IR 05000373/1993007
| ML20035B731 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/25/1993 |
| From: | Hague R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20035B722 | List: |
| References | |
| 50-373-93-07, 50-373-93-7, 50-374-93-07, 50-374-93-7, NUDOCS 9304050046 | |
| Download: ML20035B731 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No.
50-373/93007(DRP); 50-374/93007(DRP)
Docket Nos.
50-373; 50-374 License Nos. NPF-ll; NPF-18 Licensee:
Commonwealth Edison Company Executive Towers West III 1400 Opus Place Suite 300 Downers Grove, IL 60515 Facility Name:
LaSalle County Station, Units 1 and 2 Inspection At:
LaSalle Site, Marseilles, Illinois Inspection Conducted: February 8 through March 15, 1993 Inspectors:
D. Hills C. Phillips J. Roman, Illinois Department of Nuclear Safety
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Approved By:
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-K. L(Hagut, Chief Date Reactor Projects Section 1C e
Inspection Summary Inspection from Februarv 8 throuah March 15. 1993 (Renorts No. 50-373/93007
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Areas inspected: A routine, unanncunced safety inspection was conducted by the resident inspectors and an Illinois Department of Nuclear Safety
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inspector.
The inspection included followup on previously identified items
and licensee event reports; review of operational safety, monthly maintenance,
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and surveillance activities; safety assessment and ' quality verification; and l
report review.
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Results:
Two cited violations we e identified concerning the following:
l Several examples of maintenance personnel failure to follow
procedure due to inattention to detail (paragraph 5).
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Several examples of radiation protection personnel failing
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to follow radiation protection procedures during daily
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checks of continuous airborne monitors (paragraph 4.b(1)).
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9304050046 930329 PDR -ADOCK 05000373.
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One non-cited violation was identified concerning an inadequate safety evaluation in regard to fire protection issues (paragraph 4.a(2)).
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One unresolved item was identified regarding conformance of the i
onsite nuclear safety group organization changes to the plant
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licensing basis (paragraph 7).
One open item was identified involving review of seismic
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qualification test records for scram relay connectors
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(paragraph 6).
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Plant Operations l
Performance remained steady in this area.
Corrective actions to address
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inattention to detail including a plant stand down were effective for
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operations personnel.
Operator response to a failure of a scram relay to-
reset was good.
Maintenance / Surveillance j
i Performance remained steady in this area. Corrective actions to address l
inattention to detail were not entirely effective for maintenance personnel.
l Several maintenance errors were noted with the most notable causing. a safety
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relief valve blowdown event. Two failures to correctly connect the same local i
power range monitor also represented failures to adequately perform the prescribed post maintenance testing.
Enoineerino/ Technical Support
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Performance remained steady in this area. Safety evaluations performed for a j
temporary system change were inadequate as they did not recognize changes to j
the safety analysis report and therefore did not evaluate whether they e
involved an unreviewed safety question. A system engineer was not assigned to
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maintain overview of a majority of the plant continuous airborne monitors.
i Padiolooical Controls
Performance remained steady in this area. Material condition of the plant
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continuous airborne monitors was poor. Several examples were also -identified
in which radiation protection surveillances required by plant procedures were
not performed on these monitors or upon identifying equipment problems the l
information was not properly dispositioned.
Insufficient management emphasis I
on this equipment was clearly indicated. Housekeeping was good during the i
inspection period.
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Safety Assessment /0uality Verification
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Performance remained steady in this area. The current onsite nuclear safety
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group organization did not appear to meet the plant licensing basis for the independent safety engineering group.
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DETAILS i
1.
Persons Contacted W. Murphy, Site Vice President G. Spedl Station Manager
- J. Gieseker, Site Engineering and Construction Manager J. Schmeltz, Operations Manager C. Sargent, Support Services Director
- H. Reed, Technical Services Superintendent
- J. Lockwood, Regulatory Assurance Supervisor
- M. Santic, Maintenance Superintendent R. Crawford, Work Planning Assistant Superintendent
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- C. Richards, Station Quality Verification
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- S. Koenig, Regulatory Performance Administrator
- T. Hammerich, Assistant System Engineer Supervisor
- M. Depuydt, Nuclear Licensing Administrator l
- J. Arnould, Regulatory Assurance i
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- J. Lewis, Lead Health Physics Operational
- D. Leggett, Assistant Superintendent Operating
- D. Carlson, NRC Coordinator i
- Denotes those attending the exit interview conducted on March 15, 1993.
i The inspectors also talked with and interviewed several other licensee l
employees during the course of the inspection.
2.
Licensee Action on Previously Identified Items (92701 and 927021 (Closed) Violation (373/92013-01(DRP)): Multiple examples of fire
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protection program implementation problems. The licensee modified fire watch requirements and trash storage area design to ensure
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administrative fire load limits were maintained.
In addition to procedure changes, training was given to applicable work groups to emphasize fire protection program requirements. The inspectors have no further concerns in this area.
(Closed) Open Item (373/92021-03(DRP)): Completion of training on water hammer for system engineers. The licensee prepared a general information notice (GIN) that discussed relevant information in regard to water hammer at nuclear power plants. The GIN was reviewed by the i
appropriate members of the technical staff.
This item is closed.
(Closed) Unresolved Item (373/93004-01(DRP)): Complete review of duct
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tape removal from safety relief valves (SRV) during installation. This
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item is discussed in paragraph 5.b of this report and will be tracked by the violation described therein.
No violations or deviations were identified in this area.
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3.
Licensee Event Reports Followup (92700)
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.The following licensee event reports were reviewed to ensure that l
reportability requirements were met, and that corrective actions, both immediate and to prevent recurrence, were accomplished or planned in-
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. (Closed) LER 373/92010-01 Automatic Reactor Scram Due To Low Charging.
-t Header Pressure
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i (Closed) LER 374/92015-00 Reactor Water Clean Up Differential Flow -
Isolation Due To Personnel Error and Procedural Deficiency
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I (Closed) LER 373/92016-00 Unplanned Engineered Safety Feature Actuation L!
During Reactor Protection System Bus Transfer Due to Personnel Error
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(Closed) LER 373/92013-00 Diesel Generator Automatic Start Due to Vessel Level Transient Due to a Design Problem j
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(Closed) LER 373/93001-00 Instrument Spike With Resulting Division II
Low Pressure Coolant Injection Pump Start
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j (Closed) LER 373/93002-00 Unit 1 Manual Scram Due to 'A' SRV Sticking
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Open Due to Duct Tape Covering the Actuator's Air Valve-
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In addition, recent deviation reports (DVRs) were reviewed in order to monitor conditions related to plant or personnel performance and to j
detect potential development of trends. Appropriate generation and-l disposition of DVRs, in accordance with the Quality Assurance Manual,
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were also reviewed.
No violations or deviations were identified in this area.
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4.
Operational Safety Verification (71707 and 71710)
i The inspectors reviewed the facility for conformance'with the license j
and regulatory requirements, j
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On a sampling basis the inspectors observed control room activities for proper control room staffing; coordination of plant activities; adherence to procedures or technical specifications; operator cognizance of plant pa'rameters and alarms; electrical power configuration; and the frequency of plant and control room visits by station managers. Various logs and surveillance records were reviewed for accuracy and completeness.
Significant observations were:
(1)
On February 10, 1993,- the inspectors noted leakage from.
beneath the gates into the Unit 2 dryer separator pit.from the reactor cavity. The leakage was traced to a high level in the skimmer surge tank overflowing the connecting vents
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I into the reactor cavity' area. As both fuel pools were
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cross-tied, level control was more difficult. The licensee j
planned to install better vent covers. The inspectors will i
monitor any further difficulties regarding fuel pool level
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control.
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(2)
The inspectors identified that safety evaluations performed
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on October 17, 1988 and May 8,'1992, for temporary system
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change (TSC) 1-1864-88 were inadequate. The TSC,
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implemented on October '25, 1988, provided a sump flushing i
system consisting of approximately 150 feet of two inch i
polyvinyl chloride (PVC) piping and temporary plywood door i
through which the piping was routed. -These safety l
evaluations failed to identify the TSC constituted a change
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in the fire loading and combustible materials described in
the updated final safety analysis report (UFSAR), Appendix l
H, " Fire Hazards Analysis," for fire zone 6E.
In addition, j
special considerations for PVC material contained in
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Appendix A to Branch Technia.1 Position APCSB 9.5-1,
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" Guidelines for Fire Protection for Nuclear Power Plants l
Docketed Prior to July 1,1976," dated August 23,_1976, j
Section D.2(c) were not addressed.
The licensee was i
committed to Appendix A.
10 CFR 50.59.b(1) requires that
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l records of changes to the facility as descri'ed in the
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safety analysis report include a written safety evaluation
which provides a basis for the determination that the change l
does not involve ~an unreviewed safety question (USQ). This
l is a violation of 10 CFR 50.59.b(1) in that safety l
l-evaluations did not recognize the UFSAR change, and
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therefore'this change did not receive an evaluation to l
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determine whether it involved an USQ. This violation was categorized as a Severity Level V and it is not being cited j
l because the criteria specified in Section VII.B.1 of the
" General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy, 10 CFR 2, Appendix C were satisfied.
Modification P01-1-90-571, " Unit 1 Control Rod Drive (CRD)
Hydraulic Control Unit (HCU) Drain Header," also utilized plastic piping. lue above violation was considered isolated in that the safety evaluation for modification P01-1-90-571 adequately addressed fire protection-issues. The inspectors did note a minor error in the combustible loading
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calculations for which the licensee initiated discrepancy l
record 93-035. The additional fire load of the TSC was small in comparison to the design fire load of the zone and l
the allowed transient combustible limit. The inspectors did i
not consider this safety significant as large amounts of PVC
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material was not present in-the plant and fire fighting crews were equipped with breathing apparatus to-deal with
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PVC toxic gas release.
The licensee initiated changes-to l
LaSalle Administrative Procedure (LAP) 240-6, " Temporary
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System Changes," to provide additional guidance on fire protection issues.
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On a routine basis the inspectors toured accessible areas of the facility.to assess worker adherence to radiation controls and the site security plan, housekeeping or cleanliness, and control of field activities in progress.
Significant observations were:
(1)
The inspectors concluded that the material condition of the plant continuous airborne monitors (CAM) was poor. This was based upon walkdowns performed February 4-5, 1993, work request review, and examination of radiation protection surveillances.
Seven of the 15 reviewed non-technical specification CAMS were not operable with the oldest inoperability since February 1991.
Repair of the CAMS received low priority but work activities did occur as time permitted. This was indicated by two of these CAMS (not included as inoperable in the above total) which had been repaired just prior to this review. The inspectors also noted that the plant technical staff did not have a system engineer assigned to maintain an overview of this equipment.
In contrast, the dual channel primary containment CAMS were operable. These were technical specification equipment,.
received higher work priority, and a system engineer was assigned. However, the inspectors noted a discrepancy between the LaSalle Radiation Procedure (LRP) 1350-21 surveillance records for those two days and the observed condition of one of these CAMS. Resolution of this discrepancy will be tracked by existing unresolved item 50-373/92013-02.
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A review of January 1993, surveillance records indicated numerous dates the LRP 1350-20 surveillance was not performed on the two refuel floor CAMS..The inspectors
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found one of these to be unplugged.
On other dates when the
surveillance indicated this CAM was shut down, health
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physics supervision was not notified.
This condition had
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not been identified such that it could be corrected. These
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were the only CAMS found in a contaminated area.
In addition, surveillance results for two LRP 1350-24 CAMS
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indicated that they were unresponsive to the source check.
Health physics supervision had not been notified as required
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by the procedure. One of these monitors had a pre-existing i
work request but the other had not been identified as a
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problem.
t Failure to perform the daily survefilance required by LRP 1350-20 for the two CAMS and failure to notify health
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physics supervision of equipment (three CAMS) that did not j
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function properly in accordance with LRP 1350-20 and LRP 1350-24 is a violation (50-373/93007-01 (DRP)) of technical l
specification 6.2.B which requires adherence to radiation i
control procedures. These single and 24 point sampling
systems were described in the UFSAR and credited in the NRC
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safety evaluation report to detect unusual leakage from
l sources of airborne radioactivity.
l The inspectors noted that the UFSAR section 12.3.4.2 and table 12.3-17, and the applicable licensed operator training j
lesson plan incorrectly listed a CAM for monitoring the
technical support center (TSC). This CAM had been removed
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l with the addition of another type monitor in the TSC several
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I years ago. The licensee indicated the UFSAR and lesson plan
would be revised accordingly.
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(2)
The inspectors noted housekeeping to be good during the
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inspection period. Cleanup of materials was much quicker l
following the recent refuel. outage compared to the previous one. Mechanical maintenance nonoutage work request backlog i
trended up more than the previous refuel outage due to
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resources expended in the cleanup effort. However, this was
l a planned decision by licensee management and the backlog trend turned by the end of the inspection period.
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Walkdowns of select engineered safety features (ESF) were
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performed.
The ESFs were reviewed for proper valve and electrical alignments. Components were inspected for leakage, lubrication,
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l abnormal corrosion, ventilation and cooling water supply j
availability.
Tagouts and jumper records were reviewed for i
i accuracy where appropriate.
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d.
Through direct observation, interview, and review of records, an l
inspection of the Unit 2 main steam leakage control system was performed. Operating procedures, surveillance procedures and l
records, deviation and licensee event report history, and the l
outstanding work request backlog were reviewed. The inspectors identified a minor calculational error regarding flow conversion
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during the review of surveillance tests. This error did not I
change the surveillance results. A physical walkdown of the l
system showed it was properly aligned, operated and maintained.
The material condition of the system was good.
l One cited violation, one non-cited violation, and no deviations were
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identified in this area.
5.
Monthly Maintenance Observation (627031
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Station maintenance activities affecting the safety-related and important to safety systems and components listed below were observed or
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reviewed to ascertain that they were conducted in accordance with l
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and did not conflict with technical specifications.
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t The following maintenance activities were observed and reviewed:
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WR L21704 IE51-F063 Reactor Core Isolation Cooling (RCIC) Steam Isolation Valve Disassemble and Repair
WR L21705 1E51-F063 RCIC Steam Isolation Valve Motor Breaker Trips on '
Thermals WR L21964 Rebuild of the IB Reactor Building Closed Cooling Water Pump i
WR L19555 Disassemble, Clean, Inspect, and Re-Assemble the 2A Drywell
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Instrument Nitrogen Compressor l
WR Ll7076 Install Pigtail Connectors on Specified Unit 1 Local Power i
Range Monitors (LPRMs)
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WR L21004 Unit 1 LPRM 32-41 Is Connected Incorrect
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Significant observations were:
The inspectors noted several events involving maintenance personnel
failure to follow procedure due to inattention to detail. The most
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notable of these are described below. Multiple examples of similar
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occurrences, mostly involving operations personnel, were noted in inspection reports 50-373/92021,; 50-374/92021; 50-373/92028; and 50-
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374/92028. Corrective actions, including a station stand down, appeared i
effective for operations but not entirely for maintenance personnel,
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The 'A'
SRV for Unit I failed to fully close during testing on January 26, 1993, due to maintenance error during the previous
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refuel outage.
LaSalle Maintenance Procedure (LMP)-MS-06,
" Removal / Installation of Main Steam Safety Relief Valves",
Revision 5, step F.7.26.8 stated, " Remove / verify tape removed from i
actuator air valve exhaust port at bottom of cylinder head". This l
was not accomplished which caused the failure and is-an example of i
a violation of technical specification 6.2.A.a (50-373/93007-02a (DRP)). The safety significance of thi.s event was small.
Although necessitating a manual scram, the failure occurred during
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SRV testing under controlled conditions and was an analyzed
condition.
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An LPRM was twice incorrectly connected despite identified post
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maintenance testing which should have identified the errors.
i During the Unit I refuel outage in late 1992, work request (WR)
L17076 involved work instructions for fabricating and installing
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quick disconnects (pigtails) on various LPRMs. As a sketch
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normally used to assist in the connection of the LPRM was
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misplaced, a maintenance foreman instructed the workers how to
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connect the LPRM. On December 2, 1992, system engineering j
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personnel verified and documented all LPRMs to be properly connected through LaSalle Technical Surveillance (LTS)-1100-6,
"LPRM Cable Connectica Verification". During unit startup, technical staff personnel identified, through observation of instrumentation, that LPRM 32-41 was connected incorrectly. On February 16, 1993, personnel performed WR L21004 to correct the connection problem. The WR listed LTS-1100-6 as a post
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maintenance test. This surveillance was not completed after the
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maintenance; however, due to poor communications the surveillance was signed off as being completed satisfactorily in the work
request.
During unit startup while performing another surveillance listed in the work request as a post maintenance test, LPRM 32-41 was again found to be improperly connected.
i Technical staff observations during both startups represented a good attitude toward problem identification.
LaSalle Administrative Procedure, (LAP)-300-7, " Preparation and
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Control of Nuclear Work Request," Appendix C, step C.31, stated maintenance is to be performed as required by work instructions.
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Both during the refuel outage-in December 1992, and during the
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February 16, 1993, forced outage, the work on LPRM 32-41 was not performed as required by the work instructions. Also, in December
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1992, LTS-1100-6 was not properly performed to verify that LPRM i
32-41 was connected correctly.
LAP-300-7. also stated in Appendix C, step 60, that post maintenance tests listed in the WR will be
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performed. After the maintenance was performed on February 16, i
1993, a surveillane listed as a post maintenance test in WR
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L21004 that verified the correct connections, was not performed.
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(50-373/93007-02b (DRP)) of technical specification 6.2.A.a.
The safety significance of the LPRM being incorrectly connected was
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minimal due to the problem being found at low reactor _ power. The average power range monitors (APRMs) would continue to function i
with the LPRM connected improperly. The number nf operable LPRMs input to the APRMs remained abcve technical specification requirements.
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One violation and no deviations were identified in this area.
6.
Monthly Surveillance Observation (61726)
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Surveillance testing required by technical specifications, the safety
analysis report, maintenance activities, or modification activities were observed or reviewed. Areas of consideration while performing observations yere procedure adherence, calibration of test equipment, identification of test deficiencies, and personnel qualification. Areas of consideration while reviewing surveillance records were completeness, proper authorization and review signatures, test results properly
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dispositioned, and independent verification documented. The following i
activities were observed or reviewed:
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LaSalle Instrument Surveillance (LIS)-NB-105A, " Unit 1 Reactor High
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- Pressure Scram Channels A and C Quarterly Calibration"
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LIS-NB-320A, " Unit 1 Reactor Vessel High Pressure Alternate Rod Insertion (ARI) Anticipated Transient Without Scram (ATWS) Instrument Channels A and C Monthly Functional Test" l
taSalle Technical Surveillance (LTS)-1600-11, " Core Flow Calibration and
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Jet Pump Data"
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LIS-NR-101, " Unit 1 Source Range Monitor Rod Block Calibration" I
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LaSalle Electrical Surveillance (LES)-CO-102B, " Diesel Generator IB Room
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C02 System Channel Functional Test" j
LaSalle Operating Surveillance (LOS)-RI-Q5, " Reactor Core Isolation
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Cooling (RCIC) System Pump Operability, Valve Inservice Test in
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Conditions 1,2, and 3 and Cold Quick Start"
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LTS-1100-6, "LPRM Cable Connection Verification" LOS-HG-sal, " Post-Loss of Coolant Accident (LOCA) Combustible Gas
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Control System Semi-annual Functional Test and Hydrogen Recombiner Check Valve Inservice Test"
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LOS-AA-SI, Attachment F, " Equipment Operator Instrument Channel Check Data Sheet" LES-GM-103, " Inspection of 4.16 KV and 6.9 KV I.T.E. Circuit Breakers"
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LIS-NR-303B, " Average Power Range Monitor (APRM) Channels B, D, and F Rod Block and Scram Functional Test" LOS-RP-M3, " Unit 2 Turbine Stop Valve Scram Functional Test" LTS-Il00-4, " Control Rod Scram Time Testing"
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i LIS-PC-204, " Unit 2 Drywell High Pressure HPCS Initiation Quarterly
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Calibration" i
Significant observations were:
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During the performance of LIS-NR-303B on February 23, 1993, the 'F'
channel scram relay failed to reset when one of the connectors to the
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relay came off. Operator actions in regard to this problem were
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excellent. This relay had been replaced during the previous refueling.
outage (October 92 - January 93) with a different model number as part i
of a preventive maintenance program. 'This and one other scram relay i
were the first of the new model numbers to be installed. The old model had screws on the connectors to tighten the fit. The new model simply
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had a clip on device. The licensee's original analysis of the problem was that it was improperly installed.
Further analysis by the technical i
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staff revealed that the connector could have vibrated loose even if properly installed. The scram relays were safety related and seismically qualified. The inspector requested a copy of the seismic qualification testing records to determine how the test was performed and whether it met the requirements of IEEE 344-1971 to which the licensee was committed. This is an open item (50-373/93007-03 (DRP))
pending receipt of those records.
No violations or deviations ware identified in this area.
7.
Safety Assessment and Quality Verification (40500)
The inspectors performed a licensing basis document review with regard to the onsite nuclear safety group (ONSG) required in technical specification 6.1.c.5.
This group served as the independent safety I
engineering group (ISEG) prescribed by NUREG-0737, " Clarification of Three Mile Island Action Plan Requirements."Section I.B.I.2 of NUREG-0737 stated, "it is expected that the ISEG may interface with the quality assurance (QA) organization, but preferably should not be an integral part of the QA organization." The licensee submitted a technical specification change request dated July 12, 1988, to reduce the required ONSG complement from four to three dedicated, full time engineers. Justification for this change included the existence of the onsite QA group which was also independent of plant operations, impacted on the functions of ISEG, and contained one to two people licensed at the SRO level. The information was utilized in the January 31, 1989, NRC safety evaluation report when the change was granted, indicating its importance as an NRC consideration in allowing the reduction. The 0NSG-complied with NUREG-0737 requirements during initial plant licensing review and the justification conditions existed at the time of the technical specification change described above.
The current ONSG did not appear to meet the plant licensing basis for the ISEG. During the massive 1992 licensee reorganization, the 0NSG was combined with the onsite QA group to form the quality verification (QV)
group. Staffing of the resulting group was cut in half. This tended to de-emphasize the formal onsite QA organization function and rely more upon self-QA by the line organization and periodic audits by offsite
organizations (which were also reduced in number). The licensee considered technical specification 6.1.C.5 met by ensuring at least three QV personnel met ONSG qualification requirements and QV-included ONSG functions. However, the ISEG was now an integral part of the QA organization contrary to NUREG-0737 intent.
In addition, the justification for the technical specification change was. negated in that i
the onsite QA group no longer existed as a separate organization with I
partially overlapping functions and no longer contained SR0 licensed-individuals. This same concern may also apply to CEC 0's other NUREG-0737 near term operating license plants (at the-time), Byron and
Braidwood. This is considered an unresolved item (50-373/93007-04 (DRP)) pending clear determination of whether the plant licensing basis i
for the ISEG was met, i
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By letter dated November 18, 1992, the NRC requested that the licensee respond to three concerns. The first concern was with regard to the adequacy of the licensee's ultimate heat sink technical specification
surveillance.
Based on the licensee's written response, dated January 18, 1993, and two subsequent working level meetings, the NRC agrees that making accurate measurements of sediment depth in the service water tunnel is impractical and that the continued performance of LTS-600-19,
"Corbicula and Zebra Mussel Inspections", during refueling outages will
ensure operability of the ultimate heat sink.
The NRC will continue to monitor the results of this surveillance.
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The second concern involved the accuracy of the residual heat removal
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service water suction pressure guages. The licensee acknowledged that j
on occasion these guages have been out of tolerance and have required i
recalibration or replacement. There has been no evidence that the out
of tolerance conditions were caused by sedimentation. The last concern related to updating the computer software program used to determine safe
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loading margins for-temporary lead shielding to reflect plant modifications.
The licensee determined that this was a valid concern i
and will complete programmatic upgrades prior to the next refueling i
outage.
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No violations or deviations were identified in this area.
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8.
Reoort Review (90713)
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During the inspection, the inspector reviewed selected licensee reports and determined that the information was technically adequate, and that it satisfied the reporting requirements of the license, technical i
specifications, and 10 CFR as appropriate.
No violations or deviations were identified in this area.
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9.
Unresolved Items Unresolved items are matters about which more information is required in
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order to ascertain whether they are acceptable items, violations, or
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deviations. An unresolved item disclosed during the inspection is
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discussed in Paragraph 7.
10.
Ooen Items j
Open items are matters which have been discussed with the licensee, l
which will be reviewed further by the inspector, and which involve some i
action on the part of the NRC or licensee or both. An open item
disclosed during the inspection is discussed in Paragraph 6.
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Exit Interview
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The inspectors met with licensee representatives (denoted in Paragraph 1) during the inspection period and at the conclusion of l
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the inspection period on March 15, 1993'.
The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and-did not indicate that any of the information disclosed during the.
inspection could be considered proprietary in nature.
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