ML20045A821
| ML20045A821 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/03/1993 |
| From: | Powers D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20045A814 | List: |
| References | |
| 50-298-93-18, NUDOCS 9306150032 | |
| Download: ML20045A821 (14) | |
See also: IR 05000298/1993018
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-298/93-18
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Operating License: DRP-46
Licensee:
Nebraska Public Power District (NPPD)
Facility Name:
Cooper Nuclear Station (CNS)
Inspection At:
Inspection Conducted: April 26-30 and May 10-14, 1993
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Inspector:
C. E. Johnson, Reactor Inspector, Maintenance Section
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Division of Reactor Safety
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Approyed:
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d/3/98
Dr. Dale A. Powers, Chief, Maintenance Section
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Division of Reactor Safety
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Inspection Summar.y
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Areas Inspected:
Routine, announced inspection of onsite followup of licensee
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event reports and other followup.
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Results:
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A violation was identified pertaining to inadequate corrective
action and root cause determination taken in response to
performance problems with the control room ventilation radiation
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monitor (Section 2.4).
A violation was identified pertaining to an inadequate
10 CFR 50.59 safety evaluation of Design Modification DC 90-0226,
which contained an unreviewed safety question (Section 2.4).
The licensee's corrective actions pertaining to four licensee
event reports and unresolved items were adequate and met
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regulatory requirements, license conditions, and commitments.
However, the licensee's corrective actions and safety evaluation
of changes to radiation monitors were inadequate (Section 4.0).
Summary of Inspection Findinos:
Licensee Event Report 298/92-002 was closed (Section 2.1).
9306150032 930609
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Licensee Event Report 298/93-005'was closed (Section 2.2).
Licensee Event Report 298/93-006 was reviewed but not closed
(Section 2.3).
Licensee Event Report 298/91-020 was closed (Section 2.4).
Violations 298/9318-01 and 298/9318-02 were opened (Section 2.4).
Unresolved Item 298/9223-01 was closed (Section 3.1).
The licensee made a commitment to revise the Standard Clearance
Order to incorporate the use of the newly installed vents and
drains for the local leak rate testing of the reactor feedwater
check valves (Section 3.1).
Attachments:
Attachment 1 - Persons Contacted and Exit Meeting
Attachment 2 - Documents Reviewed
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DETAILS
1 PLANT STATUS
At the beginning of this inspection period, the plant was in day 52'of an 82
day scheduled refueling outage. The reactor was defueled and the reactor
cavity was flooded up with the fuel pool gates installed.
2 ONSITE REVIEW 0F LICENSEE EVENT REPORTS (92700)
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The purpose of this inspection was to determine through onsite followup of
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selected event reports, whether the licensee has taken corrective actions as
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stated in written reports of the events and if responses to the events were
adequate and met regulatory requirements, licensee conditions, and
commitments.
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2.1
(Closed) Licensee Event Report 298/92-002:
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Setooint Inaccuracy Resultino from Reference Leo Temperature Effects that
had not been Correctly Addressed
On January 20, 1992, an advance copy of Supplement 2 to General Electric
Service Information Letter 299 was received by the licensee. The purpose of
the supplement was to notify boiling water reactor owners that the information
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in Service Information Letter 299, dated July 25, 1979, had been
misinterpreted by one utility and was potentially subject to misinterpretation
by others.
A clarification of the information was provided to the licensee,
along with a recommendation that a check of level instrument setpoint
calculations be conducted.
An evaluation was performed which determined that although the Service
Information Letter had been reviewed and properly considered when it was
originally received, incorrect initial conditions and incomplete calculations
for the Reactor Water Level 1 setpoint, prescribed in the Technical
Specifications, resulted in a non-conservative setpoint performed in 1981.
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The inspector reviewed the licensee's immediate corrective actions to adjust
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the setpoint within the required limit. Temporary Procedure Change
Notice 92-005 contained corrections to the calibration and functional test
procedure.
The licensee implemented the change immediately.
The licensee had
proceduralized the calculational methodology and provided for independent
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design basis reviews.
The licensee's immediate and long-term corrective actions were determined to
be appropriate.
This licensee event report is closed.
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2.2
(Closed) Licensee Event Report 298/93-005:
Excessive Primary Containment
Leakaae Discovered durina Local Leak Rate Testina of Reactor Feedwater
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On March 10, 1993, upon completion of the local leak rate testing of the
reactor feedwater check valves, the as-found primary containment leak rate was
determined to exceed the acceptance criterion. The leakage of the reactor
feedwater check valves was too great to be accurately quantified using the
available test equipment.
A review by the licensee of the as-found leakage
rate of other components tested during this outage indicated that, had the
reactor feedwater check valves performed acceptably, the total as-found
leakage would not have exceeded the allowable containment leak rate.
This licensee event report is associated with Unresolved Item 298/9223-01
pertaining the reactor feedwater check valve leak rate and the valve and
system modifications, which have been completed.
Details of the modifications
are discussed in Section 3.1.
Local leak rate tests have been successfully
completed. The licensee committed in the exit meeting to updating the
Standard Clearance Order to incorporate the use of the new vents and drains as
part of the process for performing the local leak rate test.
Based on completion of the local leak rate tests, acceptable system
performance, and the commitment made by the licensee, this licensee event
report is closed.
2.3
(00en) Licensee Event Report 298/93-006:
Fire Barrier Doors Discovered
Open and Obstructed without a Continuous Fire Watch due to Personnel
Error
On March 16, 1993, at 11 p.m., two fire barrier doors were found open and
obstructed with no fire watch assigned as required by Technical
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Specifications.
The doors provided for passage from a stairwell into the high
pressure coolant injection pump room and residual heat removal pump room. The
door to the high pressure coolant injection pump room was obstructed with a
stanchion, while the door to the residual heat removal pump room was
obstructed with a 12-inch temporary ventilation duct that was connected to a
filter unit in the stairwell.
On April 10, 1993, at approximately 8:20 p.m., a fire barrier door to the
service water pump room at the intake structure was found open and obstructed.
A fire watch had been posted in the area continuously since April 4,1993,
because of work activities in progress.
Upon completion of work on April 10,
1993, the halon system was restored, but the integrity of the fire barrier
door was not re-established prior to departure of the fire watch at 5:20 p.m.
Review of the licensee's corrective action indicated that immediate corrective
actions were taken by removing obstructions and closing the doors. The
problems were reviewed by the outage directors at the outage coordination
meetings conducted at the beginning of each shift, and compliance with station
fire protection requirements were stressed.
Personnel involved were
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counseled. Additionally, a rovi'ng fire door patrol was initiated as a means
to assure that fire doors were closed or that a fire watch was stationed.
The inspector informed the licensee that it was not clear that'this issue was
stressed to other workers who might have the opportunity to place obstructions
in fire barrier doorways.
The inspector was informed that the licensee
planned on getting this message out to all personnel by memoranda or some
other means. The licensee was also in the process of reviewing the fire watch
implementation process to identify enhancements that will prevent recurrence.
This licensee event report will remain open pending NRC's review of the
licensee's completion of the aforementioned corrective actions.
2.4 (Closed) Licensee Event Report 298/91-020:
Failure of the Primary
Containment Intearated Leak Rate Test due to Drywell Ventilation Monitor
System and Containment Penetration leakaae
On December 10, 1991, at 12:35 a.m., during the performance of the primary
containment integrated leak rate test, the drywell ventilation monitor gaseous
detector mounting bolts were stripped from the lead shield to which_they were
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mounted.
This resulted in the detector being ejected from the lead shield
chamber.
Primary containment pressure was approximately 51.6 psig and was
being raised to the integrated leak rate test pressure of 58 psig.
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addition to this leak there were others, which were in the reactor water
cleanup system and the reactor feedwater system.
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The leakage from the drywell ventilation radiation monitor was quantified.
The monitor was then isolated and pressurization of the primary containment
continued to the required test pressure.
The drywell ventilation radiation monitor, Model I-RAK-225-lF, manufactured by
Nuclear Measurements Corporation, was designed, fabricated, and installed as
part of the original plant equipment. Over the years, numerous detector
calibration and maintenance efforts have required removal of the detector from
the shield chamber, which is primarily constructed of lead. This frequent
disassembly and reassembly resulted in degrading the shield chamber threaded
connections into which the detector was bolted.
When the unit was pressurized
during performance of the integrated leak rate test, the threaded engagement
of the retaining bolts in the shield chamber failed, resulting in ejection of
the detector.
On October 30, 1992, the licensee issued Design Change 90-0226, which replaced
the existing drywell ventilation monitor (RMV-RM-4) and the control room
ventilation radiation monitor (RMV-RM-1) during a scheduled refueling
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outage (RF15).
The licensee informed the inspector that there were many
problems over the years with these particular two monitors.
The inspector
requested from the licensee, Design Change 90-0226 and the history of the
monitors for review.
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The past and recent maintenance history revealed that over 50 nonconformance
reports were written against RMV-RM-1 and -4 (29 and 35, respectively). This
appeared to be an excessive amount of nonconformance reports for the two
monitors. The inspector inquired of the licensee what corrective actions had
been taken to resolve these problems, and if any root causes were ever
determined. The licensee stated that there has been no single root cause to
the problems, and that the monitors were obsolete and needed replacing.
The inspector reviewed a sample of these nonconformance reports. Of the
nonconformance reports reviewed by the inspector, several indicated that
different problems could have contributed to some of the failures,
i.e.,
vibration or high voltage.
However, it was not evident that the licensee
pursued either of these possibilities to determine if indeed these monitors
may have had inadequate supports or bracing to preclude vibration or whether
high voltage could have caused the monitors to be out of calibration. There
appeared to be no adequate corrective actions or root causes determined in the
nonconformance reports reviewed to prevent repetitive occurrences. Corrective
action that was taken was to replace or repair subcomponents.
Discussions on the current nonconformance reports with the licensee revealed
that these monitors were to be replaced during the outage in 1993; therefore,
the licensee's intent to repair or replace subcomponents was very limited.
The inspector informed the licensee that with no root cause determination,
there is no assurance that the new radiation monitors will resolve the
recurring problems.
Two specific examples of problems with the Control Room Ventilation Radiation
Monitor RHV-RM-1 are as follows:
(1)
Inadequate procedural guidance resulted in unintended high voltage
being admitted into the monitor during maintenance activities.
There were approximately four nonconformance reports written
pertaining to this issue on RMV-RM-1.
For instance,
Nonconformance Report 87-112 was written where failure was
attributed to procedure error, in that high voltage was not
reduced prior to the disconnection of the detector assembly of
RMV-RM-1. The vendor recommended maintenance practices required
that high voltage be turned off prior to disconnection and
reconnection of the detector assembly. The licensee's corrective
action was to revise certain procedures.
One year later,
Nonconformance Reports89-061, 89-156, and 89-157 were written for
procedure errors also relating to high voltage.
However, a
different procedure was involved for these problems.
It was
apparent that the scope of corrective measures did not capture all
applicable procedures relating to this error; therefore, recurrent
problems continued.
(2)
On January 29, 1993, Deficiency Report 93-018 was issued after the
control room ventilation radiation monitor was found by Chemistry
personnel to be malfunctioning.
In particular, the particulate
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channel's filter paper drive unit was not advancing. . Maintenance
Work Request 93-0347 was generated to investigate and repair the
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unit.
Subsequently, the filter paper drive unit was repaired.
The inspector learned, however, that the maintenance history on
this unit contained Nonconformance Report 90-125 that had also
been written against the filter paper drive unit due to a failed
belt. This earlier nonconformance report was voided after it was
determined that failure of the filter paper drive unit did not
render the monitor / channel inoperable. Consequently, Maintenance
Work Request 90-3578 was initiated to replace the belt.
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Engineering evaluated this event and concluded that a failure rate
of once every 9 years was not enough to warrant the addition of a
preventive maintenance item.
Later, in June 1991, Maintenance
Work Request 91-1868 was generated against this monitor / channel
because of another failure.
This maintenance work request also
found it necessary to replace the belt. The second failure of the
belt should have indicated to the licensee that the filter paper
drive unit could have possibly been defective causing the belt to
fail; therefore, the entire unit should have been replaced. These
examples indicate inadequate corrective action and root cause
determination to prevent recurrent problems.
The inspector concluded that corrective action and root cause determination of
RMV-RM-1 performance problems was inadequate. This is a violation of 10 CFR 50, Appendix B, Criterion XVI (50-298/9318-01).
Further review of Design Change 90-0226 by the inspector indicated that
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RMV-RM-1 and -4 were to be replaced by newer models. The inspector reviewed
the data sheets of the new monitors and determined that the new monitors were
rated at a design pressure of only 2 psig. The old monitors, specifically
RMV-RM-4 was rated at a design pressure of 58 psig, which would withstand the
design basis accident pressure.
The inspector reviewed the licensee's justification for the modification
change that included the following:
The existing control room and drywell ventilation radiation
monitors had become high maintenance items and were obsolete.
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Replacement would reduce equipment down time and maintenance cost.
The modifications to Penetration X-51 would make it possible to
protect the drywell ventilation radiation monitor from over
pressurization during the integrated leak rate test without
isolating the post accident sampling (PAS) system during the
Normally open isolation valves PC-V-43
and PC-V-229 would be required to be closed prior to the
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The inspector questioned the licensee's justification on whether this
modification considered the likelihood of a containment breach through
RMV-RM-4 during a design basis accident. The licensee told the inspector that
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RMV-RM-4 was not needed during a design basis accident because of the
following:
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(1)
There were two essential high radiation monitors installed that
would be utilized during design basis accident conditions, and
(2)
There are two manual isolation valves installed that would be
isolated during the integrated leak rate test to protect the
radiation monitor.
The licensee informed the inspector that additional justification for
containment breach was derived from General Electric Owners' Group Evaluation
of Containment Isolation Concerns, NEDC-22253, 82NED114 Class II, October
1982, Reference 9.0.1.12, which stated that potential offsite exposure from a
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break of a 1-inch inside diameter line was substantially below the 10 CFR 100
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offsite radiological limits.
The inspector reviewed the licensee's Station Safety Evaluation of this
modification change which included a 10 CFR 50.59 analysis.
Review of this
evaluation indicated that the licensee concluded the following:
The proposed change would not affect the safety function of any
system because the new RMV-RM-1 and -4 monitors met seismic
requirements.
This new design would not increase the chances for a release of
radioactivity.
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The new radiation monitors for the control room (RMV-RM-1) and
drywell (RMV-RM-4) performed the same Updated Safety Analysis
Report functions as the previous monitors.
The modification made for this design change would not create a
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possibility for an accident or malfunction of a different type
than any previously evaluated in the Updated Safety Analysis
Report.
The licensee also concluded in the 10 CFR 50.59 analysis that there were no
unreviewed safety questions that existed for this modification.
The inspector informed the licensee that his review of Design Change 90-0226
determined that Engineering did not perform an adequate 10 CFR 50.59 review,
in that, there was an unreviewed safety concern involved with the proposed
change.
This unreviewed safety concern pertained to the new radiation
monitor RMV-RM-4 design pressure rating of 2 psig, which was far below the
design basis accident pressure of 58 psig.
It appeared to the inspector that
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operation of the new radiation monitor at the time of a design basis accident
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would:
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Breach primary containment;
Subject personnel inside secondary containment to high radiation
levels as they exited (the monitor is located at a personnel
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access / egress point);
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Delay reentry into secondary containment as a result of
radioactive contamination; and
Increase the likelihood of radiation exposure to the general
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public.
The licensee informed the inspector that this concern was previously
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identified by them earlier in correspondence QAC93168 and QAC93169 dated
April 21 and April 22, 1993, respectively.
These documents addressed a
concern with the integrated leak rate test; however, they did not address the
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significant safety concern which was a breach of primary containment during a
design basis accident. The inspector questioned the licensee why no action
was ever taken to resolve their concern, especially when Design Change 90-0226
had been approved by nine groups (e.g., Engineering, Quality Assurance,
Standard Operation Review Committee). No conclusive answer was.given. Also,
no deficiency report or nonconformance report was written prior to the end of
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this inspection.
Discussions with . licensee design personnel indicated that there was no formal
hold on the field installation of this modification, and that final
configuration of this modification had not been decided. At the end of this
inspection, the inspector was informed that the licensee planned to use dual
automatic isolation valves that will be powered from a Class IE electrical
bus. This modification was in review.
Field observation by the inspector determined that the new RMV-RM-4 had been
installed in parallel with the old radiation monitor.
This change was
initiated by On-The-Spot Change No. 5.
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The inspector concluded that the licensee did not adequately evaluate this
modification, in that an unreviewed safety question existed and that the
modification was approved by the responsible organizations and installed in
the field without NRC approval.
This issue of proposing a design change that involved an unreviewed safety
question is a violation of 10 CFR 50.59 requirements (298/9318-02).
This violation does not oeet the criteria of a non-cited violation in that the
licensee identified a concern with emphasis pertaining to meeting Appendix J
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testing criteria to be close to the "as is" condition as practical. Whereas,
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a significant safety issue pertained to a breach of containment during a
design basis accident through a radiation monitor not designed to withstand
design basis accident pressure.
This licensee event report is closed.
3 FOLLOWUP (92701)
The purpose of this inspection was to perform an onsite followup inspection of
an unresolved item.
3.1
(Closed) Unresolved item 298/9223-01:
Timeliness and Effectiveness of
Corrective Actions Associated with local Leak Rate Test Failures of
Individual Feedwater Valves and Effects on Primary Containment System
Inteority
NRC Inspection Report 50-298/92-23 documented that there had been a number of
local leak rate test failures involving four feedwater system check valves
(RF-CV-13V, -140V, -15CV, and -16CV). These failures contributed to the
primary containment system exceeding its allowable leakage rate (.6 La).
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occurrence had resulted in the issuance of a nonconformance report and a
licensee event report, which documented the failures.
Each occurrence had
caused initiation of corrective actions which, by themselves, did not appear
to be effective.
Engineering provided a final written report on September 25, 1992, ir which
the results of a survey of check valves at other boiling water reactor
facilities were discussed.
Based on the survey results and a review of all
maintenance history applicable to the feedwater check valves, the following
recommendations were initiated:
A change to the feedwater check valve preventive maintenance,
which would require notification to the system engineer prior to
performing preventive maintenance to allow for a detailed
examination and evaluation of all internal components of the check
valves.
An engineering work request to install high point vents and low
point drains to provide adequate vent and drain paths during
system draining.
An engineering work request to remove the soft-seat rings, which
were not part of the original design. Also included in the plan,
was disc / seat / hinge pin bore alignment.
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which the required alignment and necessary machining was performed
in-situ and in one setup to provide the ultimate alignment and
disc / seat fitup (considered to be crucial for achieving acceptable
local leak rate test results).
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The inspector reviewed the corrective actions proposed by the licensee
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-documented in Engineering Work Request Nos.92-125 and 92-126 and Maintenance.
Work Requests Nos. 92-2719, 92-2720, 92-2721, 92-2722, 93-1097, and 93-1159.
Review of the. maintenance work requests indicated that the modification on the
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reactor feedwater valves had been completed with the exception of. the. soft
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seats, which were not replaced during this outage because-of a-six-month wait
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following purchase order issuance. The license planned to evaluate the soft
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seats at the next outage to determine if hard seats are actually necessary.
The licensee informed the inspector that use of a hard seat requires a
different disk which has to be' ordered at least 6 months in advance. After
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the completion of the. aforementioned repairs, the valves were tested.
During the inspection discussed in NRC Inspection Report 50-298/93-12, there
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were discussions on the method by which to test the valves.
The license
indicated that they would test the reactor feedwater valves by the following
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two methods:
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Fill the reactor feedwater system, secure the system,' drain down
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using the historical method, and perform the local leak rate. test.
Fill the reactor feedwater system, secure the system, drain down
using newly installed vents and drains, and then perform the local
leak rate test.
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This followup inspection determined that the licensee had conducted the two
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tests on these valves. The licensee had determined that use of the-historical.
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method of draining down had resulted in unacceptable check valve leakage;
whereas, the use of the newly installed vents and drains resulted in
acceptable check valve performance.
It is, therefore, believed that the
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licensee's long-standing problem with reactor feedwater check valve leakage
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arose when the licensee converted from the use of. water to air as the fluid
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for testing check valves. Apparently, when' draining down the system, a vacuum
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was pulled that unseated. check valves, and air pressure was not capable of re-
seating the check valve discs.
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The licensee committed to updating the Standard Clearance Order to incorporate
the new vents and drains as part of the process for performing the local leak
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rate test.
Based on completion of the local leak rate tests, acceptable
system performance, and the commitment made by the licensee in the exit
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meeting, this unresolved issue is closed.
4 CONCLUSION
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The licensee's corrective actions pertaining to four licensee event reports
and unresolved items were adequate and met regulatory requirements, license
conditions, and commitments. . However, in regard to a safety-related control
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room ventilation radiation monitor, the licensee's corrective actions and root
cause-determination to preclude repetitive problems were inadequate. Also,
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the licensee's 10 CFR 50.59 review of a design change modification to a
containment drywell ventilation radiation monitor was inappropriate.
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ATTACHMENT 1
1 PERSONS CONTACTED
1.1
Licensee Personnel
- L. Bray, Regulatory Compliance Specialist
D. Buman, Lead Electric Engineer
- M. Dean, Supervisor, Nuclear Licensing and Safety
K. Done, Supervisor, Mechanical
J. Dykstra, System Engineer
- J. Flaherty, Manager, Engineering
- R. Gardner, Plant Manager
- S. Gayler, Administrative Secretary
B. McClillen, Lead Instrumentation and Controls Engineer
S. McClure, Manager, Nuclear Engineering
- J. Meacham, Site Manager
- C. Moeller, Supervisor, Technical Staff
- S. Peterson, Senior Manager, Operations
M. Siedlik, Supervisor, Civil and Structural
- G. Smith, Manager, Quality Assurance
- M. Unruh, Manager, Maintenance
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- W. Wenzl, Manager, Site Nuclear Engineering Department
B. Wilcox, Lead Mechanical Engineer
1.2 Stone & Webster Enaineerina Corporation
R. Sanchez, Senior Principal Mechanical Engineer
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1.3 NRC Personnel
- R. Kopriva, Senior Resident Inspector
- D. Powers, Section Chief, Maintenance Section, Division of Reactor Safety
- W. Walker, Resident Inspector
In addition to the personnel listed above, the inspector contacted other
personnel during this inspection period.
- Denotes personnel that attended the exit meeting.
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2 EXIT MEETING
An interim exit meeting was conducted on April 30, 1993, and an exit meeting
was conducted on May 14, 1993.
During these meetings, the inspector
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summarized the scope and findings of the inspections. The licensee did not
identify as proprietary, any information provided to, or reviewed by the
inspector.
During the May 14, 1993, exit meeting, Mr. J. Heacham, Site Manager, made a
commitment to revise the Standard Clearance Order to incorporate the use of
the newly installed vents and drains for the local leak rate testing of the
reactor feedwater check valves.
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ATTACHMENT 2
DOCUMENTS REVIEWED
DESIGN CHANGE (DC)
DC 90-0226
MINOR DESIGN CHANGE (MDC)
MDC 80-55
MAINTENANCE WORK RE0 VESTS (MWR)
MWR 93-1159
MWR 92-2721
MWR 92-2722
MWR 92-2720
MWR 92-2719
MWR 93-1097
OVALITY ASSURANCE LETTERS
QAC93168, dated April 21, 1993
QAC93169, dated April 22, 1993
EQUIPMENT OVALIFICATION DATA PACKAGES (HIGH RADIATION MONITORS)
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NO. 225
NO. 225, APPENDIX A
NO. 227
REGULATORY DOCUMENTS
NUREG-0800, Standard Review Plan, " Containment Isolation System"
NUREG-0737, TMI Action Plan
SAFETY GUIDE II, " Instrument lines Penetrating Primary Reactor
Containment," dated March 10, 1971
PROCEDURES
Surveillance Procedure 6.2.2.1.1, "CSCS Water Level Calibration
And Functional Test," Revision 23
Maintenance Procedure 7.0.8, " Pressure Testing," Revision 7
Engineering Procedure 3.26, " Instrument Setpoint Control,"
Revision 16
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General Operating Procedure 2.1.1, "Startup Procedure,"
Revision 62
CNS Procedure 0.5.1, "Nonconformance And Corrective Action,"
Revision 9
ENGINEERING WORK RE0 VESTS (EWR)
EWR 92-079
EWR 92-126
EWR 92-125
NONCONFORMANCE REPORTS (NCR)
NCR 90-130
NCR 90-018
NCR 90-028
NCR 90-110
NCR 91-014
NCR 91-092
NCR 91-123
NCR 92-118
NCR 92-058
NCR 93-028
OTHER ITEMS REVIEWED
Integrated leak rate test results (containment)
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Local leak rate test results (reactor feedwater check valves)
Procurement records (high radiation, control room and drywell
monitors)
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