ML041350123
| ML041350123 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/14/2004 |
| From: | Kennedy K NRC/RGN-IV/DRP/RPB-C |
| To: | Ridenoure R Omaha Public Power District |
| References | |
| EA-04-078 IR-04-002 | |
| Download: ML041350123 (27) | |
See also: IR 05000285/2004002
Text
May 14, 2004
R. T. Ridenoure
Vice President
Omaha Public Power District
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550
Fort Calhoun, NE 68023-0550
SUBJECT: FORT CALHOUN STATION - NRC INTEGRATED INSPECTION REPORT AND
NOTICE OF VIOLATION 05000285/2004002
Dear Mr. Ridenoure:
On March 31, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Fort Calhoun Station. The enclosed integrated inspection report documents the
inspection findings which were discussed on April 5, 2004, with Mr. Ralph Phelps, Division
Manager, Nuclear Engineering, and other members of your staff.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, the NRC identified one violation which is cited in the
enclosed Notice of Violation (Notice) and the circumstances surrounding it are described in
detail in the subject inspection report. The violation is being cited because your staff failed to
restore compliance within a reasonable time after a violation was identified.
Additionally, the NRC identified two findings that were evaluated under the risk significance
determination process as having very low safety significance (Green). The NRC also
determined that there was a violation associated with one of these findings. This violation is
being treated as noncited violation (NCV), consistent with Section VI.A of the Enforcement
Policy. This NCV is described in the subject inspection report.
You are required to respond to this letter and should follow the instructions specified in the
enclosed Notice when preparing your response. The NRC will use your response, in part, to
determine whether further enforcement action is necessary to ensure compliance with
regulatory requirements.
Omaha Public Power District -2-
If you contest the violation or significance of the NCV, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with
copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV,
611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
NRC Resident Inspector at the Fort Calhoun Station facility.
In accordance with 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letter, and its
enclosures, 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).
Should you have any questions concerning this inspection, we will be pleased to discuss them
with you.
Sincerely,
/RA/
Kriss M. Kennedy, Chief
Project Branch C
Division of Reactor Projects
Docket: 50-285
License: DPR-40
Enclosures:
1. Notice of Violation
2. NRC Inspection Report 05000285/2004002
w/attachment: Supplemental Information
cc w/enclosures:
John B. Herman, Manager
Nuclear Licensing
Omaha Public Power District
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550
Fort Calhoun, NE 68023-0550
Richard P. Clemens, Division Manager
Nuclear Assessments
Fort Calhoun Station
P.O. Box 550
Fort Calhoun, NE 68023-0550
Omaha Public Power District -3-
David J. Bannister
Manager - Fort Calhoun Station
Omaha Public Power District
Fort Calhoun Station FC-1-1 Plant
P.O. Box 550
Fort Calhoun, NE 68023-0550
James R. Curtiss
Winston & Strawn
1400 L. Street, N.W.
Washington, DC 20005-3502
Chairman
Washington County Board of Supervisors
P.O. Box 466
Blair, NE 68008
Sue Semerena, Section Administrator
Nebraska Health and Human Services System
Division of Public Health Assurance
Consumer Services Section
301 Centennial Mall, South
P.O. Box 95007
Lincoln, NE 68509-5007
Daniel K. McGhee
Bureau of Radiological Health
Iowa Department of Public Health
401 SW 7th Street, Suite D
Des Moines, IA 50309
Chief Technological Services Branch
National Preparedness Division
Department of Homeland Security
Emergency Preparedness & Response Directorate
FEMA Region VII
2323 Grand Boulevard, Suite 900
Kansas City, MO 64108-2670
Omaha Public Power District -4-
Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
Senior Resident Inspector (JGK)
Branch Chief, DRP/C (KMK)
Senior Project Engineer, DRP/C (WCW)
Staff Chief, DRP/TSS (PHH)
RITS Coordinator (KEG)
Dan Merzke, Pilot Plant Program (DXM2)
RidsNrrDipmLipb
Rebecca Tadesse, OEDO RIV Coordinator (RXT)
FCS Site Secretary (NJC)
ANO Site Secretary (VLH)
W. A. Maier, RSLO (WAM)
G. F. Sanborn, D:ACES (GFS)
K. D. Smith, RC (KDS1)
F. J. Congel, OE (FJC)
OE:EA File (RidsOeMailCenter)
J. L. Dixon-Herrity, Senior Enforcement Specialist (JLD)
ADAMS: WYes G No Initials: __kmk___
W Publicly Available G Non-Publicly Available G Sensitive W Non-Sensitive
R:\_FCS\FC2004-02RP-JGK.wpd ML041350123
RI:DRP/C SRI:DRP/C C:DRS/OB C:DRS/EMB D:ACES
LMWilloughby JGKramer ATGody CSMarschall GFSanborn
T - KMKennedy E - KMKennedy E-KMKennedy Tapia for by E T - KMKennedy
5/12/04 5/6/04 5/11/04 5/12/04 5/11/04
D:DRP C:DRP/C
ATHowell KMKennedy
/RA/ /RA/
5/14/04 5/13/04
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NOTICE OF VIOLATION
Omaha Public Power District Docket 50-285
Fort Calhoun Station License DPR-40
During an NRC inspection conducted on January 1 through March 31, 2004, a violation of NRC
requirements was identified. In accordance with the General Statement of Policy and
Procedure for NRC Enforcement Actions, NUREG-1600, the violation is listed below:
10 CFR Part 50, Appendix B, Criterion V, states, in part, that procedures shall include
appropriate quantitative or qualitative acceptance criteria for determining that important
activities have been satisfactorily accomplished.
Contrary to the above, on January 21, 2004, the licensee failed to assure that
procedures included appropriate quantitative or qualitative acceptance criteria for
determining that important activities have been satisfactorily accomplished. Specifically,
the licensee failed to assure that Procedure OP-ST-DG-0001, Diesel Generator 1
Check, Revision 39, contained appropriate acceptance criteria for frequency when
performing a fast start of the diesel generator. The acceptance criteria did not account
for a 2 hertz speed droop of the fully loaded diesel generator when selecting the
minimum acceptable frequency. The licensee had previously received a noncited
violation (NCV 05000285/2003005-02) as a result of a similar condition.
This is a violation of very low safety significance (Green).
Pursuant to the provisions of 10 CFR 2.201, Omaha Public Power District is hereby required to
submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator,
U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington,
Texas 76011, and a copy to the NRC Resident Inspector at the facility that is the subject of this
Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This
reply should be clearly marked as a Reply to a Notice of Violation; EA-04-078 and should
include: (1) the reason for the violation or, if contested, the basis for disputing the violation or
severity level, (2) the corrective steps that have been taken and the results achieved, (3) the
corrective steps that will be taken to avoid further violations, and (4) the date when full
compliance will be achieved. Your response may reference or include previous docketed
correspondence, if the correspondence adequately addresses the required response. If an
adequate reply is not received within the time specified in this Notice, an order or a Demand for
Information may be issued as to why the license should not be modified, suspended, or
revoked, or why such other action as may be proper should not be taken. Where good cause is
shown, consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001.
Enclosure 1
-2-
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS), accessible from the
NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should
not include any personal privacy, proprietary, or safeguards information so that it can be made
available to the public without redaction. If personal privacy or proprietary information is
necessary to provide an acceptable response, then please provide a bracketed copy of your
response that identifies the information that should be protected and a redacted copy of your
response that deletes such information. If you request withholding of such material, you must
specifically identify the portions of your response that you seek to have withheld and provide in
detail the bases for your claim of withholding (e.g., explain why the disclosure of information will
create an unwarranted invasion of personal privacy or provide the information required by
10 CFR 2.390(b) to support a request for withholding confidential commercial or financial
information). If safeguards information is necessary to provide an acceptable response, please
provide the level of protection described in 10 CFR 73.21.
Dated this 14th day of May 2004
Enclosure 1
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-285
License: DPR-40
Report: 05000285/2004002
Licensee: Omaha Public Power District
Facility: Fort Calhoun Station
Location: Fort Calhoun Station FC-2-4 Adm.
P.O. Box 399, Hwy. 75 - North of Fort Calhoun
Fort Calhoun, Nebraska
Dates: January 1 through March 31, 2004
Inspectors: J. Kramer, Senior Resident Inspector
L. Willoughby, Resident Inspector
T. McKernon, Senior Operations Engineer
N. OKeefe, Senior Reactor Inspector
Approved By: Kriss M. Kennedy, Chief, Project Branch C
Division of Reactor Projects
Enclosure 2
SUMMARY OF FINDINGS
IR 05000285/2004002; 01/01/2004 - 03/31/2004; Fort Calhoun Station, Integrated Resident and
Regional Report; Surveillance Testing, Problem Identification and Resolution, Other
The report covered a 3-month period of inspection by Resident and Regional office inspectors.
One Green cited violation, one Green noncited violation, and one Green finding were identified.
The significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter 0609, Significance Determination Process. The NRC's program
for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. A violation of 10 CFR Part 50, Appendix B, Criterion V, was identified as
a result of the diesel generator test procedure not containing appropriate
quantitative or qualitative acceptance criteria to determine operability of diesel
generators when conducting the full speed starts of the diesel generators. The
licensees acceptance criteria did not account for a 2 hertz speed droop of the
fully loaded diesel generator when selecting the minimum acceptable frequency.
The licensee had previously received a noncited violation
(NCV 05000285/2003005-02) as a result of a similar condition adverse to quality.
This finding was considered more than minor because it was associated with the
procedure quality attribute of the mitigating systems cornerstone in that the
procedure did not contain appropriate quantitative acceptance criteria to ensure
the capability of the diesel generator to meet its design basis requirements. The
finding was characterized under the Significance Determination Process as
having very low safety significance because the as-found diesel generator
frequency and voltage were adequate to support the emergency core cooling
system loads and no actual loss of safety function occurred. This finding also
had crosscutting aspects associated with problem identification and resolution
because the licensee failed to correct a previously identified violation
(Section 4OA2.1).
- Green. A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, was
identified as a result of not properly translating design requirements into
procedures. Procedure AOP-17, Loss of Instrument Air, Revision 5, did not
provide adequate steps to respond to a prolonged loss of instrument air. Select
valves were equipped with air accumulators or backup nitrogen supplies to
maintain the valves operable after a loss of instrument air. The safety injection
refueling water tank recirculation valves have a 30-day design mission time
during a loss-of-coolant accident, but were provided with an accumulator capable
of lasting 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />. If the accumulator were to become expended, the valves
would fail open and divert water from containment recirculation to the safety
injection refueling water tank.
Enclosure 2
-2-
This finding was more than minor because it was related to the equipment
performance availability attribute of the mitigating systems cornerstone objective
and the design and configuration control attributes of the barrier integrity
cornerstone objective. The senior reactor analyst determined that the safety
injection refueling water tank recirculation valves would have remained closed
throughout the risk-significant portion of their mission time. Additionally, the
senior reactor analyst concluded that the likelihood of a loss-of-coolant accident
combined with a loss of instrument air was sufficiently small so that further
evaluation of the change in risk beyond the modeled mission time was not
required. Therefore, the failure to have an adequate abnormal operating
procedure for loss of instrument air represented a finding of very low risk
significance (Section 4OA5.1).
- Green. A finding was identified as a result of the licensee performing an
unauthorized modification to the coupling guard on the diesel-driven auxiliary
feedwater pump. Licensee personnel wrapped red duct tape around the guard
to reduce the excessive vibration due to broken welds on the guard. Since the
diesel-driven auxiliary feedwater pump is not safety related, the unauthorized
modification to the coupling guard was not a violation of requirements.
This finding was more than minor since it is associated with the equipment
performance reliability attribute of the cornerstone. The finding was
characterized as having very low safety significance because the pump
remained available to support unit operations. This finding also had crosscutting
aspects associated with human performance because personnel performed an
unauthorized temporary modification to a coupling guard (Section 1R22).
B. Licensee-Identified Violations
None
Enclosure 2
REPORT DETAILS
Summary of Plant Status
The unit began the inspection period at 100 percent power. On March 25, 2004, operators
commenced lowering power in preparation for a scheduled 10-day midcycle outage. The unit
was removed from service the following day. On March 28, the unit entered the first of two
planned midloop conditions for the replacement of reactor coolant pump seal packages. The
unit exited the midloop condition the following day. On March 31, the unit returned to a midloop
condition to complete the repairs to the reactor coolant pump seals. The unit was shut down in
Mode 4 at the end of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignments (71111.04)
a. Inspection Scope
The inspectors performed partial walkdowns (three inspection samples) of the following
trains of equipment during outages, operation, or testing of redundant trains. The
inspectors verified that the following systems were properly aligned in accordance with
system piping and instrumentation drawings and plant procedures:
- Low Pressure Safety Injection Pump SI-1A while Low Pressure Safety Injection
Pump SI-1B was inoperable for maintenance on January 15, 2004
- Diesel Generator 2 air start system while Diesel Generator 1 was inoperable for
maintenance on January 21, 2004
on March 3, 2004
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
a. Inspection Scope
The inspectors performed routine fire inspection tours (seven inspection samples) and
reviewed relevant records for plant areas important to reactor safety. The inspectors
observed the material condition of plant fire protection equipment, the control of
transient combustibles, and the operational status of barriers. The inspectors compared
inplant observations with commitments in the licensees Updated Fire Hazards Analysis
Report. The following fire areas were inspected:
Enclosure 2
-2-
- Fire Area 1 - Safety Injection and Containment Spray Pump Area (Room 21)
- Fire Area 2 - Safety Injection and Containment Spray Pump Area (Room 22)
- Fire Area 10 - Charging Pump Area (Room 6)
- Fire Area 20.1 - Personnel Air Lock Door Area (Room 58)
- Fire Area 23 - Pipe Penetration Area (Room 59)
- Fire Area 35A - Diesel Generator 1 (Room 63)
- Fire Area 42 - Main Control Cabinets in Control Room (Room 77)
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures (71111.06)
a. Inspection Scope
The inspectors performed a review of the flood protection measures (one inspection
sample). The inspectors reviewed the Probabilistic Risk Assessment Summary
Notebook for internal flooding events. The inspectors performed walkdowns of
Corridor 4 and Room 22 to verify that equipment was not subject to damage as a result
of internal flooding when the floor plug between Corridor 4 and Room 22 was removed.
The inspectors reviewed the internal flooding analysis that demonstrated that the
safety-related equipment in other rooms was not vulnerable to this internal flooding.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1 Quarterly Review of Requalification Activities
a. Inspection Scope
The inspectors performed a licensed operator requalification observation (one
inspection sample). On February 23, 2003, the inspectors observed licensed operator
requalification training activities, including the licensed operators performance and the
evaluators critique. The inspectors compared performance in the simulator with
performance observed in the control room during this inspection period. The focus of
the inspection was on high-risk licensed operator actions, operator activities associated
with the emergency operating procedures and the emergency plan, and previous
lessons-learned items. These items were evaluated to ensure that operator
performance was consistent with protection of the reactor core during postulated
accidents.
Enclosure 2
-3-
b. Findings
No findings of significance were identified.
.2 Biennial Review of Requalification Activities
a. Inspection Scope
The inspectors reviewed the annual operating examination test results for 2003. Since
this was the first half of the biennial requalification cycle, the licensee had not yet
administered the written examination. These results were assessed to determine if they
were consistent with NUREG 1021, Operator Licensing Examination Standards for
Power Reactors, Revision 8, Supplement 1, guidance and Manual Chapter 0609,
Appendix I, Operator Requalification Human Performance Significance Determination
Process, requirements. This review included examination of test results for a total of
51 licensed operators, which included shift-standing senior operators, staff senior
operators, shift-standing reactor operators, and staff reactor operators.
b. Findings
No findings of significance were identified.
1R12 Maintenance Rule Implementation (71111.12)
a. Inspection Scope
The inspectors reviewed the licensees implementation of the requirements of the
Maintenance Rule (10 CFR 50.65) and verified that the licensee conducted appropriate
evaluations of equipment functional failures, maintenance preventable functional
failures, the unplanned capacity loss factor, and system unavailability. The inspectors
discussed the evaluations with the licensee personnel. The following maintenance rule
items were reviewed (two inspection samples):
- Safety Injection Refueling Water Tank Recirculation Valve HCV-386 (Condition
Report 200400169)
- Condenser Evacuation In-Line Gas Radiation Monitor RM-057 (Condition
Report 200301279)
b. Findings
No findings of significance were identified.
Enclosure 2
-4-
1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)
a. Inspection Scope
The inspectors reviewed risk assessments for equipment outages (four inspection
samples) as a result of planned and emergent maintenance to evaluate the licensees
effectiveness in assessing risk for these activities. The inspectors compared the
licensees risk assessment and risk management activities against requirements of
10 CFR 50.65 (a)(4). The inspectors discussed the planned and emergent work
activities with planning and maintenance personnel. The inspectors verified that plant
personnel were aware of the appropriate licensee-established risk category, according
to the risk assessment results and licensee program procedures. The inspectors
reviewed the effectiveness of risk assessment and risk management for the following
activities:
- Outage of Diesel-Driven Auxiliary Feedwater Pump FW-54, High Pressure
Safety Injection Pump SI-2B, Charging Pump CH-1A, Air Compressor CA-1C,
and Condenser Evacuation Pump FW-8A on January 14, 2004
- Outage of Diesel Generator DG-1, Circulating Water Pump CW-1A,
Containment Cooling Coil Inlet Isolation Valve HCV-401B, Normal Range Stack
Gas Radiation Monitor Remote Ratemeter RM-062, and Accident Range Stack
Gas Radiation Monitor Remote Ratemeter RM-063 on January 21, 2004
- Outage of Boric Acid Pump CH-4A, Charging Pump CH-1A, and Circulating
Water Pump CW-1A on February 19, 2004
- Outage risk assessment and management for the unit shutdown and
maintenance outage that started on March 26, 2004
b. Findings
No findings of significance were identified.
1R14 Operator Performance During Nonroutine Evolutions and Events (71111.14)
1. Intake Blockage
a. Inspection Scope
On January 29, 2004, operators entered Procedure AOP-1, Acts of Nature,
Revision 14, as a result of high differential pressure across the intake trash grids caused
by the accumulation of debris and ice. Operators backwashed the trash grids to remove
the debris and also checked whether frazil ice conditions existed at the trash grids.
Enclosure 2
-5-
While backwashing the first trash grid, in accordance with Procedure OI-CW-1,
Circulating Water System Normal Operation, Revision 37, the operators throttled the
surface sluice isolation valve closed to allow a higher backwash pressure to develop.
Upon completion of the backwash of the first grid, the operators attempted to reopen the
surface sluice isolation valve. A pin in the reach rod assembly to the valve broke and
the surface sluice isolation valve remained in the throttled position. This caused
insufficient surface sluice flow and allowed surface ice to accumulate on the trash grids.
The operators opened the valve locally to establish surface sluice flow which kept the
surface ice from accumulating on the trash grids. The operators then proceeded to
backwash the remaining trash grids without throttling the surface sluice flow. After
backwashing the trash grids several times, the debris and surface ice were removed
from the grids and Procedure AOP-1 was exited.
The inspectors observed the backwashing of the trash grids and the evaluation of the
icing conditions. The inspectors discussed the event with operations crew involved with
the event. The inspectors reviewed Condition Report 200400345 that documented the
event (one inspection sample).
b. Findings
No findings of significance were identified.
2. Relay Failure and Loss of Letdown
a. Inspection Scope
On March 22, 2004, a relay in the reactor regulating system failed. The failure caused
the pressurizer level Channel Y setpoint to fail low, resulting in the isolation of the
letdown system. Operators allowed pressurizer level to rise to the upper end of the
control band and then stopped all charging flow into the reactor coolant system.
Operators diagnosed the problem and transferred the pressurizer level control to the
operable channel (Channel X) and re-established charging and letdown. On March 24,
the licensee replaced the failed relay and restored the charging and letdown system to
the normal alignment.
The inspectors discussed the relay failure, loss of letdown, and subsequent recovery
with the operations crew involved with the event. The inspectors performed a control
board walkdown following the event to verify that the plant was stable. The inspectors
reviewed the control room logs and Condition Report 200401111 that documented the
event (one inspection sample).
b. Findings
No findings of significance were identified.
Enclosure 2
-6-
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed operability evaluations (four inspection samples) to verify that
the evaluations provided adequate justification that the affected equipment could still
meet its Technical Specification, Updated Safety Analysis Report, and design bases
requirements. The inspectors also discussed the evaluations with cognizant licensee
personnel. The inspectors reviewed the operability evaluations and cause assessments
for the following:
- The adequacy of Restraint SISP-2 located below High Pressure Safety Injection
to Reactor Coolant Loop 1B Isolation Valve HCV-311 (Condition
Report 200401010)
- The operation of Safety Injection Cooler Outlet Control Valves PCV-2909,
PCV-2929, PCV-2949, and PCV-2969 during once through cooling (Condition
Report 200401064)
- Required flow from fire pumps to Diesel Generator 2 fire protection sprinkler
system (Condition Reports 2002025 and 200400452)
- Diesel Generator 1 start circuitry following fuse replacement (Condition
Reports 200400677 and 200400690)
b. Findings
No findings of significance were identified.
1R16 Operator Workarounds (71111.16)
a. Inspection Scope
The inspectors performed a review of the operator workarounds, the control room
deficiency, and control room burden lists. The inspectors focused on the cumulative
effects (one inspection sample) of the workarounds on the reliability and availability of
mitigating systems and the ability of operators to respond in a correct and timely manner
to plant transients and accidents. The inspectors reviewed Procedure OPD-4-17,
Control Room Deficiencies, Operator Burdens, and Operator Work Arounds,
Revision 11, that described the programs for handling workarounds and deficiencies.
In addition, the inspectors reviewed the licensees quarterly assessment of operator
workarounds dated February 5, 2004, and the planned corrective actions for the
deficiencies.
Enclosure 2
-7-
b. Findings
No findings of significance were identified.
1R19 Postmaintenance Tests (71111.19)
a. Inspection Scope
The inspectors observed and/or reviewed postmaintenance tests (four inspection
samples) to verify that the test procedures adequately demonstrated system operability.
The inspectors also verified that the tests were adequate for the scope of the
maintenance work performed and that the acceptance criteria were clear and consistent
with design and licensing basis documents. The following activities were included in the
scope of this inspection:
- Work Order 00159965-01, replace Diesel Generator 1 Turbo Oil Circulating
Pump LO-40-1, Work Order 00141420-02, replace Diesel Generator 1 governor
to control rod assembly on January 21, 2004
- Work Order 00108854, install new quick disconnect assemblies on Charging
Pump CH-1C on March 2, 2004
- Work Order 00152280-01, replace Diesel Generator Pump Motor LO-33-2-M on
March 4, 2004
- Work Order 00139297-01, disassemble, clean, visually inspect, and reassemble
Component Cooling Water Heat Exchanger AC-1A on March 15, 2004
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
On March 26, 2004, the licensee entered a planned 10-day outage to replace reactor
coolant pump seal packages. The inspectors reviewed the licensees outage shutdown
risk assessment to verify that the licensee appropriately considered risk in planning and
scheduling the outage activities. The inspectors observed the plant shutdown and
cooldown, the draining to midloop conditions, and shutdown maintenance activities. The
inspectors verified that the activities were performed in accordance with approved
procedures and Technical Specification requirements. Periodically, the inspectors
evaluated plant conditions to verify that safety systems were properly aligned and that
maintenance activities were controlled in accordance with the outage risk control plan.
The inspectors also performed containment tours and verified containment cleanliness.
Enclosure 2
-8-
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors observed and/or reviewed the performance and documentation for the
following surveillance tests (five inspection samples) to verify that the structures,
systems, and components were capable of performing their intended safety functions
and to assess operational readiness:
- OP-ST-CEA-0004, Secondary CEA Position Indication System Test,
Revision 15
- OP-ST-FO-3001, Diesel Generator 1 Fuel Oil System Pump Inservice Test,
Revision 18
- OP-ST-DG-0002, Diesel Generator 2 Check, Revision 39
- RE-ST-RX-0008, Shutdown Margin Verification During Hot Shutdown, Cold
Shutdown or Refueling, Revision 3
- OP-PM-AFW-0004, Third Auxiliary Feedwater Pump Operability Verification,
Revision 25
b. Findings
Introduction. A Green finding was identified as a result of licensee personnel performing
an unauthorized temporary modification to a coupling guard on an auxiliary feedwater
pump. Licensee personnel wrapped the coupling guard with red duct tape to reduce
vibration.
Description. During operation of Diesel-Driven Auxiliary Feedwater Pump FW-54,
licensee personnel observed excessive vibration of a coupling guard. The coupling
guard was located between the diesel engine and an electric generator mounted to the
front of the engine. The excessive vibration was due, in part, to several broken welds
on the guard. As a temporary fix to reduce the vibration, licensee personnel wrapped
the guard in red duct tape.
On January 11, 2004, the inspectors observed the coupling guard, with several cracked
support welds, wrapped with red duct tape. The inspectors informed the system
engineer and shift manager about the observation and questioned the reliability of the
pump. Specifically, the inspectors were concerned that the remaining coupling guard
welds would break during operation and the coupling would then fall onto the rotating
Enclosure 2
-9-
shaft and damage surrounding equipment needed to support pump operation. The
licensee acknowledged the inspectors concerns, removed the guard, and placed a
temporary barrier around the pump for personnel protection until a permanent repair
could be implemented.
Analysis. The inspectors evaluated the safety significance of the finding. The finding
affected the mitigating systems cornerstone and was considered more than minor
because the modification and degradation of the coupling guard affected the reliability of
the pump. The finding was evaluated using the significance determination process as
having a very low safety significance because the pump remained available to support
unit operations.
This finding had crosscutting aspects associated with human performance. The
unauthorized temporary modification to a coupling guard by licensee personnel directly
contributed to the finding.
Enforcement. The inspectors evaluated the enforcement aspects of the finding. The
finding was associated with a high risk-significant component; however, the
diesel-driven auxiliary feedwater pump is not safety-related. Therefore, the
unauthorized modification to the coupling guard was not a violation of requirements.
This finding (FIN 05000285/2004002-01) was entered into the licensees corrective
action program as Condition Report 200400156.
1R23 Temporary Plant Modifications (71111.23)
a. Inspection Scope
The inspectors reviewed Temporary Modification EC 34182 (one inspection sample) that
installed a temporary flange in place of the reactor coolant pump mechanical seal while
the mechanical seal was removed for refurbishment. In addition, the inspectors
reviewed the 10 CFR 50.59 screening associated with the modification. The inspectors
attended the plant review committee meeting that approved the temporary modification.
The inspectors performed a walkdown of the modification and verified that the
modification had no adverse impact on the safety function of the system.
b. Findings
No findings of significance were identified.
Enclosure 2
-10-
Cornerstones: Emergency Preparedness
1EP6 Drill Observation (71114.06)
a. Inspection Scope
On January 20, 2004, the inspectors observed aspects of the emergency preparedness
drill from the simulator and the technical support center (one inspection sample). The
purpose of the observations was to evaluate operator performance, licensee event
classification, notification of state and local authorities, and the adequacy of protective
action recommendations. The inspectors reviewed the licensees postdrill critiques and
discussed observations with licensee management.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
a. Inspection Scope
The inspectors reviewed the licensees performance indicator data to verify its accuracy
and completeness for the following three indicators:
- IE1 Unplanned Scrams
- IE2 Scrams With a Loss of Normal Heat Removal
- IE3 Unplanned Power Changes
The inspectors reviewed the performance indicator data for the 4 quarters of 2003. The
inspectors reviewed NEI 99-02, Regulatory Assessment Performance Indicator
Guideline, and licensee operating logs. The inspectors discussed the status of the
performance indicators and compilation of data with licensee personnel.
b. Findings
No findings of significance were identified.
Enclosure 2
-11-
4OA2 Problem Identification and Resolution (71152)
.1 Diesel Generator Surveillance Testing Procedure
a. Inspection Scope
On January 21, 2004, the inspectors observed operators perform a full speed start of
Diesel Generator 1 using Procedure OP-ST-DG-0001, Diesel Generator 1 Check,
Revision 39. The inspectors reviewed the procedure, test data, acceptance criteria, and
Technical Specifications. The inspectors evaluated the corrective actions associated
with Condition Reports 200302602 and 200302623. The inspectors discussed the
diesel generator surveillance testing procedure and the condition report corrective
actions with the licensee.
b. Findings
Introduction. A Green violation was identified as a result of the diesel generator test
procedure not containing appropriate quantitative or qualitative acceptance criteria to
determine operability of the diesel generator as required by 10 CFR Part 50,
Appendix B, Criterion V.
Description. NRC Inspection Report 05000285/2003005 documented a Green noncited
violation as a result of the diesel generator test procedure not containing appropriate
quantitative or qualitative acceptance criteria to determine operability of the diesel
generators as required by 10 CFR Part 50, Appendix B, Criterion V. Specifically, for the
conduct of the full speed start of Diesel Generator 2 performed on July 7, 2003, to
demonstrate that the diesel generator could start and accelerate to rated speed and
voltage in less than or equal to 10 seconds without prior warm up, the acceptable
frequency band listed in Procedure OP-ST-DG-0002, Diesel Generator 2 Check,
Revision 38, was between 57 and 63 hertz. The inspectors noted that the diesel
generator had a 2 hertz frequency droop when going from unloaded to fully loaded
during accident conditions based on the governor design. If the diesel generator came
up to speed at the minimum acceptable frequency of 57 hertz, as stated in the
surveillance procedure, and emergency core cooling systems loads were placed on the
diesel generator, the frequency seen by the emergency core cooling system motors
would be approximately 55 hertz. This frequency would be outside the ANSI/NEMA
MG 1-1998 criteria (57 hertz) for motor operation. The inspectors had previously asked
the licensee if all the emergency core cooling system loads would function properly
when the diesel generator was running loaded at 55 hertz, the minimum acceptable
frequency allowed in the surveillance procedure minus the 2 hertz speed droop. The
licensee had indicated that there was no calculation or design basis information to
support equipment operability at 55 hertz. The inspectors determined that
Procedure OP-ST-DG-0002 did not contain appropriate quantitative or qualitative
acceptance criteria to determine operability of the diesel generator. The licensee
entered this violation in their corrective action program as Condition Reports 200302602
and 200302623.
Enclosure 2
-12-
On January 21, 2004, the licensee performed a full speed start of Diesel Generator 1
using Procedure OP-ST-DG-0001, Diesel Generator 1 Check, Revision 39, to
demonstrate that the diesel generator could start and accelerate to rated speed and
voltage in less than or equal to 10 seconds without prior warm up. The inspectors found
that the procedure contained the same acceptance criteria for frequency, 57 to 63 hertz,
as was found in Procedure OP-ST-DG-0002 during the July 2003 test of Diesel
Generator 2. Procedures OP-ST-DG-0001 and OP-ST-DG-0002 are identical
procedures, except the -0001 procedure applies to Diesel Generator 1, and the -0002
procedure applies to Diesel Generator 2. The inspectors found that neither procedure
had been corrected by the licensee following the identification of the noncited violation
documented in NRC Inspection Report 05000285/2003005 and that the licensee had
failed to assure that a condition adverse to quality was corrected.
Analysis. The inspectors evaluated the safety significance of the finding. This finding
affected the Mitigating Systems cornerstone and was considered more than minor
because the procedure did not contain appropriate quantitative acceptance criteria to
ensure the capability of the diesel generator to meet its design basis requirements. The
finding was characterized under the Significance Determination Process as having very
low safety significance because the as-found diesel generator frequency and voltage
were adequate to support the emergency core cooling system loads and no actual loss
of safety function occurred.
This finding had crosscutting aspects associated with problem identification and
resolution. The failure of the licensee to correct the procedure when previously
receiving a noncited violation for a similar condition directly contributed to the finding.
Enforcement. 10 CFR Part 50, Appendix B, Criterion V, states, in part, that procedures
shall include appropriate quantitative or qualitative acceptance criteria for determining
that important activities have been satisfactorily accomplished. Contrary to the above,
on January 21, 2004, the licensee failed to assure that procedures include appropriate
quantitative or qualitative acceptance criteria for determining that important activities
have been satisfactorily accomplished. Specifically, the licensee failed to assure that
Procedure OP-ST-DG-0001, Diesel Generator 1 Check, Revision 39, contained
appropriate acceptance criteria for frequency when performing a fast start of the diesel
generator. The licensee had previously received a noncited violation
(NCV 05000285/2003005-02) as a result of a similar condition adverse to quality
identified on July 7, 2003. This violation of 10 CFR Part 50, Appendix B, Criterion V, is
being treated as a violation, consistent with the Enforcement Policy
(VIO 05000285/2004002-02). This violation is in the licensees corrective action
program as Condition Report 200400517.
.2 Reference to Crosscutting Findings Documented Elsewhere in the Report
Section 1R22 describes the unauthorized temporary modification licensee personnel
performed on Diesel-Driven Auxiliary Feedwater Pump FW-54. Licensee personnel
wrapped red duct tape around a vibrating coupling guard with broken support welds.
Enclosure 2
-13-
This finding had crosscutting aspects associated with human performance. The
unauthorized temporary modification to a coupling guard by licensee personnel directly
contributed to the finding.
4OA3 Event Followup (71153)
(Closed) Licensee Event Report 05000285/2002004-00: Inadequate procedural
guidance resulting in noncompliance with 10 CFR Part 50, Appendix R.
During the performance of an NRC triennial fire protection inspection, the team
identified a noncited violation 10 CFR Part 50, Appendix R, Section III.G.2, as
committed to in License Condition D of the Fort Calhoun Station operating license.
Specifically, the noncited violation was for the failure to assure that one train of systems
necessary to achieve and maintain hot shutdown conditions from either the control room
or emergency control stations is free of fire damage, as required by 10 CFR Part 50,
Appendix R,Section III.G.2. The team identified that a fire in either Fire Area 6 or Fire
Area 36A could result in the loss of all three trains of charging pumps. The licensee
credits these pumps for accomplishing the hot shutdown function of reactor coolant
inventory control. The team documented this finding in NRC Inspection
Report 05000285/2003002, Section 1R05.2. The licensee event report was reviewed by
the inspectors and no new findings were identified. The licensee documented the issue
in Condition Report 200204129. This licensee event report is closed.
4OA5 Other
.1 (Closed) Unresolved Item 05000458/2003011-03: Failure to Provide Means to Assure
Proper Emergency Core Cooling System Alignment During Prolonged Loss of
Instrument Air
Introduction. A Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III,
was identified. Specifically, the procedure for the loss of instrument air did not contain
sufficient information for addressing a loss of instrument air over an extended period.
Description. NRC Inspection Report 05000458/2003011 documented that
Procedure AOP-17, Loss of Instrument Air, Revision 5, did not contain actions to be
taken in the event of a prolonged period without an instrument air source. The plant
safety analyses stated that the plant did not need the bulk of the instrument air system
to function during design basis events. The safety-related instrument air loads were
provided with accumulators or backup nitrogen bottles which were capable of providing
pressure for a specified duration in order to maintain the functions required for safe
operation.
However, in the case of the Safety Injection Refueling Water Tank Recirculation
Valves HCV-385 and HCV-386, accumulators were provided that were designed to hold
the valves closed for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, after which time the valves would fail open. (Test data
from 1998 to present indicated the minimum accumulator bleed-down time was
Enclosure 2
-14-
39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />.) These valves were in series in the recirculation line that was common to all
trains of emergency core cooling systems, allowing water to return to the safety injection
refueling water tank. At the beginning of a loss-of-coolant accident, these valves would
be in the open position. The valves close when the injection phase is complete and the
emergency core cooling systems switch to containment sump recirculation. If these
valves lost operating air pressure and failed open during sump recirculation, the borated
water used to keep the reactor shut down and to cool the core would be diverted to the
safety injection refueling water tank. This would ultimately cause a loss of the ability to
run the containment spray pumps and then the safety injection pumps as water level
lowered in the sumps.
These valves also served as a containment boundary during the recirculation phase.
The increased leakage from the containment to the safety injection refueling water tank
would be greater than the limit used in the analysis of Updated Safety Analysis Report,
Section 14.15.8, Radiological Consequences of a LOCA; therefore, the resulting dose
assessment would also be increased.
Analysis. In accordance with Inspection Manual Chapter 0612, this finding was more
than minor because it was related to the equipment performance availability attribute of
the mitigating systems cornerstone. It affected the cornerstone objective in that the
performance deficiency affected the reliability of the emergency core cooling system to
respond continuously to the design basis 30-day accident. This finding also affected the
design and configuration control attributes of the barrier integrity cornerstone in that the
performance deficiency affected the assurance of the containment barrier to be able to
protect the public from radionuclide releases caused by accidents.
In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Attachment 1,
User Guidance for Significance Determination of Reactor Inspection Findings for
At-Power Situations, the inspectors determined that a Phase 2 significance
determination was required because two cornerstones were affected. However,
because two coincident initiating events were required for this finding to affect the plant
mitigating capability, the inspectors determined that a Phase 3 evaluation was required.
The senior reactor analyst determined that, while this finding clearly involved an
increase in the core damage frequency, the current probabilistic models were not
sufficient to quantify this change. However, the likelihood of having a loss-of-coolant
accident with a loss of instrument air, combined with emergency response personnel
failing to restore instrument air and/or bring the plant to cold shutdown conditions in
39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />, was very small.
While the design basis requires the emergency core cooling system to operate for
30 days following a design basis accident, the mission time for the risk-significant
function of the system is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This assumption exists in both the licensee's
probabilistic risk assessment and the NRCs Standardized Plant Analysis Risk Model for
Fort Calhoun Station. This assumption was developed considering that: decay heat is
lower 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into an accident, providing additional time to respond to degrading
Enclosure 2
-15-
conditions; the demands on the systems are less; and the Technical Support Center is
fully manned and capable of assisting plant staff in making affective repairs to plant
equipment.
The senior reactor analyst determined that the safety injection refueling water tank
recirculation valves would have remained closed throughout their risk-significant mission
time. Additionally, the senior reactor analyst concluded that the likelihood of a
loss-of-coolant accident combined with a loss of instrument air was sufficiently small that
further evaluation of the change in risk beyond the modeled mission time was not
required. Therefore, the failure to have an adequate abnormal operating procedure for
loss of instrument air represented a finding of very low risk significance (Green).
Enforcement. 10 CFR Part 50, Appendix B, Criterion III, requires, in part, that measures
shall be established to assure that applicable regulatory requirements and the design
basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, the measures established to assure that design information
was correctly translated into procedures were inadequate. Specifically, the requirement
to maintain air-operated valves in their required positions during a loss-of-coolant
accident with instrument air unavailable was not addressed beyond the initial
system responses when backup air accumulators and nitrogen supplies are available.
This violation of 10 CFR Part 50, Appendix B, Criterion III, is being treated as a noncited
violation, consistent with Section VI.A of the NRC Enforcement Policy
(NCV 05000285/2004002-03). This violation was entered into the licensees corrective
action program as Condition Report 200305311.
.2 Temporary Instruction 2515/153: Reactor Containment Sump Blockage (NRC
The objective of the Temporary Instruction was to support NRC review of the licensees
activities in response to NRC Bulletin 2003-01, Potential Impact of Debris Blockage on
Emergency Sump Recirculation at Pressurized Water Reactors (PWRs).
a. Inspection Scope
The inspectors reviewed the licensees response to NRC Bulletin 2003-01, Potential
Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water
Reactors, documented in Letter LIC-03-0105 dated August 3, 2003. In addition, the
inspectors reviewed NEI (Nuclear Energy Institute) 02-01, Condition Assessment
Guidelines: Debris Sources Inside PWR Containments, Revision 1, and Condition
Report 200302218 that documented the licensees response to the bulletin. The
inspectors discussed the commitments of Letter LIC-03-0105 with licensee personnel.
b. Findings and Observations
The licensee elected to implement method two of NRC Bulletin 2003-01. By
implementing method two, the licensee developed interim compensatory measures to
Enclosure 2
-16-
reduce the risk which may be associated with potentially degraded or nonconforming
emergency core cooling system recirculation functions while performing evaluations to
determine compliance. In addition, the licensee has initiated preparations to modify the
sumps in case the evaluation concludes that modifications are required. The inspectors
reviewed the licensees response to NRC Bulletin 2003-01 and noted that the licensee
had addressed the recommended six interim compensatory measures.
The inspectors reviewed portions of the completed commitments. The inspectors
observed the licensed operator training on the identification of the symptoms of a
degraded sump during a loss-of-coolant accident. The inspectors reviewed
Procedure EOP/AOP Attachments, Revision 16, and noted that the licensee had
developed Attachment 25, Methods For Refilling the SIRWT Post RAS. During the
refueling outage in the fall of 2003, the inspectors observed that the licensee was more
aggressive than previous outages in controlling foreign material and debris inside
containment.
On September 25 and 27, 2003, the licensee performed a walkdown of containment to
verify that drainage paths for recirculation were unblocked. The licensee documented
the results in Condition Report 200302218 and concluded that the drainage paths were
acceptable and unblocked.
The licensee inspected the containment emergency sumps to verify that they were free
of adverse gaps and breaches. The licensee identified some minor gaps in the screens
to both of the sumps and initiated repairs to the screens. The licensee documented the
observation in Condition Report 2000304391 and concluded that the sumps were
operable in the as-found condition. The inspectors reviewed Condition
Report 2000304391 and concluded that the minor gaps did not affect the system
operability and documented the results in NRC Inspection Report 05000285/2003006,
Section 1R15. The inspectors performed a walkdown of the sumps and did not identify
any additional adverse gaps or breaches.
4OA6 Meetings
Exit Meeting Summary
On March 31, 2004, the inspector presented the results of the significance determination
assessment discussed in Section 4OA5.1 to Mr. G. Cavanaugh, Supervisor, Station
Licensing, by telephone, who acknowledged the findings.
The results of the resident inspector activities were presented to Mr. R. Phelps, Division
Manager, Nuclear Engineering, and other members of licensee management on
April 5, 2004. The licensees management acknowledged the inspection findings and
Enclosure 2
-17-
stated that some of the material examined during the inspection was considered
proprietary. The inspectors indicated that, although examined, no proprietary
information was documented in the inspection report.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure 2
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Bannister, Plant Manager
R. Clemens, Division Manager, Nuclear Assessments
M. Core, Manager, System Engineering
M. Frans, Assistant Plant Manager
R. Haug, Manager, Chemistry
J. Herman, Manager, Nuclear Licensing
R. Phelps, Division Manager, Nuclear Engineering
M. Puckett, Manager, Radiation Protection
R. Ridenoure, Division Manager, Nuclear Operations
H. Sefick, Manager, Security and Emergency Planning
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000285/2004002-02 VIO Inadequate Diesel Generator Surveillance Test Procedure
Acceptance Criteria (Section 4OA2.1)
Opened and Closed
05000285/2004002-01 FIN Unauthorized Modification to the Diesel-Driven Auxiliary
Feedwater Pump (Section 1R22)05000285/2004002-03 NCV Inadequate Procedure for Long-term Loss of Instrument
Air (Section 4OA5.1)
Closed
05000285/2002004-00 LER Inadequate procedural guidance resulting in
noncompliance with 10 CFR Part 50, Appendix R
(Section 4AO3)05000285/2003011-03 URI Failure to Provide Means to Assure Proper Emergency
Core Cooling System Alignment During Prolonged Loss of
Instrument Air (Section 4OA5.1)
A-1 Attachment