ML060300256
| ML060300256 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/25/2006 |
| From: | Christine Lipa NRC/RGN-III/DRP/RPB4 |
| To: | Harden P Nuclear Management Co |
| References | |
| FOIA/PA-2006-0082 IR-05-012 | |
| Download: ML060300256 (36) | |
See also: IR 05000255/2005012
Text
January 25, 2006
Mr. Paul A. Harden
Site Vice President
Nuclear Management Company, LLC
Palisades Nuclear Plant
27780 Blue Star Memorial Highway
Covert, MI 49043-9530
SUBJECT: PALISADES NUCLEAR PLANT
NRC INSPECTION REPORT 05000255/2005012
Dear Mr. Harden:
On December 31, 2005, the U. S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Palisades Nuclear Plant. The enclosed report documents the inspection
findings which were discussed on January 5, 2006, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, two NRC-identified findings of very low safety (Green)
were identified. Both of these findings were determined to involve a violation of NRC
requirements. Additionally, a licensee-identified violation which was determined to be of very
low safety significance is listed in the report. However, because the violations were of very low
safety significance and because the issues have been entered into your corrective action
program, the NRC is treating these findings as a non-cited violations (NCVs) consistent with
Section VI.A.1 of the Enforcement Policy.
If you contest the subject or severity of a NCV, you should provide a response with a basis for
your denial, within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the
Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville
Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the
Palisades facility.
P. Harden -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter
and its enclosure will be available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARS) component of NRC's
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/
Christine A. Lipa, Chief
Reactor Projects Branch 4
Division of Reactor Projects
Docket No. 50-255
License No. DPR-20
Enclosure: Inspection Report 05000255/2005012
w/Attachment: Supplemental Information
cc w/encl: J. Cowan, Executive Vice President
and Chief Nuclear Officer
R. Fenech, Senior Vice President, Nuclear
Fossil and Hydro Operations
D. Cooper, Senior Vice President - Group Operations
L. Lahti, Manager, Regulatory Affairs
J. Rogoff, Vice President, Counsel and Secretary
A. Udrys, Esquire, Consumers Energy Company
S. Wawro, Director of Nuclear Assets, Consumers Energy Company
Supervisor, Covert Township
Office of the Governor
L. Brandon, Michigan Department of Environmental Quality -
Waste and Hazardous Materials Division
Michigan Department of Attorney General
DOCUMENT NAME: E:\Filenet\ML060300256.wpd
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE RIII
NAME CLipa:dtp
DATE 01/25/06
OFFICIAL RECORD COPY
P. Harden -2-
ADAMS Distribution:
HKN
LMP
RidsNrrDirsIrib
GEG
KGO
JAL3
JAE
CAA1
C. Pederson, DRS (hard copy - IRs only)
DRPIII
DRSIII
PLB1
JRK1
ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-255
License No: DPR-20
Report No: 05000255/2005012
Licensee: Nuclear Management Company, LLC
Facility: Palisades Nuclear Plant
Location: Covert, MI 49043-9530
Dates: October 1 through December 31, 2005
Inspectors: J. Ellegood, Senior Resident Inspector
J. Giessner, Resident Inspector
G. O'Dwyer, Reactor Engineer
R. Alexander, Emergency Response
Coordinator/EP Analyst
J. House, Senior Radiation Specialist
W. Snell, Senior Radiation Specialist
M. Gryglak, Reactor Inspector
S. Bakhsh, Health Physicist
Approved by: C. Lipa, Chief
Branch 4
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000255/2005012; 10/01/2005 - 12/31/2005; Palisades Nuclear Plant; Operator
Performance During Non-routine Evolutions and Events; Operability Evaluations
This report covers a 3-month period of baseline inspections. The inspections were conducted
by Region III inspectors and resident inspectors. This report includes two green findings with
associated NCVs. The significance of most findings is indicated by their color (Green, White,
Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination
Process (SDP)." Findings for which the SDP does not apply may be "Green" or be assigned a
severity level after NRC management review. 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-Revealed Finding
- Cornerstone: Initiating Events
Green: The inspectors identified one finding of very low safety significance and an
associated non-cited violation when plant personnel performed activities outside the
scope of the work package used to inspect the spent fuel pool crane. On
October 11, 2005, while raising a dry fuel storage (DFS) cask from the spent fuel pool
following loading of the cask, the emergency brake on the crane engaged. The
engaged emergency brake stopped movement of the load resulting in suspension of the
load partially out of the pool. During troubleshooting activities, the workers exceeded
the bounds of the approved work package by manipulating the brake release. This
finding represented a violation of the license by performing work contrary to
requirements specified by NUREG-0612. Corrective actions included reinforcing site
standards for procedural adherence as well as successfully lowering the DFS cask. The
licensee entered the item in the Corrective Action Program.
The finding was not suitable for evaluation under the SDP. However, because the
actions by the worker did not result in any load motion and both crane brakes remained
set, NRC management determined the finding to be of very low safety significance
(Green). This finding also affected the cross cutting area of human performance.
Cornerstone: Barrier Integrity
- Green. The inspectors identified a finding of very low significance (Green) when the
licensee failed to declare the containment air cooler, VHX-4, SW piping inoperable and
take action in accordance with licensee procedures and technical specifications when a
through-wall (pressure boundary) leak existed. This finding represented a non-cited
violation of Technical Specifications 5.4, "Procedures," in that procedures were not
properly implemented which would have resulted in declaration of inoperability of
component. Corrective actions included conducting repairs to stop the leak. The
licensee entered the item in the Corrective Action Program. The deficiency was also an
issue in the cross-cutting area of human performance in that personnel did not properly
follow the procedure for determining operability.
1 Enclosure
The inspectors determined that the issue was more than minor because the finding
impacted the barrier integrity cornerstone attribute for containment barrier performance.
The deficiency affected the barrier integrity objective of providing reasonable assurance
that physical design barriers for the containment protect the public from radionuclide
releases in that part of the boundary to a closed system for a containment penetration
was breached. The finding was of very low safety significance since the breach in the
containment boundary was small and would have very little impact on offsite dose
evaluations. (Section 1R15)
B. Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee has been
reviewed by the inspectors. Corrective actions taken by the licensee have been entered
into the licencees corrective action program. This violation and corrective actions are
discussed in Section 4OA7 of this report.
2 Enclosure
REPORT DETAILS
Summary of Plant Status
The plant operated at or near full Rated Thermal Power (RTP) during the inspection period with
the following exceptions:
- On October 18, 2005, the licensee reduced power to 35 percent power when
high vibration was indicated on the main turbine. After determining the vibrations
were related to an indication problem and not actual turbine vibration, the plant
returned to 100 percent power on October 19.
- On November 13, 2005, the licensee reduced power to 52 percent due to fouling
of cooling tower screens. The licensee returned the reactor to 100 percent
power on November 15.
- On December 14, 2005, a spurious actuation of the 1-1 EDG load sequencer
resulted in a power reduction to 81 percent due to boron addition. The licensee
returned the reactor to 100 percent power on December 15.
- On December 30, 2005, the plant shutdown to repair 3 leaking control rod drive
mechanisms and remained shutdown for the rest of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather (71111.01)
a. Inspection Scope
The inspectors reviewed the plants preparation for cold weather. This included a review
of the plants documentation including USAR and Technical Specifications (TSs),
walkdown activities, review of past adverse conditions and corrective actions as well as
a walkdown by the inspectors of equipment which could be adversely affected by cold
weather. The inspectors also reviewed the licensees cold weather procedures. The
safety system focus was on the safety injection system from the safety injection and
refueling water tank (SIRWT) and auxiliary feed water system. The documents
reviewed during this inspection are listed in the attachment.
This constitutes one sample.
b. Findings
No findings of significance were identified.
3 Enclosure
1R04 Equipment Alignment
.1 Partial Walkdowns (71111.04Q)
a. Inspection Scope
The inspectors completed two equipment alignment inspection samples by performing
partial walkdowns on the following risk-significant plant equipment:
- 1-1 emergency diesel generator during an outage for 1-2 EDG
- 'A' containment spray during an outage for 'B' containment spray
During the walkdowns, the inspectors verified that power was available, that accessible
equipment and components were appropriately aligned, and that no open work orders
for known equipment deficiencies existed which would impact system availability.
The inspectors also reviewed selected condition reports related to equipment alignment
problems and verified that identified problems were entered into the corrective action
program with the appropriate significance characterization and that planned and
completed corrective actions were appropriate and implemented as scheduled. The
documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
.1 Fire Area Walkdowns (71111.05Q)
a. Inspection Scope
The inspectors completed six fire protection inspection samples by touring the following
areas in which a fire could affect safety-related equipment:
C Condensate Pump Room (Fire Area 23)
C Emergency Diesel Room 1-2 (Fire Area 6)
C Emergency Diesel Room 1-1 (Fire Area 5)
C 1-C Switchgear Room (Fire Area 4)
C AFW Pump Room (Fire Area 24)
C East Engineering Safeguards Room (Fire Area 28)
The inspectors verified that transient combustibles and ignition sources were
appropriately controlled, and that the installed fire protection equipment in the fire areas
corresponded with the equipment which was referenced in the Updated Final Safety
Analysis Report, Section 9.6, "Fire Protection." The inspectors also assessed the
material condition of fire suppression systems, manual fire fighting equipment, smoke
detection systems, fire barriers and emergency lighting units. For selected areas, the
4 Enclosure
inspectors reviewed documentation for completed surveillances to verify that fire
protection equipment and fire barriers were tested as required to ensure availability.
The inspectors reviewed selected condition reports associated with fire protection to
verify that identified problems were entered into the corrective action program with the
appropriate significance characterization. The inspectors also verified that planned and
completed corrective actions were appropriate. The documents reviewed during this
inspection are listed in the attachment.
b. Findings
No findings of significance were identified.
.2 Fire Protection - Drill Observation (71111.05A)
a. Inspection Scope
The inspectors completed the annual inspection of evaluating the fire brigades
performance during an unannounced fire drill on December 18, 2005. The drill was
observed to evaluate the readiness of the plant fire brigade to fight fires. In addition,
reviews of procedures, fire fighting equipment, and corrective action for adverse
conditions were conducted. The inspectors evaluated the licensees critique of the drill
and actions taken as a result of the critique to verify the self-critical manner at the
debrief. Specific attributes evaluated were: (1) proper wearing of turnout gear and self-
contained breathing apparatus; (2) proper use and layout of fire hoses; (3) employment
of appropriate fire fighting techniques; (4) sufficient fire fighting equipment brought to
the scene; (5) effectiveness of fire brigade leader communications, command, and
control; (6) search for victims and propagation of the fire into other plant areas;
(7) smoke removal operations; (8) utilization of pre-planned strategies; (9) adherence to
the pre-planned drill scenario; and (10) drill objectives. This constituted one sample.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07B)
a. Inspection Scope
Regional inspectors reviewed documents associated with maintenance and inspection
of the mechanical seal heat exchangers and the bearing oil coolers for the high pressure
safety injection (HPSI) pumps (P66A & B). The heat exchangers count as two samples.
These heat exchangers were chosen based on their operational support function of
removing heat generated by the risk significant HPSI pumps. These heat exchangers
were also chosen based on the importance of the safety functions performed by the
risk-significant HPSI pumps during accidents. These heat exchangers had not been
inspected by any previous heat sink performance biennial inspection. While on site, the
inspector reviewed operability determinations, completed surveillances, vendor manual
information, associated calculations, and preventive maintenance activities; and
5 Enclosure
performed independent assessments to verify that these activities adequately ensured
proper heat transfer. The inspector also reviewed documentation to confirm that
methods used to maintain and monitor the operational effectiveness of the heat
exchangers were consistent with expected degradation and that the established
acceptance criteria were consistent with design accident requirements and accepted
industry standards. The inspectors walked down the HPSI pumps to ensure proper
installation and configuration of these heat exchangers. The inspectors verified that the
nameplates on the pumps' mechanical seal heat exchangers matched the licensees
vendor manual information.
The inspectors also reviewed documentation to verify performance of two attributes of
the ultimate heat sink (UHS.) The inspectors verified that the licensee had adequate
controls to ensure that UHS system and subcomponents were free from clogging due to
macrofouling and UHS would function properly during adverse weather conditions,(e.g.,
icing or high temperatures).
In addition, the inspectors reviewed condition reports concerning heat exchanger or heat
sink performance issues to verify that the licensee had an appropriate threshold for
identifying issues and to evaluate the effectiveness of the corrective actions to the
identified issues. The documents that were reviewed are included at the end of the
report.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification (71111.11Q)
.1 Quarterly Review
a. Inspection Scope
The inspectors completed one inspection sample pertaining to licensed operator
requalification by observing licensed operator actions in the control room simulator on
November 2, 2005. The inspectors assessed the operators ability to use plant
procedures to respond to simulated plant alarms and emergency conditions. The
inspectors assessed the operators ability to evaluate plant conditions and determine the
proper emergency action level. The inspectors assessed the licensee evaluators' ability
to evaluate the operators performance and to identify operator performance
deficiencies.
b. Findings
No findings of significance were identified.
6 Enclosure
1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13Q)
a. Inspection Scope
The inspectors completed five inspection samples. The inspectors reviewed the
following five activities to verify that the appropriate risk assessments were performed
prior to removing equipment for work. The inspectors verified that risk assessments
were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete.
When emergent work was performed, the inspectors verified that the plant risk was
promptly reassessed and managed. The inspectors verified the appropriate use of the
licensees risk assessment tool and risk categories in accordance with Administrative
Procedure 4.02, Control of Equipment, revision 29, and Fleet Procedure FP-OP-RSK-
01, Risk Monitoring and Risk Management, revision 0. Documents reviewed are listed
in the attachment.
- planned in service testing of P66A (HHSI pump) for the week of 10/2-10/8/2005;
- planned in service testing of P66B (HHSI pump) and planned P-55B (charging
pump) repack during the week of10/15-10/21/2005;
- planned work for EDG 1-2 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> outage during the week of10/22-10/28/2005;
- planned work and testing of containment spray pump P54B and P54C on
November 9, 2005; and
- unplanned work on EDG 1-2 due to fuel leak on November 21, 2005.
The inspectors also verified that condition reports related to emergent equipment
problems were entered into the corrective action program with the appropriate
significance characterization. Specific condition reports related to risk management
during maintenance activities were reviewed to verify that planned corrective actions
were appropriate and had been implemented as scheduled.
b. Findings
No findings of significance were identified.
1R14 Operator Performance During Non-routine Evolutions and Events (71111.14)
A. Inspection Scope
The inspectors completed four samples of non-routine events. For the non-routine
events described below, the inspectors reviewed operator logs, plant computer data,
and strip charts as appropriate to determine what occurred and how the operators
responded, and to determine if the response was in accordance with plant procedures:
- On October 11, 2005, during dry fuel storage loading activities, an emergency
brake for the crane engaged, suspending a dry fuel storage cask partially out of
the pool. The inspectors observed licensee activities to determine the cause of
the brake's engagement and to lower safely the load back into the pool.
- On October 19, 2005, a plant downpower to 35 percent was required by plant
alarm response procedures due to high indicated main turbine vibrations. After
7 Enclosure
investigation and troubleshooting, the high vibrations were determined to be an
indication problem and not actual vibration problem. The indication problem was
the result of a failed power supply.
- On November 13, 2005, a plant downpower to about 50 percent power was
required due to the loss of the B cooling tower caused by leaf intrusion. The
debris caused the cooling tower screens to foul relatively quickly and required
prompt action to ensure the functionality of the cooling tower was retained.
- On December 14, 2005, with the plant operating at 100 percent power, the 1-1
EDG load sequencer for design basis loads failed causing the actuation of
components which would start on the loss of offsite power with a design
accident. The failure simultaneously started the associated train's high head and
low head Emergency Core Cooling Pumps and caused the running charging
pump suction to shift from the Volume Control Tank (VCT) to the Boric Acid
Storage Tank. The charging pump injected about 200 gallons of concentrated
boric acid before being realigned to the VCT. The negative reactivity resulted in
a reduction of Power, Temperature (average), and reactor coolant system
pressure. A 10 degree F Temperature (reference) to Temperature (average)
deviation occurred, reactor coolant pressure dropped from 2060 to 1967 psig,
and power dipped from 100 to 81 percent following the boron injection and
actions taken to restore Temperature (average) and Temperature (reference).
Operators restored pressure, power, and temperature, and declared the 1-1
EDG inoperable. The inspectors observed the recovery from the transient and
the problem solving activities associated with it. All equipment responded as
expected due to the Load Sequencer Failure.
b. Findings
Introduction: The inspectors identified one finding of very low safety significance and an
associated non-cited violation when plant personnel performed activities outside the
scope of the work package used to inspect the spent fuel pool crane.
Description: On October 11, 2005, while raising a dry fuel storage cask from the spent
fuel pool following loading of the cask, the emergency brake on the crane engaged.
The engaged emergency brake stopped movement of the load resulting in suspension
of the load partially out of the pool. The licensee stopped DFS activities and developed
work instructions to inspect the crane and determine the cause for the brake
engagement.
The licensee developed a work package to inspect the crane and determine the cause
of the brake engagement. The work package developed for this activity did not include
manipulation of crane components. Status meetings held earlier in the day emphasized
that no crane component manipulations were currently planned. However, the workers,
after consultation with vendor representatives, moved the brake release to verify the
brake was set. Since early discussions regarding the scope of work clearly stated crane
components would not be manipulated, the inspectors discussed the activity with plant
management. Plant management was unaware of the manipulation prior to its
8 Enclosure
occurrence and subsequently determined the workers had exceeded the bounds of the
authorized work.
After evaluating the available data, the licensee developed and implemented a work
package to lower the cask to the floor of the spent fuel pool. Further troubleshooting by
the licensee determined the emergency brake had engaged due to an improperly set
torque limiter that engaged the brake prematurely. This torque limiter had been
adjusted using incorrect techniques in August, in part due to a failure to use a procedure
for resetting the torque limiter.
This finding also affected the cross cutting area of human performance. Specifically,
this finding addressees a failure to follow procedures.
Analysis: The inspectors concluded that working outside the bounds of a work package
on a crane with a suspended load that if dropped would damage the spent fuel pool
warranted a safety significance determination in accordance with IMC 0612. The
inspectors discussed the effects of a drop of the load with licensee personnel. Had the
load dropped, the spent fuel pool could have sustained severe damage. The inspectors
were also aware that the individuals involved in the work activity were not fully
knowledgeable of the crane's design, operation, and failure modes at the time the work
occurred. In order to compensate for the gap in knowledge, the licensee obtained
telephonic support from the crane vendor. Therefore, the inspectors concluded working
outside the bounds of the approved work package and manipulating the brake release
represented an increase in the risk of a load drop. This increase in risk is directly
associated with the reactor safety cornerstone objective of the spent fuel cooling system
as a radiological barrier.
The finding was not suitable for evaluation under the SDP. However, because the
actions by the worker did not result in any load motion and both crane brakes remained
set, NRC management determined the finding to be of very low safety significance
(Green).
Enforcement: License Amendment No. 215 approved modifications to the facility
license to increase the spent fuel pool crane capacity to 110 tons and reflect the single
failure proof design of the crane. The associated amendment request identified
NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, as a regulatory
requirement. NUREG-0612 requires licensees to have procedures in place for load
handling operations over or in the proximity to irradiated fuel. Contrary to this
requirement, the licensee performed work on the spent fuel pool crane by manipulating
its components without written procedures in place authorizing the particular task.
Because the finding was of very low safety significance and the finding was entered
into the licensee's corrective action program (CAP 01000753) this violation is being
treated as an NCV consistent with Section VI.A.1 fo the NRC enforcement Policy
9 Enclosure
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
For the three operability evaluations described in the Operability Recommendations
(OPRs) listed below, the inspectors evaluated the technical adequacy of the evaluations
to ensure that TS operability was properly justified and the subject component or system
remained available such that no unrecognized increase in risk occurred. The inspectors
reviewed the UFSAR to verify that the system or component remained available to
perform its intended function. In addition, the inspectors reviewed compensatory
measures implemented to verify that the compensatory measures worked as stated and
the measures were adequately controlled. In addition, the inspectors verified that the
condition reports generated for equipment operability issues were entered into the
licensees corrective action program with the appropriate significance characterization.
Documents reviewed are listed in the attachment.
- OPR 01000525-01, Control Room Cooler VHX-4 through-wall service water leak
- OPR000096, Intake crib degradation
accidents
b. Findings
Introduction: An inspector-identified finding of very low significance (Green) and an
associated non-cited violation of TS 5.4, "Procedures", occurred when the licensee
failed to declare the containment air cooler, VHX-4, inoperable when a through wall leak
was discovered on a service water (SW) pipe for that cooler. The SW piping serves as
part of a closed system for the containment boundary. The ASME Class III component
should have been declared inoperable with the appropriate actions taken as required by
the sites quality procedure for operability determinations, FP-OP-OL-01.
Description: On October 9, 2005, after completion of surveillance testing which
performed cycling of the inlet and outlet valves to the control air coolers service water
(SW) side, the licensee noted that level in the containment sump was rising more rapidly
than normal (from .15 gpm to .5 gallons per minute). Based on inspection and sampling
the licensee determined that the leak was from the VHX-4 control air cooler service
water supply.
The licensee marked "N/A" for an immediate operability assessment on the associated
CAP because service water to VHX-4 was not required during an accident. Although
the piping formed part of the containment boundary as a closed loop inside containment,
the licensee had previously evaluated containment integrity in Operability Evaluation
CPAL 0101971 (May 25, 2001) for leaks up to 1 gpm. Therefore, the licensee
concluded no further operability evaluation was required.
A detailed inspection on October 11, 2005 indicated the leak was coming from two
locations on a single 5/8" diameter brazed joint. Since this had occurred in the past and
was inside of the 1 gallon per minute rate of the previous evaluation the licensee
10 Enclosure
believed there was no question of operability. The inspectors questioned this
assessment since the potential impact to containment integrity was not evaluated if the
crack propagated. The inspectors questioned whether with the SW system no longer
closed, if containment pressure in an accident could act to move containment activity
through the penetration and into the SW system discharge (which discharges to mixing
basin and the lake). Although the leak was not large at the time, since it was through
wall, the licensee could not assume the flaw would not propagate. The piping is ASME
Code Class III.
The inspector reviewed the current guidance issued on September 26, 2005 in
RIS 2005-05: Revision to Guidance Formerly Contained in Generic Letter 91-18,
"Information to Licensees Regarding Two NRC Inspection Manual Sections on
Degraded and Non-conforming Conditions and on Operability" and the previous
guidance in GL 91-18 and the Part 9900 guidance: "Operable/Operability: Ensuring the
Functional Capability of a System or Component". RIS 2005-05 stated: "If the flaw is
through wall or does not meet the limits established in the Code, the component and
part of the system containing the flaw is inoperable." The previous GL 91-18 provided
similar guidance. The guidance permitted a flaw evaluation to determine if the piping
can be placed back in service. The licensee has implemented NMCs Corporate Office
Quality Procedure for Operability Determinations, FP-OP-OL-01, which states for Code
components, that the Shift manager SHALL declare a component whose pressure
boundary has leakage inoperable. For Class 3 piping, the system containing the
through wall flaw may be considered operable after it has been evaluated and found to
meet the acceptance criteria in Generic Letter 90-05. Since the licensee had not
completed a flaw evaluation, had not isolated the component and had not declared the
affected component inoperable, the inspectors concluded the licensee was not
complying with applicable requirements.
The inspectors shared the information regarding the GL 91-18 guidance with plant
management on October 13, 2005 who had been unaware of the information. The
licensee accelerated the repair of the VHX-4 cooler and completed it the same day
(October 13, 2005).
Analysis: The failure to declare the VHX-4 cooler SW piping inoperable and take action
in accordance with licensee procedures when the through wall leak existed was a
performance deficiency which warranted a significance determination. The inspectors
determined that the issue was more than minor in accordance with IMC 0612,
Appendix B because the issue impacted the barrier integrity cornerstone attributes for
containment barrier performance. The deficiency affected the barrier integrity objective
of providing reasonable assurance that physical design barriers for the containment
protect the public from radio nuclide releases because part of the boundary to a closed
system for a containment penetration was breached.
Using IMC 0609, Appendix A, "SDP Phase 1 Screening Worksheet for IE [Initiating
Events], MS [Mitigating Systems] and B [Barrier Integrity] Cornerstones" the inspectors
determined the finding impacted the barrier integrity cornerstone. Although a large
amount of water leakage could impact the flood plane and thus plant mitigation
equipment in the containment, the licensee demonstrated that the functionality of
mitigating systems was not lost. Since the finding did represent an actual open pathway
11 Enclosure
in the physical integrity of reactor containment on a designed closed system, IMC 0612,
Appendix H was used. Using Table 4.1, Containment-Related SSCs Considered for
Large Early Release Requency (LERF) Implications, due to the small size of pipe (even
if completely failed) which is less than 1-2" diameter, there is no or little impact to LERF.
Therefore this issue screens as Green. Although a complete failure of the pipe could
potentially exceed the TSs for allowed containment leakage, an additional evaluation by
the licensee indicated the impact on offsite dose would have been negligible.
This finding also affected the cross cutting area of human performance. Specifically,
this finding addressees a failure to follow procedures as well as a lack of knowledge of
procedural requirements, which, in accordance with IMC 0612, affects the cross-cutting
area of human performance.
Enforcement: Technical Specification 5.4 requires that procedures be established,
implemented, and maintained as recommended in Regulatory Guide 1.33, Revision 2.
Regulatory Guide 1.33, Appendix A, Section 1c indicated procedures should be
implemented on equipment control for safety-related equipment. The plant procedure
for implementing equipment control, Administrative Procedure 4.02, requires the Fleet
Operations Procedure, FP-OP-OL-01 "Operability Determination," be followed for
assessing the operability of plant safety related SSCs when operability is in question
(Section 9.1.2 of "Equipment Status"). Contrary to this requirement, procedure FP-OP-
OL-01 "Operability Determination" was not properly implemented in that the procedure
required the Shift Manager to declare components inoperable which had pressure
boundary component leakage (Class I, II or III). This action was not completed for the
VHX-4 cooler through wall leak, and no action was taken to evaluate the Class III
component flaw consistent with industry standards or remove the component from
service. Because this violation was associated with a finding of very low safety
significance and because the finding was entered into the licensees corrective action
program (CAP 01000525), this violation is being treated as an NCV, consistent with
Section VI.A.1 of the NRC Enforcement Policy (NCV 05000255/2005012-02). The
corrective actions included repairing the leaking component using an acceptable
procedure.
1R16 Operator Work Arounds (71111.16)
a. Inspection Scope
The inspectors completed one inspection sample regarding operator work arounds.
This was the semiannual review which evaluates workarounds for the cumulative impact
to operators in response to transients and accidents. The inspectors reviewed the
cumulative effects of deficiencies that constituted operator workarounds to determine
whether or not they could affect the reliability, availability, and potential for mis-operation
of a mitigating system; affect multiple mitigating systems; or affect the ability of
operators to respond in a correct and timely manner to plant transients and accidents.
The inspectors also assessed whether operator workarounds were being identified and
entered into the licensees corrective action program at an appropriate threshold.
Documents reviewed are listed in the attachment.
12 Enclosure
b. Findings
No findings of significance identified.
1R19 Post Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the three post-maintenance tests listed below to verify that
procedures and test activities ensured system operability and functional capability. The
inspectors reviewed the licensees test procedure to verify that the procedure
adequately tested the safety functions that may have been affected by the maintenance
activity, that the acceptance criteria in the procedure were consistent with information in
the applicable licensing basis and/or design basis documents, and that the procedure
had been properly reviewed and approved. The inspectors also witnessed the test or
reviewed the test data, to verify that test results adequately demonstrated restoration of
the affected safety function(s). Further, the inspectors reviewed condition reports to
verify that post maintenance testing problems were entered into the corrective action
program with the appropriate significance characterization. For select condition reports,
the inspectors verified that the corrective actions were appropriate and implemented as
scheduled. Documents reviewed are listed in the attachment.
C Auxiliary feedwater actuation system retest following channel repair
C VHX-4 containment air cooler service water leak repairs
C EDG 1-1 design basis accident sequencer retest
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
On December 30, 2005, the plant implemented a scheduled outage to repair the Control
Rod Drive seals on three leaking seals. The inspectors observed control room activities
during shutdown. The inspectors also completed a walkdown of accessible portions of
containment with site personnel. The inspectors evaluated these activities to ensure
licensee personnel were performing activities within TS requirements, plant procedures,
and other applicable requirements. This activity extended into the first quarter of 2006;
therefore, the remainder of the inspection will be included in NRC Inspection Report 05000255/2006-02. The inspectors completed one inspection sample.
b. Findings
No findings of significance were identified.
13 Enclosure
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors witnessed two surveillance tests and/or reviewed test data of selected
risk-significant SSCs, listed below, to assess, as appropriate, whether the SSCs met
the requirements of the TS; the UFSAR; Palisades Administrative Procedure 9.20,
TS Surveillance and Special Testing Program; Engineering Manual EM-09-02 and
EM-09-04, Inservice Testing of Plant Valves and Inservice Testing of Selected Safety
Related Pumps. The inspectors also determined whether the testing effectively
demonstrated that the SSCs were operationally ready and capable of performing their
intended safety functions. Further, the inspectors reviewed selected condition reports
regarding surveillance testing activities. The inspectors verified that the identified
problems were entered into the licensees corrective action program with the appropriate
significance characterization and that the planned and completed corrective actions
were appropriate. Additional documents reviewed are listed in the attachment.
C QO-19, Inservice Test Procedure on P66B, High Pressure Safety Injection Pump
C DWO-1, TS Surveillance Procedure: Operator Daily/Weekly Items for plant heat
balance and calculation of the calorimetric power
b. Findings
No findings of significance were identified.
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a. Inspection Scope
The inspectors performed a screening review of Revisions 11 and 12 of the Palisades
Nuclear Plant Site Emergency Plan to determine whether the changes made in
Revisions 11 and 12 decreased the effectiveness of the licensees emergency planning
program. The screening review of these revisions did not constitute an approval of the
changes and, as such, the changes are subject to future NRC inspection to ensure that
the emergency plan continues to meet NRC regulations.
These activities completed one inspection sample.
b. Findings
No findings of significance were identified.
14 Enclosure
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Job-In-Progress Reviews
a. Inspection Scope
The inspectors selected two jobs being performed in radiation areas, potential airborne
radioactivity areas, and high radiation areas (HRAs) for observation of work activities
that presented the greatest radiological risk to workers and included areas where
radiological gradients could be present. This work was estimated to result in higher
collective doses and involved dry cask fuel storage welding operations and other
selected work areas.
The inspectors reviewed radiological job requirements including radiation work permit
(RWP) and work procedure requirements, and attended as low as is reasonably
achievable (ALARA) job briefings. Job performance was observed with respect to these
requirements to ascertain whether radiological conditions in the work area were
adequately communicated to workers through pre-job briefings and radiological
condition postings. This review represented one sample.
The inspectors also evaluated the adequacy of radiological controls including required
radiation, contamination and airborne surveys for system breaches and entry into HRAs.
Radiation protection job coverage, which included direct visual surveillance by radiation
protection (RP) technicians along with the remote monitoring and teledosimetry systems
and contamination control processes was reviewed to assess the effectiveness of
worker protection from radiological exposure. This review represented one sample.
Work in HRAs having significant dose rate gradients was observed to assess the
application of dosimetry to effectively monitor exposure to personnel and to evaluate the
adequacy of licensee controls. The inspectors observed RP coverage of dry cask fuel
storage welding operations which required controlling worker locations based on
radiation survey data and real time monitoring using teledosimetry in order to maintain
personnel radiological exposure ALARA. This review represented one sample.
b. Findings
No findings of significance were identified.
15 Enclosure
2OS2 As Low As Is Reasonably Achievable (ALARA) Planning And Controls (71121.02)
.1 Problem Identification and Resolution
a. Inspection Scope
The inspectors determined that the licensees self-assessment program identified and
addressed repetitive deficiencies and significant individual deficiencies that were
identified in the licensee's problem identification and resolution process. This review
represented one sample.
Corrective action reports related to the ALARA program were reviewed and staff
members were interviewed to determine if follow-up activities had been conducted in an
effective and timely manner commensurate with their importance to safety and risk
using the following criteria:
- initial problem identification, characterization, and tracking;
- disposition of operability/reportability issues;
- evaluation of safety significance/risk and priority for resolution;
- identification of repetitive problems;
- identification of contributing causes;
- identification and implementation of effective corrective actions;
- resolution of NCVs tracked in the corrective action system; and
- implementation/consideration of risk-significant operational experience feedback.
This review represented one sample.
b. Findings
No findings of significance were identified.
Cornerstone: Public Radiation Safety
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
(71122.01)
.1 Inspection Planning
a. Inspection Scope
The inspectors reviewed the most recent Radiological Effluent Release Report for 2004,
dated March 29, 2005, along with current effluent release data to determine if the
program was implemented as described in the Radiological Environmental TS/Offsite
Dose Calculation Manual (RETS/ODCM), and the Updated Final Safety Analysis Report
(UFSAR). The effluent report was also evaluated to determine if there were any
significant changes to the ODCM or to the radioactive waste system design and
operation. There were no significant changes to the ODCM. Radioactive waste system
modifications and licensee technical reviews were evaluated to determine if those
16 Enclosure
changes would alter dose consequences to the public and if there would be any
potential impact on radiation monitor set-point calculation methodology. There were no
anomalous results in the effluent report.
The RETS/ODCM and UFSAR were reviewed to identify the effluent radiation
monitoring systems and associated flow measurement devices. Licensee records
including condition reports, self-assessments, audits, and special reports were reviewed
to determine if there were any radiological effluent performance indicator occurrences or
any unanticipated offsite releases of radioactive material for follow-up. The UFSAR
description of all radioactive waste systems was reviewed. This review represented one
sample.
b. Findings
No findings of significance were identified.
.2 Onsite Inspection
a. Inspection Scope
The inspectors walked down the major accessible components of the gaseous and
liquid release systems, including radiation and flow monitors, tanks, and vessels. This
was done to observe current system configuration with respect to the description in the
UFSAR, ongoing activities, and equipment material condition. This review represented
one sample.
The inspectors reviewed system diagrams of the radioactive liquid waste processing and
release systems to determine how liquid radwaste was processed to determine if
appropriate treatment equipment was used and that radioactive liquid waste was
processed in accordance with procedural requirements. Liquid effluent release
packages including projected doses to the public were reviewed to determine if
regulatory effluent release limits were exceeded. The inspectors reviewed system
diagrams of the radioactive gaseous effluent processing and release systems and
observed the collection and analysis of a gaseous radwaste sample to determine if
appropriate treatment equipment was used and if the radioactive gaseous effluent was
processed and released in accordance with RETS/ODCM requirements. Radioactive
gaseous effluent release data including the projected doses to members of the public
was evaluated to determine if regulatory effluent release limits were exceeded. This
review represented one sample.
The inspectors reviewed the licensees process for making releases with inoperable
effluent radiation monitors to determine if adequate compensatory sampling and
analyses were performed and to determine if an adequate defense-in-depth was
maintained against an unmonitored, unanticipated release of radioactive material to the
environment. This included projected radiological doses to members of the public.
There were no abnormal releases noted. This review represented one sample.
The ODCM was reviewed for any significant changes. Radioactive waste system
modifications including licensee technical reviews were evaluated to determine if those
17 Enclosure
changes would alter dose consequences to the public, and if there would be any
potential impact on radiation monitor set-point calculation methodology. System
modifications were reviewed to determine if they would impact the effluent monitoring or
release controls and if the changes would affect the licensees ability to maintain
effluents ALARA. The inspectors also reviewed the licensees offsite dose calculations
and discussed the process with a cognizant licensee representative. This review
represented one sample.
The inspectors reviewed a selection of monthly, quarterly, and annual dose calculations
to ensure that the licensee properly calculated the offsite dose from radiological effluent
releases and to determine if any annual RETS/ODCM (i.e., Appendix I to
10 CFR Part 50) values were exceeded. This review represented one sample.
The inspectors reviewed air cleaning system surveillance test results to determine if the
system was operating within the licensees acceptance criteria. The inspectors
reviewed surveillance test results for the vent flow rates and determined if the flow rates
were consistent with UFSAR values. This review represented one sample.
The inspectors reviewed records of instrument calibrations performed since the
last inspection for each point of discharge effluent radiation monitor and flow
measurement device. The current effluent radiation monitor alarm set point values were
reviewed for agreement with RETS/ODCM requirements. The inspectors also reviewed
calibration records of radiation measurement (i.e., counting room) instrumentation
associated with effluent monitoring and release activities. Quality control data
for the radiation measurement instruments were evaluated to determine if the
instrumentation was operating under statistical control and that any problems
observed were addressed in a timely manner. This review represented one sample.
The inspectors reviewed the results of the interlaboratory comparison program to
determine the adequacy of the quality of radioactive effluent sample analyses performed
by the licensee. The inspectors reviewed the licensees quality control evaluation of the
interlaboratory comparison test results. In addition, the inspectors reviewed the results
from the licensees quality assurance audits to determine whether the licensee met the
requirements of the RETS/ODCM. This review represented one sample.
b. Findings
No findings of significance were identified.
.3 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed the licensees self-assessments, audits, and special reports
related to the radioactive effluent treatment and monitoring program since the last
inspection to determine if identified problems were entered into the corrective action
program for resolution. The inspectors also determined whether the licensee's
self-assessment program identified and addressed repetitive deficiencies or significant
individual deficiencies that were identified in problem identification and resolution.
18 Enclosure
The inspectors also reviewed corrective action reports from the radioactive effluent
treatment and monitoring program, interviewed staff and reviewed documents to
determine if the following activities were being conducted in an effective and timely
manner commensurate with their importance to safety and risk:
- Initial problem identification, characterization, and tracking;
- Disposition of operability/reportability issues;
- Evaluation of safety significance/risk and priority for resolution;
- Identification of repetitive problems;
- Identification of contributing causes;
- Identification and implementation of effective corrective actions;
- Resolution of non-cited violations tracked in the corrective action system; and
- Implementation/consideration of risk significant operational experience feedback.
This review represented one sample.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems (71152)
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues
during baseline inspection activities and plant status reviews to verify that condition
reports were being generated and entered into the corrective action program with the
appropriate significance characterization. For select condition reports, the inspectors
also verified that identified corrective actions were appropriate and had been
implemented or were scheduled to be implemented in a timely manner commensurate
with the significance of the identified problem.
b. Findings
No findings of significance were identified.
.2 Semi-annual Trend Review
The inspectors performed a semi-annual trend review to determine that a more
significant safety issue did not exist than would be apparent in a single condition report.
The inspectors reviewed the Operations Department trend reviews and Palisades
Management Review Meeting book for October 2005. The inspectors also reviewed
condition reports to identify potential trends.
19 Enclosure
b. Findings
No findings of significance were identified.
4OA3 Event Follow-up
.1 (Closed) LER 05000-255/2005-002-01: Emergency Diesel Generator 1-2 Excessively
Loaded in Certain Postulated Post-Accident Scenarios - Condition Prohibited by TSs
and a condition that could have prevented the fulfillment of the safety function needed to
mitigate the consequences of an accident.
On February 15, 2005, the licensee identified that under certain postulated scenarios
the emergency diesel generator (EDG) 1-2 could be overloaded when the pressurizer
heaters re-energized after initially being load shed on a blackout signal. This overload,
the licensee determined, could result in an EDG 1-2 trip. The licensee determined the
cause of this event was a circuit modification which was implemented in 1986 which
removed the original plant design that blocked the heater restoration with a safety
injection signal present. Corrective actions included modifications to the circuitry to
preclude breaker closure on the pressurizer heaters with a safety injection signal
present. This finding is more than minor because it had a credible impact on safety, in
that the EDG could be tripped on overload, cause a loss of a safety related bus, and
require manual operator action to restart the EDG and manually shed the required load
during a safety injection with a loss of offsite power. The finding affects the Mitigating
System Cornerstone and was considered to have very low safety significance (Green)
using Appendix A of the IMC 0609 because the probability of events occurring which
require a safety injection with a loss of offsite power and a loss of the other EDG are
very low. This was verified using the phase 2 worksheets using the site specific risk
notebook and reviewed by a regional senior reactor risk analyst. The inspectors
validated that the higher initiating frequency events, such as a loss of offsite power
alone, would not cause the EDG 1-2 to overload. This licensee-identified finding was a
violation of TS 3.8.1 since the 1-2 EDG inoperability existed for greater than the allowed
action time and so is a condition prohibited by TSs. The enforcement aspects of the
violation are discussed in Section 4OA7. This LER is closed.
.2 (Closed) LER 50-255/2005-05-005: Reactor Protection System Actuation and Auxiliary
Feedwater Actuation
On September 1, 2005 the licensee manually tripped the reactor due to a hydrogen leak
on the main generator. Following the reactor trip, the auxiliary feedwater system
actuated to maintain steam generator level. The licensee stabilized the plant in mode
three and successfully repaired the hydrogen leak. The inspectors reviewed the LER
and no findings of significance were identified. No violation of NRC requirements
occurred. The LER is closed.
20 Enclosure
4OA5 Other (71114.04)
.1 Temporary Instruction 2515/161 - Transportation of Reactor Control Rod Drives in
Type A Packages
a. Inspection Scope
The inspectors conducted interviews with cognizant licensee personnel who conducted
record reviews to verify that: (1) the licensee had undergone refueling activities since
calender year 2002; and (2) did not ship irradiated control rod drive mechanisms in
Department of Transportation Specification 7A, Type A packages during the time frame
2002 to the present.
b. Findings
No findings of significance were identified.
.2 Emergency Action Level and Emergency Plan Changes (URI 05000255/2003008-03)
The inspectors discussed with the licensee staff the January 22, 2004, Integrated
Inspection Report which identified an unresolved item regarding previous changes to the
emergency plan which potentially resulted in the use of a non-standard emergency
action level classification scheme. The inspectors advised the licensee that this issue
will continue to be evaluated in 2006.
These activities did not constitute an inspection sample.
.3 Operation of an Independent Spent Fuel Storage Installation (ISFSI) (60855.1)
a. Inspection Scope
Loading Campaign
The inspectors reviewed the loading procedures and observed activities associated with
the loading and transfer of two NUHOMS 32 PT casks. During the lift of the first loaded
canister and transfer cask out of the Spent Fuel Pool (SFP), the mechanical brake on
the fuel building crane engaged, resulting in the cask being suspended in the SFP pit
with the top of the cask approximately 5 feet above the water surface. The inspectors
observed and evaluated the licenses response during the event. The inspectors also
reviewed the cranes annual inspection records, work orders associated with trouble
shooting activities, and the Root Cause Analysis Report associated with this event. The
inspectors evaluated the adequacy of the short and long-term corrective actions that the
licensee proposed and initiated to prevent future occurrences of similar issues.
The inspectors reviewed radiological surveys to confirm that the cask radiation and
contamination levels did not exceed requirements specified in the license. The
inspectors reviewed a number of condition reports that related to the dry fuel storage
project. The inspectors also evaluated a completed documentation package for
Horizontal Storage Module 9. Specifically, the inspectors reviewed the welding records,
21 Enclosure
the vacuum drying and helium leak testing records, as well as the visual and dye
penetrant records.
10 CFR 72.212 Evaluation and 10 CFR 72.48 Screenings
The inspectors reviewed the licensees Title 10 Code of Federal Regulations (CFR)
72.212 evaluation to verify that it addressed the NRC issuance of two new exemptions
that related to the licensees fuel selection process and 10 CFR 50.68(b) requirements.
The inspectors also reviewed a number of 10 CFR 72.48 screening documents to verify
that changes made to the dry fuel storage process or the cask components did not
adversely impact the design of the cask.
Fuel Selection
The inspectors reviewed the licensees fuel selection process to verify that the licensee
incorporated all of the physical, thermal, and radiological fuel acceptance parameters
specified in the NRC-granted exemptions into the fuel selection process. The inspectors
reviewed the fuel selection procedure, qualification records for each assembly to be
loaded in the first Dry Shielded Canister and the Loading Plan. The inspectors also
reviewed the loading procedure and the complete Fuel Move Sheets for the same
canister.
Training
The inspectors reviewed the licensees training program for the training of new welders.
The inspectors reviewed qualification records for the new welders, including on-the-job
evaluations and the final written examination records. The inspectors observed the new
welders weld the first cask. The inspectors also verified that other personnel obtained
the necessary training. The inspectors evaluated the licensees approach to train
personnel to unload a cask during an emergency.
b. Findings
No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. P. Harden and other members of
licensee management on January 5, 2006. Licensee personnel acknowledged the
findings presented. The inspectors asked licensee personnel whether any materials
examined during the inspection should be considered proprietary. No proprietary
information was identified.
22 Enclosure
.2 Interim Exit Meetings
Interim exit meetings were conducted for:
- The access control to radiologically significant areas program, the ALARA
planning and controls program, and the radioactive gaseous and liquid effluent
treatment and monitoring systems program under the occupational and public
radiation safety cornerstones with Mr. P. Harden on November 4, 2005.
- Biennial Heat Sink Performance with Mr. D. Mims, Palisades Site Director and
Mr. G. Hettel, Plant General Manager, on December 2, 2005.
- Independent Spent Fuel Storage Installation with Mr. G. Hettel, Plant General
Manager and others on December 2, 2005.
- Emergency Preparedness Inspection with Mr. T. Blake on December 21, 2005.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
- The licensee identified a design error which resulted in overloading an EDG
during certain accidents. This resulted in EDG inoperability for longer than the
7 days permitted by TS 3.8.1 Action B. This was identified in the licensees
CAP 01001432 and LER 05000-255/2005-002-01. This finding was of very low
safety significance because the accident scenarios were sufficiently infrequent
that the inoperability had a very low impact on plant risk
ATTACHMENT: SUPPLEMENTAL INFORMATION
23 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
P. Harden, Site Vice President
T. Blake, Nuclear Safety Assurance Manager
B. Brogan, Probabilistic Safety Analysis Engineer
N. Brott, Senior EP Coordinator
M. Carlson, Engineering Director
B. Dotson, Regulatory Compliance
C. Fritts, Programs and Analysis Supervisor
J. Hager, Heat Exchanger Program Engineer
R. Harvill, Program and Analysis Engineer
G. Hettel, Plant Manager
L. Lahti, Licensing Manager
D. Malone, Regulatory Affairs
C. Moeller, Radiation Protection Supervisor
B. Patrick, Radiation Protection Manager
C. Plachta, Radiation Protection Supervisor
B. Rice, Dry Fuel Storage Project Manager
R. Schmidt, HPSI System Engineer
J. Schwan, former HPSI System Engineer
K. Smith, Operations Manager
M. Sullivan, Chemistry Supervisor
M. Sweet, EP Coordinator
R. Tiffany, Site Maintenance Rule Coordinator
J. Voskuil, Engineer
K. Yeager, Assistant Operations Manager
Nuclear Regulatory Commission
M. Padovan, Project Manager, NRR
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000255/2005012-01 NCV Spent Fuel Pool Crane Manipulated Outside bounds of
Approved Procedures (Section 1R14)05000255/2005012-02 NCV Failure to Declare VHX-4 Cooler Inoperable with a
Through-wall Piping Leak (Section 1R15)
Closed
05000255/2005012-01 NCV Spent Fuel Pool Crane Manipulated Outside bounds of
Approved Procedures (Section 1R14)
1 Attachment
05000255/2005012-02 NCV Failure to Declare VHX-4 Cooler Inoperable with a
Through-wall Piping Leak (Section 1R15)
05000255/2005-002-01 LER Emergency Diesel Generator 1-2 Excessively Loaded in
Certain Postulated Post-Accident Scenarios
05000255/2005-005-00 LER Manual Reactor Trip Due to Hydrogen Leak on the Main
Generator
Discussed
05000255/2003008-03 URI Emergency Action Level and Emergency Plan Changes
2 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety but rather that
selected sections or portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a documents on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
1R01 Adverse Weather Protection
SOP-23 Attachment 8 CL CWCL; Cold Weather Checklist; November 29, 2005;
Revision 21
WO 00030748 01; Perform Cold Weather Checklist; October 25, 2005
AR01005931; NRC Questions Screen House Ventilation Configuration, Revision 0
CE007604; Evaluate whether MV-ES3243, SIRW TK LT ISOL, Cold Weather Protection
is Adequate; October 29, 2003
CAP03800; Intake Structure Ventilation Operation Contrary to UFSAR Described
Operation; October 8, 2003
1R04 Equipment Alignment
Palisades Nuclear Administrative Procedure 4.02; Control of Equipment; Revision 29
Risk Report for Risk Measures for In Service Items if Taken OOS; Generated
October 24, 2005
DWG M 208 1A; Piping and Instrument Diagram - Service Water System; Revision 53
DBD 1.02; Design Basis Document for Service Water System; Revision 7
DBD-2.03; Containment Spray System; Revision 7
SOP-4; Containment Spray System; Revision 22
DWG 204; Safety Injection Containment Spray and Shutdown Cooling System;
Revision 31
1R05 Fire Protection
Palisades Nuclear Plant Fire Hazards Analysis; Revision 5
USNRC SER for GL 86-10; Fire Protection; September 1, 1978
EA-APR-95-032; Evaluation Of Fire Detection and Suppression System Installed in 1-C
Switchgear Room to Satisfy Appendix R Requirements; October 7, 1995
Palisades Fire Drill Critique and Drill Guide for 18 December Fire Drill; Revision Original
IR07 Heat Sink Performance
Work Order 24011171; Rebuild P67A LPSI Pump, Replace Bearings, Mechanical Seal
and Heat Exchanger with New Design per EAR-99-238 & CPAL-99-2533; January 18,
2001
SAR00806809; Pre-NRC Biennial Ultimate Heat Sink Snapshot Self-Assessment;
November 10, 2005
Procedure No T-223: Component Cooling Water Flow Verification; October 31, 2003,
Revision 13
Letter from V. C. Hall to D. J. Olver; HPSI & LPSI Pump Cooling Spec Comments;
July 20, 1967
Letter from G. J. Parks to J. D. Alderdink; (D255/1993 and 7873/0394); October 28,
1986
3 Attachment
Letter from Durametallic to Palisades; T. E. Cook to G. Szczypka; Cooling requirements
for HPSI pumps; October 28, 1986
Letter from Durametallic to Palisades; Cooling requirements for CS pumps; T. E. Cook
to T. Peterson; September 23, 1986
PPAC SWS026A-4B; Diver Inspection of Traveling Screens & Associated Equipment;
February 14, 2005
WO 24422270-9; Diver Pumped Sand & Mussels from between Trash Racks &
Traveling Screens; February 17, 2005
PPAC SWS175A; Diver Inspection/Cleaning of Intake Bay; completed June 14, 2004
PPAC CWS086C; Diver Inspection/Cleaning of Intake Bay; completed January 12, 2005
PPAC SWS175; Diver Inspection/Cleaning of Intake Bay; completed June 21, 2005
Chemistry Operating Procedure COP-16A; CCW System Chemistry; Revision 13,
March 2, 2005
Chemistry Operating Procedure COP-3; ESS System Chemistry; Revision 25,
May 3, 2005
Procedure No. QO-16, Inservice Test Procedure - Containment Spray Pumps -
Section 5.5 performed on December 21, 2004; Revision 23
EM-09-16; Heat Exchanger Condition Assessment Program; Revision 4
Plant Industry Experience Traveler for 1993 Op Ex, RHR operation results in CCW
water hammer; April 24, 1995
NOS Observation Report 2004-004-8-024; Generic Letter 89-13 Program;
December 15, 2004
M0001GA 8001; HPSI Pump Vendor Manual; Current Compilation ONP-12, Acts of
Nature; Revision 19
OPR 110 (associated with CAP 49234); CCW temperature to ESS pumps may exceed
design after some LOCA scenarios; August 23, 2005
NMC RFQ 20315; Request for bids on CCW to ESS pumps flow requirements;
September 30, 2005
CAP006870; P-67A minor bearing fault indication found during RO-98; November 18,
1999
CAP032055; HPSI Pump P-66A Seal Cooling Heat Exchanger Missing Bolt;
November 8, 2002
CAP032245; CPAL-97-1363 didnt rigorously evaluate increased CCW Temp on ESS
pump seals during one potential scenario; November 26, 2002
CAP034799; CCW Flow to P-66A and P-67B found low during T-223; April 1, 2003
CAP049234; Post-LOCA Analyzed CCW Temperature to ESS pumps Exceeds Design
Value during two scenarios; August 16, 2005
A/R No. 01006110, A/R Type- CAP; NRC Identified Incorrect Model Number Used in
Request for Quote; originated December 2, 2005
A/R No. 01006556, A/R Type- CAP; NRC Identified FSAR HPSI cooling statement
should be addressed in CCW Rerate; originated December 6, 2005
1R11 Licensed Operator Requalification
License Operator requalification, simulator evaluation Cycle 05E, November 21, 2005
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
Risk assessment for Work Week 2540;10/2-10/8/2005 (yellow) for In service testing of
P66A (HHSI pump)
4 Attachment
Risk assessment for Work Week 2542; 10/15-10/21/2005 (yellow) for In service testing
of P66B (HHSI pump) and P-55B (charging pump) repack
Risk assessment for Work Week 2543; 10/22-10/28/2005 (yellow) for EDG 1-2 planned
90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> outage (yellow)
Risk assessment for Work Week 2547; 11/13-11/19/2005 (yellow) for In service testing
of containment spray pump P54B and P54C
Risk assessment for Work Week 2543; 11/20-10/26/2005 (yellow) for EDG 1-2
unplanned outage (yellow)
1R14 Operator Performance During Nonroutine Evolutions and Events
Palisades Challenge Board summary document; Final Actions Prior to Power
Escalation, Turbine Vibration Forced Outage; October 18, 2005
WO 00028215 01; L-3: Contingency- Dry Fuel Storage 200; October 11, 2005
RCE0100065901; Root Cause Analysis Report CAP01000659
WO 00028215 53 L-3: Contingency- Dry Fuel Storage 200; October 12, 2005
1R15 Operability Evaluations
FP-OP-OL-1; Corporate Office Quality Procedures: Operability Determination;
Revision 1
CAP 01000525; Containment Sump Fill rate increased During QO-1; November 9, 2005
OPR 01000525-01; VHX-4 Leaking Service Water to Containment Sump; October 12,
2005
OPR - CAP033264; CCW Design Temperature of 140 degrees; February 5, 2003
CAP049616; Degradation of Palisades Intake Crib; September 9, 2005
OPR000096; Palisades Intake Crib; April 19, 2005
1R16 Operator Workaround
A010 01007214; Feed Reg Valve Position Indication Not Screened as an OWA;
December 21, 2005
Procedure No 4.12; Operator Workaround Program; Revision 2
Operator Burden List; December 21, 2005
EOP-1.0; Standard Post-Trip Actions; Revision 12
1R19 Post Maintenance Testing
WO 00028548 01; Auxiliary Feedwater Actuation System; November 3, 2005
QI- 39; Auxiliary Feedwater Actuation System Logic Test; October 6, 2005
WO 00029819 01; VHX-4 Containment Air Cooler Leak Repairs; October 13, 2005
Engineering Assistance Request (EAR -2001-0367); Permanent Plugs for Containment
Air Cooler VHX-4; June 19, 2001
QO-1; Safety Injection; Revision 49; December 15, 2005
WO 00110216 01; D/G 1-1 Load Sequencer; December 14, 2005
1R22 Surveillance Testing
QO-19; Inservice Test Procedure - HPSI Pumps and ESS Check Valve Operability Test;
Revision 24, performed October 20, 2005
EA-ELEC08-0001; Engineering Analysis - Uncertainty Calculation for the Secondary
Calorimetric Heat Balance; Revision 2
DWO-1; TS Surveillance Procedure: Operator Daily/Weekly Items Modes 1,2,3 and 4;
Revision 71
5 Attachment
EA-HAR-91-10; Engineering Analysis - Heat Balance Adjustment for Moisture Content
of Steam; Revision 0
EA-BWB-96-01; Engineering Analysis - Heat Balance Calculation Using the Ultrasonic
Flowmeter Measurement Device; Revision 5
Inspection Procedure 61706; NRC Inspection Manual Core Thermal Power Evaluation;
July 14, 1986
1EP4 Emergency Action Level and Emergency Plan Changes
Palisades Nuclear Plant Site Emergency Plan; Revisions 11 and 12
2OS1 Access Control to Radiologically Significant Areas; and
2OS2 ALARA Planning And Controls
RCA Entries 100 Millirem or Greater; November 1, 2004 - November 1, 2005
Performance Indicator Data for Occupational and Public Radiation Safety; August 8,
2005
CAP049175; HRA Identified During Routine Monthly Survey; August 10, 2005
CAP049040; Worker Contaminated on Right Shoe; August 3, 2005
CAP049176; Contamination Found on Floor in 602' Miscellaneous Waste Tank Room;
August 10, 2005
CAP049351; HEPA Vacuum Inappropriately Stored; August 24, 2005
CAP049809; Neutron Dose Measurement; September 21, 2005
RWP 536; Containment Entries by Operations with Reactor Critical; October 21, 2005
RWP 537; Containment Entries by Maintenance Dept. with Reactor Critical; October 27,
2005
RWP 539; Dry Fuel Storage - Cask HSM007; Revision 2
FHS-M-39B; Fuel Loading and DSC Sealing Operations for NUHOMS 32PT Dry Fuel
Loading Operations; Revision 10
CE012085; RV-2203 Gas Header to VCT Relief Apparently Leaking By; November 30,
2004
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
2004 Annual Radioactive Effluent Release Report; March 29, 2005
Radiochemistry Cross Check Data for the 2nd, 3rd, and 4th Quarters of 2004
Operability of Main Steam Line Gross Gamma Activity Monitor; January 10, 2005
RT-85C; Fuel Handling Area Ventilation Filter Testing; Revision 7
NUCON; I-131 Removal Efficiency Determination of Adsorbent Samples; September 14,
2004
SA011366; Snapshot Report: Self-Assessment RETS; August 1, 2005
2005-003-8-022; Nuclear Oversight Observation Report; October 10, 2005
RR-9B; Radwaste Discharge Monitor Calibration; January 29, 2005
RR-9I; Waste Gas Discharge Monitor RIA-1113 Calibration; February 12, 2005
RR-84A; Iodine/Particulate Effluent Monitor RIA-2325 Calibration; August 27, 2005
RR-84B; Noble Gas Effluent Monitor RIA-2326 Calibration; October 11, 2005
RR-84C; Noble Gas Effluent Monitor RIA-2327 Calibration; June 22, 2005
DWR-10; Stack Effluent Sampling Calculations and Records, Revision 24
Gamma Spectroscopy Report; Stack Gas Iodine; November 2, 2005
CA026442; Radiation Monitoring System-Action Plan Development; January 17, 2005
ACE003633; RIA-1113 (Waste Gas Monitor) Switch Failure; September 6, 2005
6 Attachment
CAP048116; Rad Effluent Releases from the VCT Due to Leakage of RV-2203 and
CK-CVC2073; June 1, 2005
CAP048646; Record Keeping Enhancement for RETS/REMP Sampling; July 7, 2005
CAP049734; Release Rate Verification for T-91 Utility Water Tank Not Documented
Correctly; September 16, 2005
CAP049719; Offsite Dose Calculation Not Completed Within Required Time Frame;
September 15, 2005
CAP049880; RIA2327 High Range Noble Gas Stack Monitor Rad Level Spike;
September 26, 2005
Effluent Data 4th Quarter 2005; November 3, 2005
CH 4.39; Lower Limit of Detection Data, Detectors 1, 2 and 3; Revision 13
Gamma Spectroscopy Calibration Data, Detectors 1, 2 and 3
4OA2 Problem Identification and Resolution
Palisades Management Review Meeting; October 2005
Operations Department Monthly Performance Report; June and November 2005
4OA3 Event Follow-up
LER 05000-255/2005-05-005; Reactor Protection System Actuation and Auxiliary
Feedwater Actuation; October 24, 2005
LER 05000-255/2005-002-01; Emergency Diesel Generator 1-2 Excessively Loaded in
Certain Postulated Post-Accident Scenarios; September 27, 2005
4OA5 Other Activities
Loading Campaign
Condition Reports, generated between August 2005 and December 2005 for Dry Fuel
Storage Project
Procedure, FHS-M-39B; "Fuel Loading and DSC Sealing Operations for NUHOMS 32PT
Dry Fuel Loading Operations," Revision 12
Procedure, FHS-M-39C; "Dry Fuel Loading Operations Loaded NUHOMS DSC/Transfer
Cask Transfer to ISFSI"
Procedure, FHS-M-40B; "NUHOMS 32PT Unloading," Revision 1
Root Cause Analysis Report, No. CAP 01000659; "Crane operator heard loud noise
during lift with L-3 crane"
Work Order, No. 24422117; "L-3, annual inspection (NDE), Perform yearly inspection of
crane, using procedure MSM-13"
Work Order, No. 2821501; "L-3: Contingency-Dry Fuel Storage 200," October 11, 2005
Work Order, No. 2821553; "L-3: Contingency-Dry Fuel Storage 200," October 12, 2005
Work Order, No. 3087201; "Contingency-Dry Fuel Storage 200," December 1, 2005
Work Order Package, No. 2424201; "HSM-009; 2005 DFS Loading: Load DSC,"
November 23, 2005
10 CFR 72.212 Evaluation and 72.48 Screening
72.48 Screen, No. 05-0772; "Fuel Loading and DSC Sealing Operations for NUHOMS
32PT Dry Fuel Loading Operations"
72.48 Screen, No. 05-0793; "Fuel Loading and DSC Sealing Operations for NUHOMS
32PT Dry Fuel Loading Operations"
72.48 Screen, No. 05-0783; "10 CFR 72.212 and Certificate of Compliance Evaluation
Report for NUHOMS-32PT System"
7 Attachment
72.48 Screen, No. 05-0791; "Dry Fuel Loading Operations-Loaded NUHOMS
DSC/Transfer to ISFSI" 72.48 Screen, No. 05-0802; "Fuel Loading and DSC Sealing
Operations for NUHOMS 32PT Dry Fuel Loading Operations"
Report No. PNP 721004; "Palisades 10CFR 72.212 and Certificate, Compliance
Evaluation Report for NUHOMS 32PT System," Revision 2, October 7, 2005
Fuel Selection
DSC Loading Plan, DCS Serial Number PNP-32PT-K10-S125, Procedure
No. EM-04-56, Attachment 2, Revision 1
Fuel Assembly Qualifications, Procedure No. EM-04-56, Attachment 1, Revision 1
Procedure, No. EM-04-56; "Fuel Selection for Dry Fuel Storage," Revision 1
Procedure No. FHSO-17A; "MSB/DSC Loading Procedure," Fuel Move Sheets,
Attachment 1, Revision 3
Training
Training Records; "2005 Dry Fuel Storage Load Campaign"
8 Attachment
LIST OF ACRONYMS USED
ADAMS Agency-Wide Document and Management System
ALARA As Low As Is Reasonably Achievable
AR Action Request
CAP Corrective Action Program
CCW Component Cooling Water
CFR Code of Federal Regulations
CR Condition Report
DC Direct Current
EDG emergency diesel generator
EOP Emergency Operating Procedures
ESS Engineered Safety System
HPSI High Pressure Safety Injection
IMC Inspection Manual Chapter
ISFSI Independent Spent Fuel Storage Installation
LPSI Low Press Safety Injection
NCV Non-Cited Violation
NMC Nuclear Management Company
ODCM Offsite Dose Calculation Manual
PARS Publicly Available Records
PI Performance Indicator
PRA Probabilistic Risk Assessment
PSA Probabilistic Safety Assessment
REMP Radiological Environmental Monitoring Program
RETS Radiological Environmental Technical Specifications
RP Radiation Protection
RWP Radiation Work Permit
SDP Significance Determination Process
SFP Spent Fuel Pool
SSC Structures, Systems, and Components
TI Temporary Instruction
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
VHRA Very High Radiation Area
9 Attachment