ML20236B289
| ML20236B289 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/10/1987 |
| From: | Bruce Bartlett, Cummins J, Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236B276 | List: |
| References | |
| 50-482-87-15, NUDOCS 8707290041 | |
| Download: ML20236B289 (14) | |
See also: IR 05000482/1987015
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APPENDIX B
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U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-482/87-15
LP: NPF-42
Docket: 50-482
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Licensee: Wolf Creek Nuclear Operating Corporation (WCNOC)
P. O. Box 411
Burlington, Kansas 66839
Facility Name:
Wolf Creek Generating Station (WCGS)
Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas
Inspection Conducted: June 1 to July 4, 1987
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Inepectors:
M4Nd
b'li444AxC6
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' . /E. Cummins, Senior 1[esident Inspector,
Date
perations
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M4A
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?-m-97
B. L. Bartlett, Resident Reactor Inspector,
Date
Operations
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Approved:
7!20 /I 7
D. R. Hunter, Chief, Reactor Project Section B
Date
Reactor Projects Branch
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8707290041 870722
ADOCK 050DO482
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Inspection Sunmary
Inspection Conducted June 1 through July 4, 1987 (Report 50-482/87-15)
Areas Inspected:
Routine, unannounced inspection including plant status,
operational safety verification, monthly surveillance ~ observation, monthly
maintenance observation, review of licensee event reports (LERs),10 CFR Part 21
report followup, onsite event followup, physical security verification,
radiological protection, battery bank surveillance, independent reactor
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coolant system (RCS) leak rate measurement, and engineered safety features
(ESF)walkdown.
Results: Within the 12 areas inspected, two violations were identified
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(failure to comply with TS, paragraph 5; and-failure to perform a containment
purge system surveillance, paragraph 4).
Two unresolved items are identified
in paragraphs 3.c and 5.
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DETAILS
1.
Persons' Contacted
Principal Licensee Personnel
- G. D. Boyer, Plant Manager
- 0. L. Maynard, Manager, Licensing
- C. M. Estes, Superintendent of Operations
- M. D. Rich, Superintendent of Maintenance
- M. G. Williams, Superintendent of Regulatory, Quality, and Administrative
Services
- W. J. Rudolph, QA Manager-WCGS
- A. A. Freitag, Manager, Nuclear Plant Engineering (NPE)
M. Nichols, Plant Support Superintendent
G. Pendergrass, Licensing
- W. M. Lindsay, Supervisor, Quality Systems
- C. J. Hoch, QA Technologist
- K
Petersen, Supervisor Licensing
E. Lehmann, NSE Engineer
- J. Allen, NSE Engineer
The NRC 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 July 9,
1987.
2.
Plant Status
The plant operated in Mode 1 during the inspection period, except during
the time period described below:
On June 29, 1987, the reactor tripped from approximately 92 percent power,
when Main Feed amp "A" tripped and the steam generator water levels
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reached the Lo-Lo level reactor trip setpoint.
The plant was returned to
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power operations in Mode 1 on July 1, 1987.
3.
Operational Safety Verification
The NRC inspectors verified that the facility is being operated safely and
in conformance with regulatory requirements by direct observation of
licensee facilities, tours of the facility, interviews and discussions
with licensee personnel, independent verification of safety system status
and simiting conditions for operations, and reviewing facility records.
The NRC inspectors, by observation of randomly selected activities and
interview of personnel verified that physical security, radiation
protection, and fire protection activities were controlled.
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By observing accessible components for correct valve position and
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electrical breaker position, and by observing control room indication, the-
NRC inspectors confirmed the operability'of selected. portions of
safety-related systems.
The NRC> inspectors also visually inspected safety
components for leakage, physical damage, and other impairments that could
prevent them from performing their designed functions.
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Selected NRC inspector observations are discussed below:
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On June 8,1987, at 3:20 p.m. (CDT) the licensee made a verbal report
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to the NRC duty officer per 10 CFR 50.72-(b)(1)(ii) concerning a leak
rate test on the containment personnel airlock.
The NRC inspector
reviewed Defect / Deficiency Report'(DDR)87-054 and LER 482/87-023
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which stated that on June 1,- 1987, the personnel ~ airlock barrel
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leakage test (STS PE-014A) failed to meet its ' leakage criteria.
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At the time of the test, the leak rate was beyond the range of the
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test instrumentation to measure and it was not until June 5, 1987,
that the leak rate was verified to be greater than 0.6 La by-
calculation.
At the time
PE-014A failed, the shift supervisor (SS) was
informed c
a test deficiency as required by procedure; however,
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since'the
was not informed that the leak rate was greater than
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O.6 La, the SS only entered TS 3.6.1.3 for an inoperable airlock,
failing to realize that he should have also entered TS 3.6.1.1
because of a lack of containment integrity.
Because of a combination
of circumstances, the requirements:of TS 3.6.1.1 were incidentally'
complied with. The NRC inspector concluded that the licensee had
complied with the applicable TS requirement by chance rather than by
design.
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b.
On June 14, 1987, at approximately 11:18 a.m. CDT licensee security
personnel reported to the control room that a loud rushing noise.was
coming from the hydrogen (H ) skid.
The H skid is located
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approximately 1000 feet north of the control room-and outside of the
protected area barrier (PAB).
H fires are not normally visible during daylight; however, the shift
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supervisor (SS) could see a heat wave effect when viewing the skid on
the security cameras.
The SS called out'the site fire brigade and
the City of Burlington Fire Department and then entered 0FF-Normal
Procedure OFN 00-016, " Fire Response."
The H bottles were sprayed with water to cool them and then the
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affected bottle was isolated, terminating the H leak.
Thw SS did
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not declare a Notification of Unusual Event (NOVE) because with the
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wind out of the east, the release of flammable gas did not.present a
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general hazard to other personnel or.the plant.
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Due' to the high summer temperatures, the site fire brigade leader was
overcome with heat exhaustion and spent the night in the hospital;
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however, he returned to work the next day.
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c.
During a review of environmental qualification (EQ) of certain
equipment, the licensee's staff determined that the main steam
isolation valve (MSIV) nitrogen (N ) accumulator pressure switches
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were not properly EQ qualified.
The switches are Barksdale
Model BDB1T-A65SS.
These switches were temporarily removed from the
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MSIV electr' al circuit pending further engineering review.
Engineering review determined that if these switches shorted to
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ground, a second ground would be-required to; affect MSIV operability
since it was a DC circuit. -The licensee is'considering the
desirability of future modifications to'these pressure switches
either to qualify them fully or to ensure that their failure would
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not affect the rest of the valve in any'way.
This will remain an
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unresolved item pending NRC issurance of enforcement guidance
(482/8715-03).
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d.
On June 24, 1987, during a routine tour of the control room on
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backshift, the NRC inspector observed the supervising operator (50)
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studying a textbook while on duty.
When questioned, the 50 stated
that the textbook was directly job related since' he _was taking the
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course to satisfy NRC requirements for a shift technical advisor and'
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studying was approved by his management.
The NRC inspector informed
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the 50 that NRC did not agree that studying was directly job related
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but the NRC inspector would contact Region IV for guidance.
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June 26, 1987, in a telecon Region IV informed the licensee that
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studying course textbooks while'on shift would not be in accordance
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with Regulatory Guide 1.114, Revision 1, " Guidance On Being Operator
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At the Controls Of a Nuclear Power Plant."
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No violations or deviations were identified.
4.
Monthly Surveillance Observation
The NRC inspectors observed selected portions of the performance of
surveillance testing and/or reviewed completed surveillance test
procedures to verify that surveillance activities were performed in
accordance with TS requirements and administrative procedures.
The NRC
inspectors considered the following elements while' inspecting surveillance
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activities:
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Testing was being accomplished by qualified personnel in accordance
with an approved procedure.
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The surveillance procedure conformed to TS requirements.
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Required test instrumentation was calibrated,
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Technical Specification limiting conditions for operation (LCO) were
satisfied.
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Test data was accurate and complete.
Where appropriate, the NRC
inspectors performed independent calculations of selected test data
to verify their accuracy.
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The performance of the surveillance procedure conformed to applicable
administrative procedures,
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The surveillance was performed within the required. frequency and the
test results met the required limits.
Surveillance witnessed and/or reviewed by the NRC inspectors are listed
below:
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STS CR001, Revision 5, " Shift Logs for Modes 1, 2, and 3,"' performed on
May 31, 1987.
STS BB001, Revision 4, "RCS Water Inventory Balance," performed April 1
and 25, 1986, and May 9, 1987.
STS IC-640A, Revision 4 " Slave Relay Test K640A Train A Motor Driven
Auxiliary Feedwater Pump Start," performed on June 17, 1987.
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Selected NRC inspector observations are discussed below:
On July 1, 1987, the licensee determined that required TS
Surveillance 4.3.3.11 was not being adequately performed due to
Surveillance Procedure STS IC-275B, Revision 4, " Analog Channel
Operational Test Containment Purge System Radiation Monitor GT RE-33,"
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which the licensee. c,ad established to accomplish this surveillance, was
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inadequate.
Performance of STS IC-275B between October 1985 and
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October 1986 and December 1986 and July 1987 failed to demonstrate that
automatic isolation of the containment purge pathway occurred when
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radiation monitor (GT RE-33) indicated measured levels above the Alarm / Trip
Setpoint as required by TS Surveillance 4.3.3.11.
Between October and
December 1986, during refueling outage, the required surveillance was done
by Surveillance Procedure STS GP001, Revision 5, " Containment Penetration
Integrity Verification." Up to July 20, 1985, STS IC-275B contained
adequate instructions for verifying that the containment purge pathway was
isolated.
Revision 3 to STS IC-2758, deleted these instructions from the
STS. The instructions were deleted as 'a part of the corrective action to
WCGS Defect / Deficiency Report No,85-105 which reported that Surveillance
STS 1C-275B, as it was then written, provided a flow path from containment
through the containment purge system to the environment without specifying
that this action required a release permit. .However, an alternata method
of demonstrating that the containment purge pathway automatically isolated
on a radiation monitor measured level above the Alarm / Trip Setpoint, as
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required by TS, was not provided.
This is the third reported instance
where the licensee has failed to perform a TS required surveillance (see
NRC Violations 482/8541-03 and 482/8534-02) and is, therefore, an apparent
violation (482/8715-02), even though the licensee identified the problem.
5.
Monthly Maintenance Observation
The NRC inspector observed maintenance activities performed on
safety-related systems and components to verify that these activities were
conducted in accordance with approved procedures, Technical
Specifications, and applicable industry codes and standards. The
following elements were considered by the NRC inspector during the
observation and/or review of the maintenance activities:
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LCOs were met and, where applicable, redundant components were
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Activities complied with adequate administrative controls,
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Where required, adequate, approved, and up-to-date procedures were
used.
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Craftsmen were qualified to accomplish the designated task and
technical expertise (i.e., engineering, health physics, operations)
was made available when appropriate.
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Replacement parts and materials being used were properly certified,
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Required radiological controls were implemented,
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Fire prevention controls were implemented where appropriate.
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Required alignments and surveillance to verify post maintenance
operability were performed.
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Quality control hold points and/or checklists were used when
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appropriate and quality control personnel observed designated work
activities.
Selected portions of the maintenance activities accomplished on the work
requests (WR) listed below were observed and related documentation
reviewed by the NRC inspector:
No.
Activity
WR 00702-87
Line No. 027 downstream of EJ-V033, below minimum wall
thickness
WR 02158-87
Hydro pressure may not have been adequate for repair
performed on WR 10370-85
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Selected NRC inspector observations are discussed below:
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During a routine review of WRs on June 23, 1987, the NRC inspector
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developed questions concerning WR 02158-87.
This WR had been written to
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resolve a quality first allegation concerning the test equipment
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configuration for a hydrostatic test.
Licensee personnel stated that the
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analysis for WR 02158-87 had just been completed and due to test equipment
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configuration that the hydrostatic pressure test performed for WR 10370-85
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could not be verified to meet the code requirement.
WR 10370-85 was written to repair EF137-HBC?4 which is a section of
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safety-related, ASME Code Class 3, piping located immediately downstream
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of Valve EF V090. On September 8,1985, a small amount of water was
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observed to be leaking from under the insulation and the repair was
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started on November 15, 1986. During the repair effort, the pipe was
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opened up and internal inspections revealed wall thinning due to erosion
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caused by flow through a throttled butterfly valve. At that time, a base
metal repair returning the minimum wall thickness to greater than
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0.320 inches and a hard surface deposit was performed. As required by the
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code for through-wall repairs, a hydrostatic pressure test was performed;
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however, as stated above this test was later found to have been performed
at approximately 215 aounds instead of the required 224 pounds or greater.
In discussions with tle licensee, the NRC inspector became aware that the
licensee was apparently unaware of the TS surveillance requirement of
4.0.5.
In a telecon on June 23, 1987, NRC gave verbal permission to WCN0C to
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delay performing a new pressure test, to meet code, until the next
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refueling outage since the pressure test could not be performed without
shutting down the unit. On June 24, 1987, the licensee informed the NRC
inspector that they had identified two other sections of safety-related
piping where code exemptions would be requested.
Line Number EF-080-HBC-24
immediately downstream of Valve EF V058 and Line EJ032-HBC-18 immediately
downstream of Valve EJ V033 were below minimum wall thickness due to
erosion and would be repaired as required by the code; however, they could
not be tested per the code with the unit in Mode 1.
Verbal permission was
given by NRC later that afternoon in a telecon~to wait until the refueling
outage to test or replace the piping.
Valve EF V058 is the essential service water discharge valve from the
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"A" train component cooling water heat exchanger, while Valve EF V090
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performs the same function on "B" train and has the same configuration as
EF V058. The licensee was unable to adequately explain their lack of
timely action to identify the wall erosion below EF V058 once the wall
erosion below EF V090 was identified.
This failure to comply with TS 4.0.5 and to notify the commission in
accordance with 10 CFR Part 50.55a(g)(5)(iii) is an apparent violation
(482/8715-01).
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When asked about TS 4.0.5, the licensee stated that they considered the
components fully operable, since they had not failed, even though the
components no longer met code.
The question of whether the licensee should declare a component inoperable
immediately upon determination that it fails to meet ASME Code will remain
an unresolved item pending further NRC review (482/8715-04).
6.
Review of Licensee Event Reports (LER)
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During this inspection period, the NRC inspectors performed followup on
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Wolf Creek LERs. The LERs were reviewed to ensure.
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Corrective action stated in the report has been properly completed or
work is in progress.
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Response to the event was adequate.
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Response to the event met license conditions, commitments, or other
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applicable regulatory requirements.
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The information contained in the report satisfied applicable
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reporting requirements.
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Generic issues were identified.
The LERs discussed below were reviewed and closed:
LER 86-028-001: TS Violation-Due to RHR Inoperability
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This LER was discussed in NRC Inspection Report 50-482/86-13 and concerns
the closure of one valve in one train of residual heat removal (RHR)
causing all of RHR to be inoperable.
The licensee has written procedures
to define actions to be taken if the RHR system temperature increases
during a plant heatup due to checkvalve back leakage. This LER was made
required reading for licensed operators and to reduce the problem of
checkvalve back leakage, the time :: pan that the reactor coolant system is
maintained at lower pressures during startups is being reduced.
This LER
is closed.
LER 86-034:
Inoperable Containment Isolation Valve
During a containtaent isolation valve local leak rate test (LLRT), it was
discovered that tha total leak rate through one valve was greater than the
TS limit of 0.6 La due to a mispositioned handwheel. The inner containment
isolation valve was operable throughout the time that the outer.
containment isolation valve was mispositioned. This LER is discussed
further in paragraph 4.
This LER is closed.
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LER 85-082:
Engineered Safety Features (ESF) Actuation-Partial Loss of
Off-Site Power
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Failure of a potential transformer in the 69 KV offsite power system
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caused the loss of offsite power to the 4160 V ESF Bus, NB01.
Per design
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Emergency Diesel Generator "A" started and picked up the ESF loads as they
were sequenced onto the NB01 bus.
All ESF components functioned per
design and the plant continued to operate in Mode 1 at 100 percent power
during the event and the subsequent restoration.
This LER is closed.
LER 86-062:
Control Room Ventilation Isolation Signal (CRVIS)
Power leads to a chlorine monitor were lifted as a part of an isolation
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clearance order.
Lifting the power leads caused an unanticipated ESF
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actuation (CRVIS) which'is required to be reported to the NRC. -To prevent
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a reoccurrence, the licensee changed the administrative procedure for
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clearance order control to include a step reminding personnel to conduct a
thorough review prior to removing equipment from service to identify
potential ESF actuations.
This LER is closed,
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7.
10 CFR Part 21 Report Followup
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The NRC inspector, by review of documents and discussions with licensee
personnel, verified that the 10 CFR Part 21 reports discussed below had
been reviewed and appropriately acted on by the licensee.
(Closed) P21-1987-87-18:
Inoperability Due to Retainer Nut Problems With
Discs in Main Steam Line Check Valves
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The licensee determined that there were not any components supplied by
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Schutte & Koerting Co. used at WCGS. ~WCGS does not use the dual check
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valve design in the main steam isolation valves (MSIV) that were reported
in the Part 21.
Anchor Darling dual disc gate valves were used at WCGS.
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This information was documented in WCGS Industry Technical Information
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Review and Evaluation (ITIP) No. 00359.
(Closed) P21-1987-87-20:
6" Model C Valve and Mercury Check-Devises
Failed to Open
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The licensee has determined that WCGS did not use the Automatic Sprinkler
Corp. Model "C" Valver reported by this Part 21 to have a problem.
This
information was documented in WCGS ITIP No. 00361.
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(Closed) P21-1987-87-38:
After Being Energized for a Long Period a 130 V
DC Relay in the DG Control Circuit Failed to Drop Out Af ter Being
De-Energized
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The emergency diesel generators at WCGS were supplied by Colt Industries.
WCGS does not use the emergency diesel generators or the Square D Company
Class 8501, Type KPD13 relay reported in this Part 21 as having a problem.
This information was documented in WCGS ITIP No. 00389.
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8.
Onsite Event Followup
The NRC inspectors performed onsite followup of nonemergency events that
occurred during this report period.
The NRC inspectors reviewed control
room logs and discussed the events with cognizant personnel.
The NRC
inspectors verified the licensee had responded to the events in accordance
with procedures and had notified the NRC and other agencies as required in
a timely fashion.
The events that occurred during this report period are
listed in the table below.
The NRC inspectors will review the LER for
each of these event and will report any findings in a subsequent NRC'
inspection report.
Date
Event *
Plant Status
Cause
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6/15/87
CRVIS
Mode 1
Blown Fuse
(100 percent)
6/23/87
CPIS/CRVIS
Mode 1
Technician Error
(100 percent)
6/29/87
Rx Trip
Mode 1
Main Feed Pump Trip
(92 percent)
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CRVIS - Control room ventilation isolation
CPIS - Containment purge isolation
Selected NRC inspector observations are discussed belowf
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The reactor trip which occurred on June 29, 1987, was caused by "A" main
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feedwater pump tripping which caused a reactor trip on Lo-Lo steam
generator water level.
All ESF features equipment operated as designed.
Both NRC resident inspectors were in the control room during the trip and
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verified that operator actions were timely, well coordinated, and
appropriate to the circumstances.
The NRC inspectors.will follow up and
review the LER for this event.
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No violations or deviations were identified.
9.
Physical Security Verification
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The NRC inspectors verified that the facility physical security plan (PSP)
was being complied with by direct observation of licensee facilities and
security personnel.
The NRC inspectors by observation of randomly selected activities verified
that search equipment was operable, that the protected area barriers and
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vital area barriers were well maintained, that access control procedures
were followed and that appropriate compensatory measures were followed
when equipment was inoperable.
No violations or deviations were identified.
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10.
Radiological Protection
By performing the following activities, the NRC' inspectors verified that
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radiologically related activities were _ controlled in accordance with the
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licensee's procedures and regulatory requirements:
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Reviewed documents such as active radiation work ~ permits and the-
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health physics shift turnover log.
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Observed personnel activities in the radiologically ' controlled
area (RCA)suchas:
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Use of the required dosimetry equipment.
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" Frisking out" of the RCA.
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Wearing of appropriate anti-contamination clothing where
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required.
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Inspected postings of radiation and contaminated areas.
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Discussed activities with radiation workers and health physics
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supervisors.
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No violations or deviations were identified.
11.
Independent Reactor Coolant System Leak Rate Measurement
The NRC inspector verified the adequacy of the licensee's calculational
technique to determine the RCS leak rate and independently verified that
the RCS leak rates were within the TS limits. Using the Computer-
Program RCSLK9 supplied by NRC, the NRC inspector updated the plant
parameter list to reflect Wolf Creek specific numbers and then taking data-
from leak rate. surveillance previously performed by the licensee ran the
RCSLK9 program.
Comparing the results of the RCSLK9 program to the
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licensee's surveillance, the difference was less than 0.2 gpm.
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No violations or deviations were identified.
12.
Battery Bank Surveillance
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On June 19, 1987, the licensee's staff was performing a routine review of
surveillance procedures and discovered that STS MT-019, Revision4, "125 V
DC Class 1E Quarterly Battery Inspection," performed on November 8,1986,
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failed to meet its TS criteria, yet, the associated action statement was
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not entered. The licensee's review showed that when STS MT-019 was
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performed on November 8,1986, that the average compensated specific
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gravity for Battery Bank NK11 was 1.199 and for NK13 was 1.205 which were
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below the TS limit of greater than 1.205 for all connected cells. -
TS 3.8.2.2 requires that as a minimum 125-volt Battery Bank NK11 and NK13
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and its associated full capacity chargers NK21 and NK23 or Battery
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Bank NK12 and NK14 and their chargers be operable in Modes 5 and 6.
On
November 8 and 10, 1986, at 12:25 a.m. (CST) refueling commenced and the
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unit entered Mode 6.
Since TS 3.8.2.2 allows one train of vital DC to be
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inoperable in Modes 5 and 6, there was not a TS violation at that time;
however, on November 16, 1986, at 11:19 a.m. the licensee deenergized
Bus NB02 making B train vital DC sources inoperable and should have
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entered the action statement to TS 3.8.2.2, which among other things
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prohibited core alternations. On November 17, 1986, at 3:05 a.m. Bus NB02
was reenergized.
Further licensee review has shown that while Battery Banks NK11 and NK13
were at all times capable of performing their design functions, it
was not until January 10, 1987, that STS MT-019 was reperformed proving
that the battery banks were fully operable. On December 11, 1986, the
unit entered Mode 4 with battery banks inoperable in violation of TS.
This violation of TS was identified by the licensee corrective actions
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have been taken to prevent recurrence, and it was not a violation that
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could reasonably be expected to have been prevented by the licensee's
corrective action for a previously identified violation. However, similar
violations have been previously cited. -Violation 482/8628-01 issued
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December 22, 1986, stated that on November 24, 1986, STS MT-016,
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Revision 3 " Standby Diesel Generator Inspection," performed on
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November 1,1986, was noted to have not been signed off by the test
performer, the supervisor, or the quality control supervisor / lead as
required. Violation 482/8632-01 issued March 13, 1987, stated that on
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December 5, 1986, STS MT-019 performed October 16, 1985, was noted to have
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average specific gravity out-of-limits which the reviews failed to
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identify or take corrective action. While these violations were
identified after STS MT-019 was performed on November 8,1986, their
corrective actions would not have corrected this violation. This failure
to properly review completed STSs and identify out-of-specification data
shows a lack of attention to detail. The NRC inspectors will continue to
follow the licensee's corrective actions to verify their adequacy to
ensure that surveillance performed in the future are adequately reviewed.
13.
Engineered Safety Features System (ESF) Walkdown
The NRC inspectors verified the operability of ESF systems by walking down
selected accessible portions of the systems. The NRC inspector verified
valves and electrical circuit breakers were in the required position,
power was available, and valves were locked where required. The NRC
inspectors also inspected system components for damage or other conditions
that could degrade system performance.
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The ESF system walked down during this inspection period and the documents
utilized by the NRC inspectors during the walkdown are listed below:
System
Documents
Auxiliary Feedwater System (AL)
Drawing M12AL01(Q), Revision 0, " Piping
and Instrumentation Diagram Auxiliary
Feedwater System"
Checklist (CKL) AL120, Revision 10
" Auxiliary Feedwater Normal Lineup"
SYS AL-120, Revision 6, " Feeding Steam-
Generators With a Motor Driven or Steam
Driven Auxiliary feedwater Pump"
No violations or deviations were identified.
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14.
Unresolved Item
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Unresolved items are matters about which more information is required in
order to ascertain whether they are acceptable items, items of
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noncompliance, or deviations.
Two unresolved items disclosed during the
inspection are discussed in paragraphs 3.c and 5.
15.
Exit Meeting
The NRC inspectors met with licensee personnel to discuss the scope and
findings of this inspection on July 9, 1987.
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