significance Determination Process described in Manual
Chapter 0609, Appendix C, to analyze the significance of the finding. The inspectors
determined that the finding was of very low significance because: (1) it was not an
ALARA finding, (2) it was not an overexposure, (3) it did have a substantial potential for
overexposure, and (4) it did not compromise the ability to assess dose.
In addition, this finding had cross-cutting aspects associated with human performance.
When operators failed to notify radiation protection personnel of the operation of the
waste backwash transfer pumps as required by Procedure 0-OI-77D, Backwash
Receivers and Phase Separators System, it directly contributed to the finding.
Enforcement: TS 5.7.1 requires that high radiation areas with dose rates not exceeding
1.0 rem per hour at 30 cm from the radiation source shall have each entryway
barricaded and conspicuously posted as a high radiation area. In addition, access to
such areas shall be controlled by a radiation work permit that includes specification of
Enclosure
19
radiation dose rates in the immediate area and that entry personnel are knowledgeable
of them. The licensee violated this requirement by failing to barricade and
conspicuously post an area that had dose rates of greater 100 mrem per hour. The
finding was documented in the licensees corrective action program as PER 72725.
Because this violation was of very low safety significance and was entered into the
licensees corrective action program, it is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy: NCV 50-259, 260, and 296/200503-03,
Failure to Barricade, Conspicuously Post, and Control a High Radiation Area.
(3) Introduction: Two examples of a Green self-revealing NCV of TS 5.4.1 were identified
for failure of workers to comply with radiation work permit requirements.
Description: On March 22, 2004, an operator entered a posted high radiation area on
an RWP that did not allow entry into high radiation areas. The operator followed a
chemistry technician into the B/D residual heat removal heat exchanger room and did
not notice the high radiation area posting. The operator entered the room on
RWP 04370011 which stated, in part, No entry into high radiation areas on this RWP.
The electronic dosimeter setpoints were 40 mrem for dose and 100 mrem per hour for
dose rate. The operator received a electronic dosimeter dose rate alarm. Radiation
dose rates in the area were 600 mrem per hour on contact and 300 mrem per hour at
30 cm from the radiation source.
On October 4, 2004, a craft worker entered an area in the overhead, greater than 6 feet,
of the Unit 1 reactor building 593-foot elevation without contacting radiation protection
as required by the RWP. The worker did not review the planned work with radiation
protection prior to entry and did not monitor his electronic dosimetry prior to reaching the
dose alarm setpoint. The worker received a dose alarm and a total dose of 11 mrem.
The electronic dosimeter dose alarm setpoint was 10 mrem. A survey of the overhead
area indicated dose rates of 200 mrem per hour on contact, 60 mrem per hour at 30 cm,
and 25 mrem per hour general area from overhead piping.
Analysis: The failure to follow RWP instructions is a performance deficiency. The
finding is greater than minor because it is associated with the cornerstone attribute of
program and process and it adversely affected the cornerstone objective because not
following RWP instructions does not ensure the adequate protection of the worker
health and safety from exposure to radiation. Because the finding involved workers
unplanned, unintended dose or potential for such a dose which could have been
significantly greater as a result of a single minor reasonable alteration of the
circumstances, the inspector used the Occupational Radiation Safety Significance
Determination Process described in Manual Chapter 0609, Appendix C, to analyze the
significance of the finding. The inspectors determined that the finding was of very low
safety significance because it did not involve: (1) ALARA planning and controls,
(2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired
ability to assess dose.
In addition, this finding had cross-cutting aspects associated with human performance.
When the worker in the first example failed to follow the requirements of his RWP with
Enclosure
20
respect to entry into high radiation areas, it directly contributed to the finding. When the
worker in the second example failed to follow the requirements of his RWP with respect
to entry into the overhead area requiring notification of radiation protection personnel
and receipt of a briefing of radiological conditions prior to entry, it directly contributed to
the finding.
Enforcement: TS 5.4.1 requires that written procedures shall be established,
implemented, and maintained covering access control to radiation areas including a
RWP system. Station Procedure RCI-9.1, Radiation Work Permit Preparation and
Administration, Revision 45, Section 7.4, states, in part, ...workers shall comply with all
RWP protective and special instructions. This requirement was violated when the first
worker failed to comply with the RWP requirement which prohibited entry into high
radiation areas. This requirement was also violated when the second worker failed to:
(1) review planned work with radiation protection prior to entry, (2) notify radiation
protection prior to working in the overhead greater than 6 feet, and (3) monitor electronic
dosimetry frequently and exit the area prior to exceeding an alarm setpoint. The finding
was documented in the licensees corrective action program as PERs 47669 and 69750.
Because this violation was of very low safety significance and was entered into the
licensees corrective action program, it is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy: NCV 50-259, 260, and 296/2005003-04,
Two Examples of Failure to Comply with Radiation Work Permit Requirements.
2OS2 ALARA Planning and Controls
a. Inspection Scope
ALARA: The inspectors evaluated ALARA program guidance and implementation for
on-going tasks associated with the Unit 2 Cycle 13 (U2C13) refueling outage. In
addition, post-outage ALARA activities associated with the Unit 3 Cycle 11 (U3C11)
refueling outage and ALARA planning and performance for recovery efforts on Unit 1
were evaluated. The inspectors reviewed and discussed with licensee staff ALARA
work plan documents, including dose estimates and prescribed ALARA controls for
selected outage work activities expected to incur significant collective doses. The
inspectors reviewed the implementation of dose-reduction initiatives for high person-rem
expenditure tasks. These elements of the ALARA program were evaluated for
consistency with the methods and practices delineated in applicable licensee
procedures.
The implementation and effectiveness of ALARA planning and program initiatives during
work in progress were evaluated. The inspectors made direct field observations of
Unit 2 work activities involving: preparations of radioactive shipments; maintenance of
the CRDMs; drywell cleanup and scaffolding removal; and refueling floor activities,
including the decontamination of the #2 I/S camera. The inspectors also observed work
activities in U1, including the painting of control rod drive accumulators in a
contaminated area. The inspectors interviewed radiation workers and RPT staff to
assess their understanding of dose reduction initiatives and their current and expected
final accumulated occupational doses at completion of the task.
Enclosure
21
Projected RWP dose expenditure estimates from U3C11 and Unit 1 recovery efforts
were compared to actual dose expenditures, and noted differences were discussed with
cognizant ALARA staff. Changes to dose budgets relative to changes in job scope were
identified and discussed. The inspectors attended pre-job briefings and evaluated the
communication of ALARA goals, RWP requirements, and industry lessons-learned to
job crew personnel.
Implementation and effectiveness of selected program initiatives with respect to
source-term reduction were evaluated. Chemistry program actions, including feedwater
iron control and depleted zinc oxide injection initiatives, and their resultant effect on
reactor coolant system (RCS) and equipment dose rate trending data were reviewed,
discussed, and compared to previous data. The implementation of routine flushing
procedures, hydrolazing, and de-sludging to reduce the source term were discussed
with cognizant licensee personnel. The effectiveness of selected shielding packages
installed for the current outage was assessed through completion of independent
radiation surveys and comparison to applicable licensee survey records and expected
planning data. Cobalt reduction initiatives for RCS valve replacement activities were
reviewed and discussed in detail.
The plant collective exposure histories for calendar years (CY) 2001, 2002 and 2003,
taken from data reported to the NRC pursuant to 10 CFR 20.2206), were reviewed and
discussed with licensee staff, as were established goals for reducing collective
exposure. The inspectors reviewed the applicable guidance and examined dose
records of declared pregnant workers during CY 2003 and 2004 to evaluate current
gestation doses for declared pregnant workers.
ALARA activities were evaluated against the requirements specified in 10 CFR 19.12;
10 CFR Part 20, Subparts B, C, F, G, H, and J; and approved licensee procedures. In
addition, licensee performance was evaluated against 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 2OS1 and 2OS2 of the report Attachment.
Problem Identification and Resolution: Licensee corrective action documents
associated with ALARA activities were reviewed and assessed. The inspectors
evaluated the licensees ability to identify, characterize, prioritize, and resolve the
identified issues in accordance with the corrective action program. Specific
self-assessments, audits, and PERs reviewed and evaluated in detail for this inspection
area are identified in Section 2OS2 of the report Attachment.
b. Findings
No findings of significance were identified.
Enclosure
22
Cornerstone: Public Radiation Safety [PS]
2PS2 Radioactive Material Processing and Transportation
a. Inspection Scope
Waste Processing and Characterization The inspectors reviewed the plants solid
radioactive waste system description in the Browns Ferry Nuclear Final Safety Analysis
Report (FSAR) and process control program (PCP). The most recent radiological
effluent release report was reviewed for information on the types and amounts of waste
disposed. The scope of the licensees audit program was reviewed to verify that it met
the requirements of 10 CFR 20.1101. The inspectors walked down the accessible
portions of the liquid and solid radioactive waste processing systems to verify and
assess that the current system configuration and operation agreed with the FSAR and
PCP. The liquid radioactive waste evaporator usage history and lay-up status was
discussed with radwaste and operations personnel to determine its potential to create an
unmonitored release pathway.
The inspectors reviewed the radiological operating report for any documented changes
to the radwaste processing systems and discussed the observations with radwaste and
operations department personnel. The inspectors reviewed the plants process for
transferring radioactive resin and sludge discharges into shipping/disposal containers to
determine if appropriate waste stream mixing and/or sampling procedures and
methodology for waste concentration averaging provided representative samples of the
waste product for waste classification purposes. The inspectors reviewed current
10 CFR 61 analysis results and the procedures for obtaining the samples to support the
analysis. The scaling factors used for radioactive waste streams and calculations used
for determining the amount of hard-to-detect nuclides were reviewed. The program was
reviewed to verify compliance with 10 CFR 61.55-56 and Appendix G of 10 CFR 20.
The inspector reviewed the program for provisions that would ensure that the waste
stream composition accounted for changes in operational parameters and would remain
valid between required periodic updates.
Transportation: The inspectors observed the preparation and shipment of contaminated
laundry to a vendor facility. The observations included packaging, surveying, labeling,
placarding, vehicle checks, drivers briefing and emergency instructions, a review of
shipping papers provided to the driver, and licensee final verification of shipment
readiness. The inspectors were unable to witness a Type B shipment because none
were scheduled during the inspection period. The inspectors reviewed shipping
documentation for several shipments that had occurred in the previous year. The
inspectors reviewed the Quality Assurance (QA) surveillance documentation verifying
compliance with the Certificate of Compliance for the Type B packages that included
spent resin and crushed irradiated control rod blades.
The inspectors observed and interviewed the radwaste workers who were involved in
the shipments and reviewed their training records.
Enclosure
23
Transportation program implementation was reviewed against regulations detailed in
10 CFR, Part 20; 10 CFR, Part 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. Documents reviewed during the inspection are listed in Section 2PS2 of the
report Attachment.
Problem Identification and Resolution Five PERs and one self-assessment were
reviewed in detail and discussed with licensee personnel. The inspectors assessed the
licensees ability to characterize, prioritize, and resolve the identified issues in
accordance with licensee procedure SPP-3.1, Corrective Action Program, Revision 7.
Documents reviewed for problem identification and resolution are listed in Section 2PS2
of the report Attachment.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
.1 Emergency Preparedness (EP)
a. Inspection Scope
The inspectors reviewed the licensees procedure for developing the data for the EP PIs
which are: (1) Drill and Exercise Performance (DEP); (2) Emergency Response
Organization (ERO) Drill Participation; and (3) Alert and Notification System (ANS)
Reliability. The inspectors examined data reported to the NRC for the period April 2004
to March 2005. Procedural guidance for reporting PI information and records used by
the licensee to identify potential PI occurrences were also reviewed. The inspectors
verified the accuracy of the PI for ERO drill and exercise performance through review of
a sample of drill and event records. The inspectors reviewed selected training records
to verify the accuracy of the PI for ERO drill participation for personnel assigned to key
positions in the ERO. The inspectors verified the accuracy of the PI for alert and
notification system reliability through review of a sample of the licensees records of
periodic system tests.
The inspection was conducted in accordance with NRC Inspection Procedure 71151,
Performance Indicator Verification. The applicable planning standard, 10 CFR 50.9
and NEI 99-02, Revision 3, Regulatory Assessment Performance Indicator Guidelines,
were used as reference criteria.
b. Findings
No findings of significance were identified.
Enclosure
24
.2 Radiation Protection
a. Inspection Scope
The inspectors sampled licensee records to verify the accuracy of reported PI data for
the periods listed below. To verify the accuracy of the reported PI elements, the
reviewed data were assessed against guidance contained in NEI 99-02, Revision 3, and
the PI Frequently Asked Questions List.
Occupational Radiation Safety Cornerstone: The inspectors reviewed the Occupational
Exposure Control Effectiveness PI results for the period of January 2004 through
March 2005. For the assessment period, the inspectors reviewed electronic dosimeter
alarm logs and licensee procedural guidance for collecting and documenting PI data.
Report Section 2OS1 contains additional details regarding the inspection of controls for
exposure-significant areas and the review of related PERs. Documents reviewed are
listed in Sections 2OS1 and 4OA1 of the report Attachment.
Public Radiation Safety Cornerstone: The inspectors reviewed the RETS/ODCM
Radiological Effluent Occurrences PI results for the period of January 2004 through
March 2005. For the assessment period, the inspectors reviewed cumulative and
projected doses to the public and two PERs related to RETS/ODCM issues. The
inspectors also reviewed licensee procedural guidance for collecting and documenting
PI data. Documents reviewed are listed in Section 4OA1 of the report Attachment.
b. Findings
No findings of significance were identified.
4OA2 Identification & Resolution of Problems
.1 Routine Review of Problem Evaluation Reports
a. Inspection Scope
The inspectors performed a daily screening of all PERs entered into the licensees
corrective action program. The inspectors followed NRC Inspection Procedure 71152,
Identification and Resolution of Problems, in order to help identify repetitive equipment
failures or specific human performance issues for follow-up.
b. Findings and Observations
There were no specific findings identified from this overall review of the PERs issued
each day.
Enclosure
25
.2 Semiannual Trend Review
a. Inspection Scope
As required by Inspection Procedure 71152, the inspectors performed a review of the
licensees corrective action program and associated documents to identify trends that
could indicate the existence of a more significant safety issue. The inspectors review
included the results from daily screening of individual PERs (see Section 4OA2.1
above), licensee quarterly trend reports and trending efforts, and independent searches
of the PER database. The inspectors review nominally considered the six-month period
of January 2005 through June 2005, although some PER database searches expanded
beyond these dates. The inspectors also interviewed the Human Performance
Manager, responsible corrective action program personnel, and key departmental
personnel regarding integrated trend analysis and assessment. Furthermore, the
inspectors verified whether adverse or negative trends and issues identified in the
licensees PERs, quarterly reports and trending efforts were entered into the corrective
action program (CAP).
b. Findings and Observations
No violations of NRC requirements were identified. However, the inspector identified
certain adverse trends related to corrective action extensions as well as inadequate
procedure use and adherence. The consistent, high number of PER action extensions
over the past five quarters was specifically discussed with the Human Performance
Manager. Although this trend was not previously recognized, the licensee now plans to
include it as a specific review item in the forthcoming CAP self-assessment. The
adverse trend related to inadequate procedure use and adherence was based on a
significant number of actual inspector-identified findings. These findings were of a
minor nature, each of which was discussed in detail with responsible licensee personnel
and management, and subsequently entered into the licensees CAP. Inadequate
procedure use and adherence had already been recognized by the licensee as a human
performance area in need of significant improvement and for which specific human
performance initiatives were being developed and instituted.
During review of the departmental quarterly trend analysis and summary reports, the
inspector identified that these reports lacked consistency and standardization among
departments, exhibiting a considerable variation in thoroughness and completeness.
Furthermore, little or no information was provided by the departments regarding the
effectiveness of prior actions and efforts to address previously recognized adverse
trends.
.3 Focused Annual Sample Review
Failure to identify that all safety-related valve functional requirements were incorporated
into the Motor Operated Valve Testing Program.
Enclosure
26
a. Inspection Scope
The inspectors reviewed PERs and corrective action documents and functional
evaluations related to functional testing for Motor Operated Core Spray Injection valves.
PERs 80624 and 97-001770 were reviewed in detail to ensure that the full extent of the
issue was identified, an appropriate evaluation was performed, and appropriate
corrective actions were specified, prioritized, and completed. The inspectors also
evaluated licensee actions against the requirements of the licensees corrective action
program as specified in SPP-3.1, Corrective Action Program, and 10 CFR 50,
Appendix B.
b. Findings and Observations
There were no findings of significance identified during this Problem Identification and
Resolution (PI&R) Annual Sample review. The inspectors determined that in 1998, as
part of a Testing Methodology review, the licensee recognized that valve settings and
testing requirements for some Motor-Operated Valves were non-conservative. Changes
were made to the program. However, four Core Spray valves, which are also
Containment Isolation valves, were not recognized as having a safety-related closure
function in a post-LOCA environment. Therefore, valve settings and testing
requirements for these valves were not changed. This resulted in in a lack of assurance
that these valves would fully close to their required leak tightness requirements in a
post-LOCA environment. These valves were the Core Spray Pump discharge valves for
Division I and II on Units 2 and 3 (2-FCV-75-25, 2-FCV-75-53, 3-FCV-75-25, 3-FCV-75-
53).
The licensee did subsequently determine that the torque switch settings for the Unit 3
valves were non-conservative and re-adjusted them to meet the more conservative
requirements in April of 2000. However, the licensee did not recognize that this
condition affected the same valves on Unit 2. Not until March of 2003, during testing of
the Unit 2 Division II injection valve, 2-FCV-75-53, was the valve torque switch setting
adjusted, but the need to adjust Unit 2 Division I, 2-FCV-75-25, valves torque switch
settings was again not recognized.
In April of 2005, during MOVATs testing of 2-FCV-75-25, Division I Core Spray
discharge valve, the licensee recognized that the valve would not meet the current
requirements. Due to refueling outage constraints, the re-setting of the torque switch
was deferred and a procedure change was made to secure the Division I Core Spray
pumps prior to closing 2-FCV-75-25. Closing this valve with the associated pump
secured is within the capability of the original torque setting.
The inspectors determined that the licensee failed to identify a condition adverse to
quality in 1998 which ultimately resulted in the Core Spray valves not meeting licensing
requirements for a time frame of 4 to 7 years. Furthermore, once recognized, the
licensee failed to adequately identify the extent of condition. This constituted a
performance deficiency and a finding. This finding is minor because the valves would
partially close to a small equivalent orifice size and leakage would be into a closed
Enclosure
27
system whose pressure retaining capability exceeds the post-LOCA conditions at which
time the valves would be required to close. This issue was determined to have cross-
cutting aspects associated with Problem Identification and Resolution
.4 Cross-Cutting Aspects of Findings
Section 1R12, describes an NRC-identified Maintenance Rule finding with cross-cutting
aspects related to PI&R due to untimely and ineffective corrective actions to ensure
primary containment isolation valves in the PSC keep fill were capable of performing
their intended safety function.
4OA3 Event Follow-up
.1 (Closed) Licensee Event Report (LER) 05000296/2005-001-00, Automatic Reactor Scram Due to False Main Transformer Differential Signal
At 1629 hours0.0189 days <br />0.453 hours <br />0.00269 weeks <br />6.198345e-4 months <br /> Central Standard Time on February 11, 2005, the Unit 3 reactor
scrammed from 100% power. The scram was caused by a simultaneous false trip
signal generated to the main generator circuit breaker 234, switchyard circuit breakers
5264 and 5268, and a main generator trip. This signal was generated when a PK block
(disconnect device 26W), which had been pulled as part of a clearance for breaker
5264, was re-inserted as part of a switching order from the Load Dispatcher for
returning the breaker to service. When the PK block 26W was inserted (out of
sequence of the switching order), the associated current transformer (CT) circuit was
momentarily grounded resulting in a false differential. The correct sequence of the
switching order was to actuate the trip cutout switches for the differential trip functions
prior to inserting any of the PK blocks. The generator trip resulted in a turbine trip and
opening of the output breakers causing a power-load unbalance trip. The control valve
(CV) fast closure caused the reactor to SCRAM. All rods inserted. Reactor water level
lowered, as expected, and was recovered by normal feedwater flow. All expected
Primary Containment Isolation System (PCIS) isolations were received along with the
auto start of Control Room Emergency Ventilation (CREV), and the three Standby Gas
Treatment (SGT) trains. The root cause of this event was determined to be personnel
error, in that the licensed operator failed to follow the task sequence identified in the
switching order. The inspectors reviewed the LER for completeness and accuracy and
to verify the licensee had developed appropriate corrective actions. For enforcement
and inspector followup activities, see NRC Inspection Report 05000260,296/2005002.
This LER is closed.
.2 (Closed) LER 05000260,296/2003-004-00, Cable Separations Design Error Related to
Appendix R Requirements.
Introduction: A violation of 10 CFR 50, Appendix R, III.G.1 and III.G.2 was identified for
failure to protect cables from fire damage which could result in the propagation of the
fire in a manner that could adversely affect safe shutdown. The NRC applied
enforcement and reactor oversight process discretion to the violation.
Enclosure
28
Description: In June of 2003, during a review and validation of Appendix R-related
calculations for restart of Unit 1, the licensee became aware of potential deficiencies
with the Unit 2 and 3 calculations regarding associated circuits for certain 4-kV electrical
distribution boards and loads. To evaluate and correct these deficiencies, PER 55116
was initiated. On July 7, 2003, the licensee determined that electrical cable routing
associated with the Unit 2 and 3 recirculation pump electrical boards was not in
compliance with Appendix R requirements. More specifically, the physical routing of
control power cables in the reactor building cable spreading room for the recirculation
pump power circuit breakers was in the proximity (i.e., within 20 feet) of the 4-kV power
supply cables feeding the very same circuit breakers. Consequently, certain single fires
could hypothetically cause electrical faults in the associated control power cables
resulting in fuse failures that would disable remote operation from the main control room
and automatic tripping of the 4-kV recirculation pump breakers. Whereupon, the same
fire could then damage the power supply cables causing high fault currents along the
entire length of these cables with no capability to automatically trip the power supply
breakers to isolate the fault. In the case of this unlikely scenario, the power supply
cable insulation could potentially ignite and spread anywhere along its cable run from
the reactor building to the turbine building, adversely affecting both trains of safe
shutdown equipment in other fire areas/zones.
Once this common vulnerability was identified, the licensee promptly implemented
compensatory measures (i.e., roving fire watches) until such time as a permanent fix
could be effected. To eliminate this vulnerability and restore compliance with Appendix
R, the licensee developed and implemented DCN 60035. This DCN modified the control
circuitry of the Unit 2 and 3 4-kV reactor recirculation pump boards to install coordinated
fuses that would isolate any fire-induced shorts in associated trip circuits which could
have prevented automatic tripping of the 4-kV recirculation pump feeder breakers.
Physical modifications of Unit 2 and 3 were accomplished in December 2003 and March
2004, respectively. The inspectors reviewed and verified that the corrective actions of
PER 55116 and this LER were completed. The inspectors also specifically reviewed
and verified completion of DCN 60035. Based upon the results of this inspection, the
licensees compensatory measures were considered prompt and their corrective actions
were timely.
Analysis The inspectors concluded that the licensees identification of inadequate
electrical cable routing associated with the Unit 2 and 3 recirculation pump electrical
boards constituted a performance deficiency because it violated requirements in the
area of fire protection, and was due to an error in Appendix R analysis that should have
been previously recognized by the licensee. In addition, the finding was associated with
the protection against external factors attribute and affected the objective of the
Mitigating Systems cornerstone to ensure the availability, reliability, and capability of
systems that respond to initiating events. Therefore, the finding is greater than minor.
According to NRC Inspection Manual Chapter 0305, Operating Reactor Assessment
Program, Section 06.06.2, Violations in Specified Areas of Interest Qualifying for
Enforcement Discretion, the NRC did not evaluate the significance of this finding
through use of the Significance Determination Process.
Enclosure
29
Enforcement 10 CFR 50, Appendix R, III.G.1, requires that fire protection features shall
be provided for components important to safe shutdown such that one train of systems
necessary to achieve and maintain hot shutdown conditions from the control room is
free of fire damage. Specific requirements for protection of cables are contained in
III.G.2. Contrary to these requirements, cables were not protected such that one train of
systems necessary to achieve and maintain hot shutdown conditions was not free of fire
damage or not independent from the area under consideration. As discussed in NRC
Regulatory Issue Summary 2004-003, Revision 1, and the NRCs Enforcement Policy,
the NRC may refrain from issuing enforcement action for violations associated with
fire-induced circuit failures provided the licensee has taken prompt compensatory
actions, and implements corrective actions within a reasonable time frame. For this
case, the licensee did not dispute that the requirements of Appendix R were violated;
and the conditions for applying enforcement discretion case were met. Consequently,
pursuant to NRC Enforcement Manual Section 8.1.7.1, Fire-Induced Circuit Failures, the
NRC has decided to exercise enforcement discretion. This LER is considered closed.
4OA4 Cross-Cutting Aspects of Findings
Section 2OS1 describes a self-revealing finding involving inadequate controls of high
radiation areas and a self-revealing finding (with two examples) involving adherence to
RWP requirements that have cross-cutting aspects associated with human
performance. In one finding, the failure of operators to notify radiation protection of the
operation of the waste backwash transfer pumps directly contributed to the licensee not
recognizing in a timely manner that the conditions for a high radiation area existed. In
the other finding, failure of the radiation workers to comply with RWP requirements
directly contributed to the intended radiological controls not being fully implemented.
4OA5 Other
.1 (Closed) TI 2515/163 - Operational Readiness of Offsite Power
a. Inspection Scope
The inspectors collected data pursuant to TI 2515/163, "Operational Readiness of
Offsite Power." The inspectors reviewed the licensee's procedures related to General
Design Criteria 17, "Electric Power Systems;" 10 CFR 50.63, "Loss of All Alternating
Current Power;" 10 CFR 50.65(a)(4), "Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants;" and the Technical Specifications for the offsite
power system. Documents reviewed for this TI are listed in the attachment.
b. Findings and Observations
No findings of significance or current operability issues were identified. In accordance
with TI 2515/163 reporting requirements, the inspectors provided the required data to
the headquarters staff for further review.
Enclosure
30
.2 Review of Institute of Nuclear Power Operations (INPO) Report
In May 2005, the inspectors reviewed the final report of the Browns Ferry INPO Plant
Evaluation (October 2004). This report did not identify any safety or risk significant
issues that had not been previously recognized and/or examined by the NRC.
.3 Visual Inspection of Plant Systems, Structures, and Components in Containment
a. Inspection Scope
The inspectors performed visual inspections of the interior of the Browns Ferry Unit 2
containment during Refueling Outage 13. This included observation of accessible
portions of plant systems, structures, components, instrumentation lines, and electrical
cables inside the containment to observe material condition and inspect for aging
conditions that might not have been previously recognized and addressed in the License
Renewal Application.
The observations of general material conditions included: inspection of piping
components for evidence of leaks or corrosion, inspection of coatings (piping, tanks,
and structural components), and inspection of electrical cables and instrumentation lines
for indications of deterioration. With the exception of some minor degradation of
coatings on the containment liner plate, the material condition at Browns Ferry was good
and no significant aging management issues were identified. The degraded coatings
are being identified and also have been entered into their program to be repaired.
b. Findings
The material condition at Browns Ferry was good and no significant aging management
issues were identified. No findings of significance were identified.
4OA6 Management Meetings
.1 Exit Meeting Summary
During the inspection period regional inspectors discussed the results their inspections
with the plant manager, and other responsible staff. A followup exit was held by
teleconference on June 2, 2005 with Mr. Kurt Krueger and other responsible staff to
discuss the final disposition of radiation protection findings. Then, on July 7, 2005, the
resident inspectors presented the integrated inspection results to the Plant Manager,
Mr. Bruce Aukland, and other members of his staff, who acknowledged the findings.
The inspectors confirmed that proprietary information was not provided or examined
during the inspection period.
.2 Annual Assessment Meeting Summary
During this inspection period, on May 24, 2005, the NRCs Chief of Reactor Projects
Branch 6 and the Senior Resident Inspector assigned to the Browns Ferry Nuclear Plant
Enclosure
31
met with the Tennessee Valley Authority (TVA) to discuss the NRCs Reactor Oversight
Process (ROP) and the Browns Ferry annual assessment of safety performance for the
period of January through December 2004. The major topics addressed were: the
NRCs assessment program, the results of the Browns Ferry assessment, and NRC
inspection plans. Attendees included Browns Ferry site management, members of site
staff, and corporate management.
This meeting was open to the public. The presentation material used for the discussion
is available from the NRCs document system (ADAMS) as accession number
ML052090076. ADAMS is accessible from the NRC Web site at
http://www/reading-rm/pdr.html (the Public Electronic Reading Room).
4OA7 Licensee-Identified Violations
The following finding of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which meets the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
- TS 5.7.1 requires that each high radiation area be barricaded and conspicuously
posted as a high radiation area and entry into such areas shall be made only
after entry personnel are knowledgeable of them. On August 9, 2004, three
individuals working in the Unit 1 drywell entered into a high radiation area without
being briefed on the radiological conditions. The highest dose rates within the
high radiation area were 150 mrem per hour on contact and 120 mrem per hour
at 30 cm from the source for a localized hotspot on a pipe. The general area
dose rate adjacent to the hotspot was 50 mrem per hour. The area was posted
and barricaded as a high radiation area; however, the barrier and posting had
been moved to an area approximately six feet above the floor that was not
conspicuous to the workers. As recorded by electronic dosimetry, the highest
radiation field actually entered by any of the workers was 39 mrem per hour.
The finding was of very low safety significance because it did not involve:
(1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential
for overexposure, or (4) an impaired ability to assess dose. The licensee
entered this finding into its corrective action program as PER 66755.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
SUPPLEMENTAL INFORMATION
PARTIAL LIST OF PERSONS CONTACTED
Licensee
B. Aukland, Nuclear Plant Manager
W. Crouch, Nuclear Site Licensing & Industry Affairs Manager
R. DeLong, Site Engineering Manager
A. Elms, Nuclear Plant Operations Manager
A. Feltman, Emergency Preparedness Supervisor
H. Hodges, ISI Program Coordinator
R. Jones, Unit 1 Restart Manager
J. Kennedy, Performance Improvement Manager
R. Kerwin, Acting Site Nuclear Assurance Manager
J. Lewis, Nuclear Plant Operations Manager
D. Logalbo, Unit 3 Outage Manager
B. Marks, Emergency Services - Corporate
R. Marks, Site Support Manager
R. Marsh, Operations Superintendent
M. Mitchell, Radiation Operations Manager
J. Mitchell, Site Security
D. Nye, Maintenance & Modifications Manager
C. Ottenfeld, Radiation Protection Manager
J. Parshall, Emergency Preparedness - Corporate
D. Pond, Emergency Preparedness - Corporate
M. Skaggs, Site Vice President
J. Sparks, Unit 2 Outage Manager
J. Steele, Daily Scheduling Manager
M. Welch, TVA Level III
K. Welch, Systems Engineering Manager
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000260, 296/200503-01 NCV Untimely and Ineffective Corrective Actions To
Ensure RHR Keep Fill Containment Isolation
Valves Fulfill Their Safety Function Per
10 CFR 50.65 (a)(1) (Section 1R12)
05000259, 260, 296/200503-02 NCV Failure to Control a High Radiation Area with Dose
Rates Greater than 1.0 Rem Per Hour
(Section 2OS1.1)
05000259, 260, 296/200503-03 NCV Failure to Barricade, Conspicuously Post, and
Control a High Radiation Area (Section 2OS1.2)
05000259, 260, 296/200503-04 NCV Two Examples of Failure to Comply with Radiation
Work Permit Requirements (Section 2OS1.3)
Attachment
2
Closed
05000296/2005-001-00 LER Automatic Reactor Scram Due to False Main
Transformer Differential Signal (Section 4OA3.1)
05000260, 296/2003-004-00 LER Cable Separations Design Error Related To
Appendix R Requirements (Section 4OA3.2)
05000259, 260, 296/2515/163 TI Operational Readiness of Offsite Power (Section
4OA5.1)
Discussed
None
LIST OF DOCUMENTS REVIEWED
Section 1R08: Inservice Inspection Activities
Procedures
NDE Procedure, N-UT-76, PDI Generic Procedure for the Ultrasonic Examination of Ferritic
Piping Welds, Revision 4
NDE Procedure N-VT-3, Visual Examination of Welds, Fit-Ups, and Dimensional Examination
of Weld Joints, Revision 24
NDE Procedure, N-PT-9, Liquid Penetrant Examination of ASME and ANSI Code Components
and Welds, Revision 27
Other Type Documents
Self Assessment Report # BFN-ENG-03-003 Browns Ferry Inservice Inspection Program
2nd Quarter Program/Component Status, Containment Inservice Inspection (IWE)
Section 1R12: Maintenance Effectiveness
MR - Corrective Action Program Documents
Problem Evaluation Reports (PER)
05-78722, B CREV Started Unexpectedly
05-78273, 0-SR-3.8.1.9 (A OL) DG A Load Acceptance Test
04-53996, Unit 3 Fuel Leak (CDE 2004-08-01 and 10 point plan)
01-46033, Fuse 3-FU1-256-1F Fuse Cleared (failed)
02-57988, C RHRSW Pump Room Penetrations
05-77985, MR Performance Criteria Exceeded
02-45690, Unit 2 Load Reject Trip
02-39166, Unit 2 Main Transformer Phase Bushing Failure Induced Trip
04-64906, Unit 2 Hi HI IPRM Level Trip
3
Site Working Procedures
0-SR-3.8.1.9 (A OL), DG A Load Acceptance Test, competed end of recent Unit 2 outage
O-OI-57C, 208V/120V AC Electrical System, Revision 77
Work Orders
05712889, 0-BKR -067-A/7A testing
05713773, Relay 2-PLT-099-0002BK4 is not Calibrated (TYPICAL)
Miscellaneous
System Related Event Log Unit 2 System 256 (ECCS Inverters)
TSAIL (Technical Specification LCO Logs) for Unit 2
System Status [Health] Reports for the last year for all a(1) systems
Expert Panel Meeting Minutes (Example: R40 041122 292, 4/17/05)
Calculation CDQO-303-2003-00260, 2002 Maintenance Rule Structures Inspection, 1/4/03
CDE 2004-07-04, Unit 2 Scram due to a Turbine Generator Load Reject (TYPICAL)
Section 1EP1: Exercise Evaluation
Plans and Procedures
Radiological Emergency Plan (Generic Part), Revision 75
Plan Effectiveness Determination (Generic Part), Revision 75
EPIL-1, Procedures, Maps, and Drawings, Revision 19
Section 1EP4: Emergency Action Level (EAL) and Emergency Plan Changes
Plans and Procedures
EPIP-1, Emergency Classification Procedure, Revision 38
EPIP-3, Alert, Revision 29
EPIP-4, Site Area Emergency, Revision 28
EPIP-5, General Emergency, Revision 33
EPIP-6, Activation and Operation of the TSC, Revision 23
Records and Data from 06/08/2005 exercise
Site Emergency Directors Journal (maintained by TSC Clerk)
Control Room Simulator Log
TSC Maintenance Managers Log
Team Tracking Form (from EPIP-7, Appendix A) as completed for all OSC teams
4
Section 2OS1: Access Controls to Radiologically Significant Areas
Problem Evaluation Reports
PER 41716, On 2/24/04 painters had to stop work due to airborne contamination, 2/25/04
PER 41752, Torus evacuated due to airborne alpha concentrations, 3/2/04
PER 41952, Operations failed to notified RP prior to placing HWC in service, 4/02/04
PER 44288, Operator received dose rate alarm in radwaste building, 2/6/04
PER 45297, Mechanic was contaminated during operation of RHR system, 3/8/04
PER 47083, A review of ED alarms from entering on wrong RWPs, 4/8/04
PER 47174, Draining equipment pit increased dose rates at handrails3/24/04
PER 47319, Laborer entered HRA on wrong RWP and received dose rate alarm, 3/23/04
PER 47474, Hot spot discovered on hose reading 1700/200 millirem per hour, 3/25/04
PER 47448, Two QC inspectors entered on wrong RWP and received dose rate alarm, 3/26/04
PER 47569, Maintenance worker entered HRA on wrong RWP, 3/16/04
PER 47669, Operator entered HRA on wrong RWP, 3/22/04
PER 47696, A 150 R/hr hot spot discovered during cleanout of the 3A hotwell, 3/21/04
PER 47708, 5 persons contaminated as the result of a spill, 3/10/04
PER 47772, Door to locked high radiation area found unsecured, 3/6/04
PER 47873, During removal of Control Rod Drives inconsistent dose rates recorded, 3/09/04
PER 48066, On the 23rd a Unit 1 laborer picked up 70 mrem picking up trash bag, 2/24/04
PER 48181, During investigation of dose rate alarm an individual entered on wrong RWP,
2/19/04
PER 60387, Operator received dose rate alarm, 5/1/04
PER 60421, Respiratory protection during Unit 3 equipment pit decontamination, 5/1/04
PER 62250, Individual in U2 on a U1 RWP and received dose rate alarm, 5/27/04
PER 63286, Two level 2 PCs from improper removal of contamination barrier, 6/15/04
PER 65860, Wrong scaffold location specified costs 60 mrem, 7/27/04
PER 66755, Unauthorized entry into a posted high radiation area, 8/9/04
PER 69750, Working on wrong RWP, 10/03/04
PER 72725, Emergent high radiation area in radwaste building, 11/24/04
PER 73009, Emergent high radiation area trend, 12/02/04
PER 74189, Laborers signed onto the wrong RWP, 12/28/04
PER 75141, Two workers working in overhead not briefed and in wrong unit, 1/19/05
Audits and Self-Assessments
BFN-RP-04-002, High Radiation Area Controls
BFN-RP-04-006, Radworker Practices/Radcon Practices
Radiation Work Permits
05290001 U2C13 Outage -Refuel Floor Activities
05280853 U2C13 Outage - Under-Vessel Work Activities
05282113 U2C13 Drywell Maintenance Carpenter Support
05282198 U2C13 Drywell Replace Reactor Head Vent Piping
05272072 U2C13 Reactor Building ISI
05272092 U2C13 Reactor Building IWE
05282003 U2C13 Drywell ISE
5
Procedures
SPP (Standard Programs and Processes)-3.1, Corrective Action Program, Revision 7
SPP-5.1, Radiological Controls, Revision 5
RCI (Radiological Control Instructions)-2.1, External Dosimetry Program Implementation,
Revision 52
RCI-9.1, Radiation Work Permit Preparation and Administration, Revision 45
RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Revision 48
RCI-23, Hot Spot Tracking Program, Revision 7
RCI-26, Radiation Protection Standards and Expectations, Revision 3
0-OI (Operating Instruction)-77D, Backwash Receivers and Phase Separators System,
Revision 24
0-OI-77D, Backwash Receivers and Phase Separators System, Revision 25
Miscellaneous Documents
BFNP Internal Dose Assessment Greater than 50 mrem CEDE, Intake Date June 15, 2004,
including whole body count, fecal analysis, and dose assessment
Spreadsheet: BFN U2C13 Investigation Counts (internal dose assessments for individuals
exceeding 20 mrem CEDE, dated April 7, 2005
Abacus-Plus Printouts: whole body count results for contaminated individual, initial estimate 46
mrem CEDE, intake date April 6, 2005
Report: BFN - Unit Two Cycle 13 Bioassay Measurements, dated March 31, 2005
Section 2OS2: ALARA Planning and Controls
Self-Assessments and Audits
BFN-RP-04-008, Focused Self-Assessment: Electronic Dosimeter Issues, 12/1-12/03
BFN-RP-04-006, Focused Self Assessment: Radworker Practices / Radcon Practices,
10/6-18/03
BFR-RIM-03-002, Unit 1 Restart ALARA Program Self-Assessment, 5/12-15/03
BFR-RRC-04-002, Cobalt Source Term Reduction, 9/20-24/04
BFR-RRC-04-001, Unit 1 Restart ALARA Program Self-Assessment, 8/16-19/04
Records, Reports and Data
Fiscal Year 2004 Annual ALARA Report (Units 2 and 3)
U3C11 Outage ALARA Report
Unit 1 Radiological Protection/ALARA Status Update, 3/21/05
Browns Ferry Nuclear Plant Site Dose Reduction Strategy, 9/22/04
Departmental Dose Reduction Plans (2004): Radiation Protection, Operations,
Chemistry/Environmental, Modifications, Outage and Scheduling, Training,
Maintenance, and Engineering
Table: List of Recurrent Work with ALARA Concerns and Actions Taken to Lessen Impact or
Prevent Recurrence
Graph: TVA-Browns Ferry Unit 2 Recirc Pipe Dose Rates, July 1991 - January 2005
Graph: TVA-Browns Ferry Unit 3 Recirc Pipe Dose Rates, November 1995 - Feburary 2005
Spreadsheet: Collective Exposure Trends.xls
Spreadsheet: Declared Pregnancies, 2004
Exposure Report for Declared Pregnant Female and HIS-20 Dosimetry History by Individual for
two declared pregnant females (declarations dated 9/27/04 and 11/10/04)
Table: BFN Annual ALARA Goals Dose Distribution by Section, 2002-2004
6
Chemistry CWPS Treatment Dose Results (graphs depicting Chemistry/Environmental Dose
Performance comparing 2005 actual and goal with 2004 actual)
As Low As Reasonably Achievable/Radwaste Committee Meeting Minutes: 6/23/04, 6/30/04,
7/7/04, 7/14/04, 7/3/04, 7/30/04, 8/6/04, 8/13/04, 8/20/04, 10/29/04, 12/15/04, 12/17/04,
1/7/05, 1/14/05, 1/18/05, 1/25/05, 2/1/05, 2/22/05
Survey No. 030105-3, Initial Contamination Survey of DW after the bellows flooding, 2/28/05
ALARA Planning Reports (APR)
APR 04-0041, U3C11 Outage - Refuel Floor Activities, Revision 2
APR 04-0062, U3C11 Outage - Encapsulate 3B and 3C RWCU Heat Exchangers, Revision 1
APR 05-0063, U2C13 Outage - Torus Desludge/Coating Inspection/Repair & ECCS Ring
Header Cleaning, Revision 0
APR 04-0041 Post Job Report
APR 04-0062 Post Job Report
Temporary Shielding Requests (TSR)
TSR 04-0019, U2C13 Outage Drywell, 1/14/05
TSR 04-0032, U1/U2 Reactor Building 664' Elev. Refuel Floor Spent Fuel Storage Pool
Transfer Canal, 12/6/04
TSR 05-0006, Unit 2 Drywell 550' Elev. Sub-Pile Room Floor Drain Grating Cover, 3/23/05
TSR 04-0041, U2C13 Reactor Cavity Inner and Outer Bellows, 3/21/05
TSR 02-0005, U3 Control Rod Drive System Drive Water Filters 3A and 3B
Radiation Work Permits (RWP)
RWP 05220742, Maintenance on RHR System (High Rad, Dose Control, Various Dress)
RWP 05222222, U2 RWCU HX Room Repair/Replace 2-FCV-069-0002 (High Rad, Dose
Control, Various Dress)
RWP 05260081, U2C13 Turbine U/S Condenser Modifications / Repairs (Various Dress)
RWP 05260087, U2C13 Turbine I/S Condenser Modifications / Repairs (High Rad/Tyvec
Hood/Various Dress)
RWP 05272322, U2C13, Rx Bldg, Disassemble/Repair Valve 2-74-57
RWP 05280002, U2C13, Drywell, Miscellaneous Maintenance
RWP 05280004, U2C13, Drywell, Miscellaneous Maintenance, Outside Support
RWP 05282032, U2C13 Drywell Snubber Maintenance (High Rad, Various Dress)
RWP 05282112, U2C13 Drywell Maintenance Carpenter Support (High Rad, Various Dress)
RWP 05282042, U2C13 Drywell MSRV Maintenance (High Rad, Various Dress)
RWP 05280912, U2C13 Drywell OPS Support Drywell/ Reactor Steam Tunnel (High Rad,
Various Dress)
RWP 05290025, U2C13 RFF Vessel Maintenance / Support (High Rad, Resp. and Various
Dress)
RWP 05290025, U2C13, Refuel Floor, Camera #2 I/S Decon
RWP 05333222, Replace/Repair RWCU Pump Mech. Seal (High Rad, Dose Control, Various
Dress)
RWP 05330692, Maintenance on RWCU System (High Rad, Dose Control, Various Dress)
RWP 05040613, Maintenance in F and D Valve Room (LHRA, Dose Control, Various Dress)
7
Problem Evaluation Reports (PER)
PER 79075, Demineralizer Contamination, 3/21/05
PER 69454, Potential contamination in clean RCA trend PER, 9/28/04
PER 78229, Poor radworker practices resulted in contamination of technician, PCE 20050028,
3/22/05
PER 74390, Wrong valve thermographed in LHRA, 1/4/05
PER 74443, C-zone not address in RadCon brief, 1/5/05
PER 75415, Employees picked up excessive dose because lead blankets not placed in correct
spot, 2/27/05
PER 49384, Actual dose accrued to complete WO#03-005760-000 and WO#03-015920-000
was 732 mrem vs. 300 mrem estimate, 10/7/03
PER 77623, U1 drywell spill during filling of reactor outer bellows with demin water, 3/1/05
PER 76843, U1 personnel concerned with expectations to stay in RCA for 7 1/2 hours per day,
2/16/05
PER 65400, Non-DOP tested vacuum use, 7/14/04
PER 68875, Need for empirical dose study to develop dose goals, 9/16/04
PER 68883, Lack of knowledge of Radiological Survey Maps, 9/16/04
Procedures
RCI-1.1, Field Operations Program Implementation, Revision 116
RCI-15.1, Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable,
Revision 31
RCI-15.2, Temporary Shielding, Revision 20
RCI-15.3, ALARA/Radwaste Committee, Revision 16
RCI-15.4, ALARA/Radwaste Volume Reduction Suggestion Program, Revision 6
RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Revision 48
RCI-24, Control of Vacuum Cleaners and Portable HEPA Units within the RCA, Revision 20
RCI-26, Radiation Protection Standards and Expectations, Revision 3
SPP-5.1, Radiological Controls, Revision 5
SPP-5.2, ALARA Program, Revision 2
SPP-6.1-3, TVAN Pre-Job Brief Checklist
RCDP (Radiological Control Department Procedures)-10, Personnel Contamination Reporting,
Revision 3
Section 2PS2: Radioactive Material Processing and Transportation
Procedures, Manuals, and Guides
SPP-5.0, Radiological and Chemistry Control, Revision 1
SPP-5.1, Radiological Controls, Revision 5
SPP-5.2, ALARA Program, Revision 2
SPP-5.6, Controlling Byproduct and Source Material, Revision 4
SPP-5.7, Radwaste Management, Revision 1
SPP-5.8, Special Nuclear Material Control, Revision 5
SPP-5.9, Radiological Control and Radioactive Material Shipment Augmented Quality
Assurance Program, Revision 6
RCDP-1, Conduct of Radiological Controls, Revision 2
RCDP-2, Conduct of Chemistry, Revision 1
RCI-1.1, Field Operations Program Implementation, Revision 116
RCI-7, Receipt of Radioactive Materials, Revision 16
8
RCI-15.1, Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable
(ALARA), Revision 31
RCI-15.3, ALARA/Radwaste Committee, Revision 16
RCI-26, Radiation Protection Standards and Expectations, Revision 3
RCI-27, Source Term Reduction and Control, Revision 2
RWTP (Rad Waste Technical Procedure) -100, Radioactive Material/ Waste Shipments,
Revision 2
RWTP -101, 10 CFR 61 Characterization, Revision 0
RWTP-102, Use of Casks, Revision 1
0-OI-77E, Solid Radwaste, Revision 33
Browns Ferry Nuclear Plant Process Control Program Manual (PCP), Revision 3
CAP Documents
Self Assessment Report BFN-RP-05-001, Radwaste Reduction (Focused), 12/13-15, 2004
PER 40218, Incoming rad material was not escorted by RP for 10 minutes while paperwork
issue was being resolved
PER 40971, Use of painted carbon steel is questioned for moisture separator/dryer strongbacks
vs Stainless Steel for ease of decon.
PER 48938, Laborer received facial contamination and minor uptake while cleaning radwaste
phase separators
PER 54155, Nuclear Assurance audit identified minor labeling problems in PASS area.
PER 75241, Rad waste shipment #050122 was above allowable weight by 4000 lb on one axle
when weighed by DOT weigh station. Discrepancy in weight of sealand container
43,560
(DOT) 37,200 (TVA crane load cell) over all shipment weight 76,600 vs DOT limit 80,000 lb.
Shipping Records and Radwaste Data
Shipment 050208, Cross around valves to Wylie, 2/9/05
Shipment 050126, RWCU/ CWPS Resin to Barnwell SC, 01/26/05
Shipment 040104, LPRMS to GE Reuter Stokes, 1/10/04
Shipment 040509, Cask containing irradiated hardware to Barnwell (Crushed control rod
blades), 5/27/04
Shipment 050408, Laundry to Unitech, 4/7/05 (Shipment observed)
Shipment 030716, CNS 14-170 cask of dewatered resin, 7/21/03
10 CFR 61 Analysis Report (Smears from all 3 units), 7/14/04
10 CFR 61 Analysis Report (CWPS Resin), 7/1/04
10 CFR 61 Analysis Report ( RWCU Resin) 7/6/04
Copies of training certificates for personnel involved in transportation of radioactive waste/
material
Section 4OA1.1: Performance Indicator Verification - Emergency Preparedness
Procedures
EPIL (Emergency Preparedness Instruction Letter) -15, Emergency Preparedness Performance
Indicators, Revision 9
Records and Data
9
Documentation (scenario, time line, event notification forms, player logs and completed
procedures) of ERO off-year exercise on 08/24/2004
Documentation of DEP opportunities from Operations Simulator evaluations on 04/21/2004,
04/26/2004, 06/21/2004, 12/07/2004
Selected training records of drill/exercise participation by ERO personnel during 2003-2004
Siren weekly and monthly tests 6/04 thru 3/05
Siren Maintenance records 6/04 thru 3/05
Section 4OA1.2: Performance Indicator Verification - Radiation Protection
Procedures
SPP-3.1, Corrective Action Program, Revision7
SPP-3.4, Performance Indicator and MOR Submittal Using INPO Consolidated Entry,
Revision 2
Documents/Records
Browns Ferry Nuclear Plant (BFN) - Units 1,2 and 3- Annual Radiological Environmental
Operating Report (AREOR) January Through December 2003 [ML041400033]
Browns Ferry, Units 1,2, and 3- Annual Radioactive Effluent Release (ARER) Report, January
through December 2003 [ML041240353]
Effluent Release Permit-Continuous Plant Main Stack March 2005
Effluent Release Permit-Batch Liquid Release December 2004
Preliminary Effluent Release Summary -January Through December 2004
Internal Dose Assessment including both in-vivo and in-vitro analysis results, dated 6/15/2004
Section 4OA5.1: TI 2515/163 - Operational Readiness of Offsite Power
Procedures
TVAN SPP-7.1, On-line Work Management
0-AOI-57-1A, Loss of Offsite Power (161 and 500 KV)/Station Blackout
0-GOI-300-1, Attachment 15.23, Emergency Load Curtailment - TVA Power System Alerts
OPDP-9, Emergent Issue Response
Documents
Memorandum dated June 17, 2004, Browns Ferry Nuclear Plant Grid Operating Guide For BFN
- West Point 500 KV Transmission Line And/Or BFN - Trinity 161KV Transmission Line
Outage(s)
PER 83217
BFN SENTINEL computer tool guidance