ML20062H163
| ML20062H163 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/23/1990 |
| From: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20062H159 | List: |
| References | |
| 50-482-90-32, NUDOCS 9012040133 | |
| Download: ML20062H163 (12) | |
See also: IR 05000482/1990032
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APPENDIX
U.S. NUCLEAR REGULATORY COMMIS$1CN
REGION IV
NRC Inspection Report:
50-482/90-32
Operating License: NPF-42
Docket:
50-482
Licensee: Wolf C*eek Nuclear Operating Corporation (WCNOC)
P.O. Box 411
Burlington, Kansas 66839
Facility Name: Wolf Creek Generating Station (WCGS)
Inspection At: WCGS, Coffey County, Burlington, Kansas
Inspection Conducted:
September 15 through October 24, 1990
Inspectors:
M. E. Skow, Senior Resident Inspector
Project Section D, Division of Reactor Projects
L. L. Gundrum, Resident Inspector
Project Section D, Division of Reactor Projects
Approved:
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A. T. Vowell, Chief, Project Section D
Date
DivisiknofReactorProjects
Inspection Summary
Inspection Conducted September 15 through October 24,1990 (Report 50-482/90-32)
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Areas Inspected:
Routine. unannounced inspection including plant status,
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operational safety verification, surveillance observations, maintenance
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observations', prompt onsite response of an event at operating power reactors,
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review of licensee event reports, and followup on previously identified items.
Results: Within the areas inspected, one noncited violation was identified
(failure to follow procedure).
The noncited violation concerned failure to
properly utilize the industry technical information program (paragraph 8).
The
inspector found a spread of contamination outside of a designated " contaminated
area," which appeared to be an isolated event in the licensee's radiological
controls area (paragraph 3).
Examples of good quality verification were
demonstrated by the identification and resolution of defects found in certain
snubber parts assemblies and in the licensee's response to a pinned spring can
(paragraph 4).
Insufficient human factors considerations contributed to
procedural weaknesses in a surveillance test procedure (paragraph 4).
Control
of troubleshooting and maintenance was good (paragraphs 3, 4, and 5) although
inattention to detail caused control of two auxiliary feedwater valves to shift
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to the auxiliary shutdown panel (paragraph 5). A partial loss of offsite power
caused engineered safety features actuations including an emergency diesel
generator start. All systems functioned properly (paragraph 6).
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DETAILS
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1.
Persons Contacted
Principal Licensee Personnel
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- B. D. Withers, President and CEO
J. A. Bailey, Vice President, Nuclear Operations
- F. T. Rhodes, Vice President, Engineering and Technical Services
- G. D. Boyer, Plant Manager
- R. S. Benedict, Manager, Quality Control (QC)
H. K. Chernoff, Supervisor, Licensing
- M. E. Dingler, Manager, Nuclear Plant Engineering (NPE) Systcms
- R. B. Flannigan, Manager, Nuclear Safety Engineering (NSE)
- C. W. Fowler, Manager, Instrumentation and Control (I&C)
- R. C. Hagan, Manager, Nuclear Services
- R. W. Holloway, Manager, Maintenance and Modifications
- W. M. Lindsay, Manager, Quality Assurance (QA)
- R. L. Logsdon, Manager, Chemistry
- 0. L. Maynard, Manager, Regulatory Services
T. S. Morrill, Manager, Radiation Protection
- D. G. Moseby, Supervisor, Operations
W. B. , Nor'.on, Manager, Technical Support
- C.
E. Parry, Director, Quality
- A. L. Payne, Manager, Supplier Materials Quality
J. M. Pippin, Manager, NPE
- C, M. Sprout, Section Manager, NPE, WCGS
J. D. Weeks, Manager, Operations
- S. G. Wideman, Senior Engineering Specialist
"M. G. Williams, Manager, Plant Support
- J. A. Zell, Manager, Training
The inspectors also contacted other members of the licensee's staff during
the inspection period to discuss identified issues.
- Denotes those personnel in attendance at the exit meeting held on
October 24, 1990.
2.
Plant Status (71707)
The plant operated at 100 percent power during the inspection period,
except for brief reductions to 98 percent power to conduct surveillance
tests.
3.
Operational Safety Verification (71707)
The objectives of this inspection were to ensure that the facility was
being operated safely and in conformance with license and regulatory
requirements to ensure that the licensee's management control systems were
effectively discharging the licensee's responsibilities for cont'nued safe
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operation. The methods used to perform this inspection included direct
observation of activities and equipment, tours of the facility, interviews
and discussions with licensee personnel, independent verification of
safety system status and limiting conditions for operation (LCOs),
corrective actions, and review of facility records.
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Areas reviewed during this inspection included, but were not limited to,
control room activities, routine surveillances, engineered safety
feature (ESF) operability, radiation protection controls, fire protection,
security, plant cleanliness, instrumentation and altrms, deficiency
reports, and corrc cive actions.
Selected inspecto' observations are
discussed below:
a.
3prend of Contamination Beyond a Contamination Boundary
On September 19, 1990, during a routine tour, the inspector found
water stlashing from the "B" centrifugal charging pump (CCP).
The
licersee had shut off the positive displacement charging pump to
repa.r the speed controller and was running the "B" CCP in the
interim.
.eak off past the pump bearing and seal was dripping into a
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temporary p'astic catch device.
However, the device was collapsed
and ineffective at collecting the leak-off. During the time that the
"B" CCP was n,t running, the leak-off, although not contained by the
device, did not spread beyond the roped-off contaminated area of the
pump and its foundation. With the room cooler and pump both in
operation, however, the additional air currents appeared to cause the
leak-off water to splash beyond the roped off area.
The splashing
was small but sufficient to cause a spread of contamination. The
inspector notified health physics (HP) of the potential contamination
and the HP technician found that the floor in the area of the
splashing was contaminated.
HP took appropriate actions to rope off
the area to limit further spread of contamination.
This appeared to
be an isolated event in the licensee's evaluations of and actions
taken to limit the spread of contamination,
b.
Spurious Tripping of Diesel Generator Room Supply Fan "B"
On October 3, 1990, control room personnel were reviewing existing
plant status during shift turnover and noted that there was no
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indication that the diesel generator (DG) Room Supply Fan "B" was
operating.
Personnel were dispatched to verify the problem, and it
was determined that the DG Supply Fan "B" supply breaker had tripped.
The breaker was reset and the fan began running.
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the breaker tripping was not known, however, the DG "B" was declared
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inoperable at the time that the condition of the fan trip was first
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identified at 7:09 a.m.
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Technical Specification (TS) Action Statement 3.8.1.1d.1 requires
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that with one DG inoperable, the required systems, subsystems,
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trains, components, and devices that depend on the remaining DG be
operable or if such conditions cannot be satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, be
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in at least " Hot Standby," within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, followed by
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placing the plant in " Cold Shutdown" within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Diesel Generator "A" was considered to be administrative 1y inoperable
becausa the airconditioning unit that supplies cooling to the
Train "A" DG switchgear was out of service for maintenance.
The
plant safety review committee (PSRC) met at 8 a.m. to discuss the
situation and evaluate the criteria for requesting a temporary waiver
of c.ompliance from TS 3.8.1.1d.1, while DG Room Supply Fan "B" was
troubleshot.
Electrical maintenance personnel continued work on air conditioning
Unit SGK-05A and determined that a problem existed on the thermal
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overloads associated with the molded case circuit breaker. As work
proceeded on both components, licensee management contacted RIV to
request a temporary waiver of compliance.
A conference call with
RIV, Office of Nuclear Reactor Regulation (NRR), and licensee
management was held at approximately 10:30 a.m.
Since the licensee
did not restore DG "A" to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, a shutdown
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of the plant was required.
The licensee planned to initiate reducing
power at 11:09 a.m., 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after entering the 6-hour power
reduction requirement.
This was based on the ability to conduct an
orderly shutdown within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. A temporary waiver of compliance
was not required, however, because the breaker for SGK-05A was
replaced and an operability run was successfully performed.
The troubleshooting efforts on the DG Room Supply Fan "B" continued.
The control circuit was monitored and no further instances of the
breaker tripping occurred. On the basis of previous troubleshooting
(WR 04839-90) results, the nuclear plant engineering group reviewed
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and approved a change to increase the instantaneous trip setting of
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the breaker from 2 to 3.
The PSRC approved the change at 10:50 p.m.
on October 5, and the breaker was declared operable on October 5,
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1990, at 3 p.m.-
Additionally, a modification was being pursued by
the licensee to eliminate the potential for a high starting current,
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which may have tripped the breaker.
Notwithstanding the above, HP controls appeared to be effective during the
. inspection period.
The involvement of HP personnel was evident during the
plant tours and observation of the maintenance activities.
Plant security was found to be generally-professional and in accordance
with the security plan. The inspector observed portions of security
drills conducted during both dayshift and backshift during this inspection
period.
Control of troubleshooting and maintenance activities were good.
4.
Surveillance Observations (61726)
The purpose of this inspection was to ascertain whether surveillance of
safety-significant systems and components was being conducted in
a%ordanca with TS.
Methods used to perform this inspection included
direct observation of licensee activities and review of records.
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The following surveillances were witnessed and/or reviewed by the
inspectors:
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STS AL-201, Revision 9, " Auxiliary Feedwater System Inservice Valve
Test," performed September 28, 1990.
During performance of STS AL-201, the licensee was testing the stroke
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time for the auxiliary feedwater system flow control valve to Steam
Generator"D"(ALHV-5).
Lack of procedural clarity resulted in
initial problems during the test.
First, during the closure of the
valve, the operator was to observe an indicator light on the ESF
status panel being lit. However, the indicator light was lit white
while the valve was open and turned red when it began to close.
The
procedure did not specify which color to observe. The procedure also
specified that the AL HV-5 (a two-speed valve) was to be slow closed
to obtain the required stroke time.
The procedure did not specify
the open speed for the open stroke time. As a result, the operator
opened the valve in slow speed and found that the stroke time was
unacceptable before he realized the intent of the procedure was for
AL HV-5 to be stroked open in fast speed.
The inspector observed the
operators prepare a procedure _ change to address the indicator light
question, and the operators stated that they would verify which open
speed was correct and initiate a procedure change to specify the
correct opening speed of AL HV-5.
STS SF-001, Revision 1, " Control and Shutdown Rod Operability
Verification," performed October 12, 1990.
On October 12, 1990, Procedure STS SF-001, " Control and Shutdown Rod
Operability Verification," was being performed in accordance with
TS 4.1.3.1.2.
This surveillance verifies the operability of each
full-length rod not fully inserted in the core by movement of at
least 10 steps in any one direction. At 7:33 a.m., during the
perfo_rmance of the surveillance test, Control Bank A, Group 1 Rods,
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H-6 and H-10 digital rod position indication (DRPI), did not show
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movement'when rod movement was demanded. No alarms were received.
I&C personnel were sent to investigate the problem under Work
Request (WR) 05061-90 and-the surveillance test was terminated.
At-8:57 a.m., I&C reported that the stationary grippers for rods H-6
and -10 were not releasing.
Since Control Rods H-6 and -10 were
still capable of being tripped because a trip signal would remove
power from the s'tationary grippers,-TS 3.1.3.1, Action 4, was defined
at 9:12 a.m.
Action 4 of TS 3.1.3.1 requires that the inoperable
rods be returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in " Hot
Standby" within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Troubleshooting and repair work
was accomplished and control room personnel performed STS SF-001
satisfactorily. At 5 p.m., TS 3.1.3.1 was exited.
The inspector witnessed the troubleshooting, card replacement, and
surveillance test.
The troubleshooting appeared to be well
organized, preplanned, and performed by experienced personnel. One
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inspector observation noted was the acceptance criteria for voltage
measurements and traces were not specified in the steps developed
prior to performing the troubleshooting but were verified from the
vendor manual that was at the job site and documented in the WR,
Mechanical Snubber Assembly Spare Parts Testing
On September 12, 1990, the 1::ensee identified that three of five
replacement parts for 1/2-inch size mechanical snubber assemblies had
failed 20 percent load tests onsite.
The onsite test is performed to
20 percent of the rated load of 650 pounds. One earlier failure
occurred in 1987.
There were two of these spare part assemblies
installed in the plant and the licensee removed one of the snubbers.
In the other case, previous analysis by the licensee showed that the
snubber was not required and would not affect the piping by either
locking up or by breaking free.
The licensee returned the removed
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part assembly, as well as other representative samples of spares,
from the warehouse and provided them to the vendor for further
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testing. .The results of the vendor tests indicated that the parts
were found to routinely fail below the expected 1500-pound acceptance
point.
The removed part failed at about 125 pounds, while four
others failed at between 145 and 655 ,aunds.
The failure mechanism
appeared to be inadequate crimping of two rods into a cylindrical end
piece.
The rods pulled free of the end piece during tension tests.
The crimp dents were described by the licensee as being shallow
compared to similar parts from another lot. The licensee purchased
the entire lot of 20 assemblies from the vendor.
The licensee also
stated that the vendor, PSA, issued a Part 21 report, but that they
do not believe there are other defective parts in distribution.
Spring Can Hanger Inspection
A spring can hanger was found pinned by the licensee, and was
discussed in NRC Inspection Report 50-482/90-31.
This condition
restricted pipe movernent on one side. The licensee issued a
programmatic deficiency report, PDR OP 90-181. A review of
documentation by the licensee did not reveal the need to have the
spring can hanger pinned. As a result, the licensee was unable to
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determine the root cause of this deficiency.
To assure that other
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hangers did not remain pinned, the licensee conducted a sample
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inspection of 32 of 198 safety related hangers and 80 of
902 nonsafety-related hangers of similar configuration. There were
no pins found installed in the inspected hangers.
Insufficient human factors considerations contributeo to procedural
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weaknesses in Surveillance Test STS AL-201.
Root cause analysis and
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quality. verification were good regarding defects found in certain snubber
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parts and in response to a pinned spring can.
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5.
Maintenance Observations (62703)
The purpose of inspections in this area was to ascertain that maintenance
activities on safety-related systems and components were conducted in
accordance with approved procedures and TS.
Methods used in this
inspection included direct observation, personnel interviews, and records
review.
Portions of selected maintenance activities regarding the WRs
were observed. The WRs and related documents reviewed by the inspectors
are listed below:
Rework check valves in the auxiliary feedwater lines (WR 04333-89).
"B" DG supply fan breaker tripped (WR 04941-90).
On October 8, 1990, troubleshooting was performed on Control Room
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Cabinet RP053BB to determine if there were variations in the voltage
being supplied to the control circuitry for the DG supply fan.
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troubleshooting activity involved the connection of two channels of a
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four-channel recorder to the 115-volt power supply.
Inadvertently,
leads were swapped and the -15-volt supply was connected to a common
terminal which shorted out the power supply. As a result, control of
auxiliary feedwater (AFW) System Valves HV-7 and HV-12 shif ted to the
auxiliary shutdown panel.
This condition was annunciated in the
control room.
The I&C technician removed the leeds immediately and
control room personnel were sent to the auxiliary shutdown panel to
restore the switches to the remote position.
The annunicator cleared
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and the indication showed control being in remote. Although the
event was caused by personnel error, work controls appeared to be
adequate for the circumstances.
The same activity was observed by
the inspector on October 9, 1990, and no problems were noted.
"B" RHR pump breaker inspection and testing (WR 5196-90).
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Replace pipe insulation to condensate storage tank (WR 07062-90).
Install valve for isolating BB-PCV-8034 (WR 02930-90).
Control Rod Group 1 did not move (WR 05061-90).
With the exception of inattention to detail by a technician, which caused
control of auxiliary feedwater valves to transfer to the auxiliary
shutdown panel, control of troubleshooting and maintenance was good.
6.
Prompt Onsite Response to Events at Operating Power Reactors (93702)
At 1:33 a.m., CDT, on October 25, 1990, the plant experienced a partial
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Four breakers in the switchyard opened
automatically to isolate the east bus. The power supply from the east bus
is used as the normal supply to the "A" train emergency bus, NB01. Upon
loss of power to NB01, the "A" emergency diesel tenerator (EDG)
automatically started and closed in on NB01 in 8 seconds. The shutdown
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sequencer functioned properly to restore power to safety loads.
operators also manually started the "B" essential service water (ESW) pump
in accordance with procedures. The loss of "A" train electric power
causes the service water isolation valves to ESW to f ail closed for both
Operators then established an alternate lineup to power NB01
ESW trains.
from another independent offsite power source at 4:25 a.m. and restored
safety systems to their normal lineup. The plant remained at 100 percent
power throughout the event, supplying power through the west bus.
Electricians from system operations initially found a Phase "A" fault in
Breaker 345-90, one of three main breakers on the east bus.
Breakers 345-60 and -120 opened to isolate the east bus and Breaker 345-80
The operators and
opened to isolate 345-90 from the west bus.
electricians next worked to open the disconnect links to isolate the
faulted Breaker 345-90 and restored power to the east bus through 345-60
This in turn enabled the operators to restore Bus NB01 to the
and -120.
normal lineup at 6:15 a.m.
During the event, additional actuations occurred to the control roo
containment purge isolation systam (CPIS). These actuations occurred
The licensee
because power was lost to the assv: dated radiation monitors.
made a 4-hour nonemergency report to the NRC as required by 10 CFR Part 50.72(b)(2)(11) because the auto atic actuation of the ESF
systems.
The inspectors will review the LER for the event and will report any
findings in a subsequent inspection repcrt.
7.
Review of Licensee Event Reports (LERs) (92700)
During this inspection period, the inspectors performed folls
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WCGS LERs. The LERs were reviewed to ensure that:
Corrective action stated in the report has been properly completed or
work is in progress;
Response to the event was adequate;
Response to the event met license conditions, commitments, or other
applicable regulatory requirements;
The-information contained in the report satisfied applicable
reporting requirements; and
Generic issues were identified.
The LERs discussed below were reviewed and closed:
482/90-007, "Feedwater Isolation Signal Because of
(Closed) LER
Surveillance Test Requiring a High Feed Rate and Auxiliary Feedwater
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Actuation and Reactor Trip Signal During Recovery."
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During the performance of $fS AL-211, Revision 3, " Turbine Driven
Auxiliary Feedwater System Flow Path Verification and Inservice Check
Valve Test," the levels in Steam Generators A and C reached the
high-high level. A feedwater isolation signal and main turbine trip
signal occurred.
The cause of the event was procedural. inadequacy
that led the operators to believe that the four check valves (one per
each steam generator) were to be tested simultaneously. This required
feeding 140,000 pounds-mass per hour which was higher than steam flow
for the existing plant conditions.
Procedure STS AL-211 was revised
on June 28, 1990, to clarify that only one check valve be tested at a
time. Operations personnel have been reminded of their authority to
suspend testing if necessary to preclude plant transients. On the
basis of these corrective actions, this LER is closed,
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b.
(Closed) LER 482/90-003, " Control Room Ventilation Isolation and
Containment Purge Isolation Caused by High Gaseous Activity During
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Containment Vent."
On March 20, 1990, shortly after initiation of a containment vent,
CPIS and CRVIS isolation signals were received.
The probable cause of
the high activity was a vent line installed from the pressurizer
relief tank (PRT) to the containment shutdown purge ductwork that had
been left open while the purge was not in progress.
This allowed an
accumulation of higher activity air to build up in the duct. When
the purge was started, the concentration of radioactivity in the air
being exhausted was sufficient to generate the isolation signals.
Caution statements have been added to Procedure GEN 00-007,
Revision 13, "RCS Drain Down," dated July 26, 1990. These cautions
state that when venting to the purge ductwork from the PRT or vessel
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head, the duration of time when the purge is shut down should be
minimized.
These changes to the procedure should prevent farther
events of this type.
This LER is closed.
8.
Followup on Previously Identified NRC Items (92701 and 92702)
a.
(Closed) Unresolved Item 482/8929-02, Resolve Inadequacy of
Containment Cooler Supports. This item concerned the delay from the
time the licensee became aware of a question concerning the seismic
qualification of the containment cooler supports and when the issue
was resolved.
The licensee initiated Programmatic Deficiency
Report (PDR) NP 90-01, on November 18, 1989, after determining that
personnel had failed to follow KGP-1311, Revision 2, " Industry
Technical Information Program," that required placing applia b?e
incoming vendor technical information in industry technical
information program (ITIP). The licensee performed a review of r,ther
problem information requests (PIRs) provided by Bechtel and found
that there were other PIRs that were not entered into the licensee's
ITIP.
Those items were subsequently entered in the ITIP for
tracking.
The failure to follow KGP-1311 is an apparent violation of
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TS 6.8.1.a (482/9032-01).
Because the criteria specified in
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Section V.A of the Enforcement Policy was satisfied, a cited violation
will not be issued. The unresolved item and the noncited violation
are closed.
b.
(Closed) Violation 4 2/8905-03, Failure to Update Drawing. This
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item concerned a drawing in the control room that had not been
redlined to reflect field implementation and partial closeout of a
plant modification.
The licensee revised Procedure ADM 01-042,
Revision 16. " Plant Modification Request Implementation," to require
that drawings be marked with red lines to indicate changes before
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affected systems are returned to service.
This includes
modifications that are completely installed as well as partially
installed.
In addition, ADM 01-228, Revision 2, " Temporary
Modifications," requires that control room drawings be marked with a
green line to reflect installed temporary modifications.
This item
is closed,
c.
(Closed) Violation 482/8908-01, Inadequate Procedures.
This item
concerned failure to include consideration of a potential release of
radioactivi contamination in a fire h uards analysis.
The licensee
performed an analysis that found that the dose rate of the material
in the area of consideration would have needed to be greater than
1000 mrem /hr to exceed 10 CFR Parts 20 and -100 limits.
The licensee
performed surseys of the areas and found the highest dose rate
measured was 15 mrem /hr.
Further, the inspector reviewed Procedure
KPN-D-316, Revision 3, " Fire Protection Review," and associated Forms
KEF-D-316-1 and -2.
The procedure requires the use of the checklists
in the forms during fire protection review applicability screening.
Form KEF-D-316-2, Step 3.6, questions an increase ir. the amount of
radioactive material that can be released in the event of a fire.
This item is closed.
d.
(Closed) Violation 482/8927-01, TS Surveillance Requirement
Improperly Performed.
This item concerned performance of a
surveillance test of service water flow through the containment
coolers.
Since the violation was issued, TS Surveillance
Requirement 4.6.2.3.a.2 has been changed to verify a flow rate
greater than 1,850 gallons per minute (gpm) rather than the previous
minimum of 2,200 gpm.
Procedure STS EF-925, Revision 3, " Containment
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Coolers Flow Verification," requires the new flow rate minimum.
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addition, Precaution 2.2 states that system lineups other than those
specified by the pro:edure would require an evaluation to determine
the condition of the containment coolers.
This item is closed.
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e.
(Closed) Violation-482/9026-02, Failure to Follow Procedure.
Post-trip reviews did not include strip charts that reflected real
time, and Section 6 of the review was not completed.
SeP ion 6
documents the actual or suspected cause of the trip and any abnormal
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or degraded indications identified during the transient.
These were
required by Procedure ADM 02-400, Revision 7, " Post-trip Reviews."
However,_it was noted that Revision 7 of the procedure also lacked
clarity.
Revision 8 to the procedure, issued October 4, 1990,
clarified both items of concern.
In addition, a memo was issued
September 11, 1990, to all shift supervisors discussing the ADM 02-400
changes and asking them to fill out a training attendance form
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documenting that the information had been discussed with each crew.
All licensed personnel in the trcining department were required to
complete the training attendance form. On the basis of these
corrective actions, this violation is considered closed.
9. -
Exit Meeting (30703)'
The inspectors met with licensee personnel (denoted in paragraph 1) on
October 24, 1990. The inspectors summarized the scope and findings of the
inspection.
The licensee did not identify as proprietary any of the-
information provided to, or reviewed by, the: inspectors.
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