IR 05000289/1990012
| ML20058M293 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/01/1990 |
| From: | Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058M277 | List: |
| References | |
| 50-289-90-12, 50-320-90-05, 50-320-90-5, NUDOCS 9008100039 | |
| Download: ML20058M293 (13) | |
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U. S. NUCLEAR REGULATORY COMMISSION l
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Docket / Report No. 50-289/90-12 License: DPR150 l
50-320/90-05
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Licensee:-
GPU Nuclear Corporation i
.P. O. Box 480 f
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.Middletown, Pennsylvania 17057 Facility:
.Three Mile Island Nuclear Station, Units l'and 2'
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Locationi Middletown Pennsylvania Datesi May 15, 1990 June 26, 1990
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-Inspectors:
.R..Brady, Resident Inspector i
D. Johnson, Resident-Inspector i
T, Moslak, Resident-Inspector
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F. Young, Senior' Resident Inspector
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i Approved by: ' _k
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TO W. RuYand,.3ection Chief ~ '
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Reactor Projects Section No. 4B h:
Division of Reactor' Projects
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Inspection Summary:- Combined Inspection Report Nos 50-289/00-12 and
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50-320/9'D-05 for May 15, 1990 - June 26, 1990
s Areas R) viewed:
The NRC staff' conducted routine safety inspections of Unit 1 m'
' power operations and Unit 2 gleanup activities.
The inspectors reviewed
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plant operations, maintenance / surveillance, radiological practices, and
. engineering support activities as they related to plant safety.
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- addition,"a: review of housekeeping activities associated with Unit 2 i
reactor. building was performed.
Licensee action on previous inspection Li findings were also reviewed,
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Results:
The. highlights of this inspection are given in the executive summary; j
of this report,
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9008100039 900801'
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TABLE OF CONTENTS i
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1.0 Introduction and Overview.................
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1.1 Licensee Activities.......,.......,.
I 1.2' NRC Activities...................
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1.3.Persofi:nContacted..................
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- 2.0 Operations '(M707/64 704)*........... -......
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2.1 Facility Inspection..................
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2.2 Steam Generator Level Verification Test.
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-2.3 -Fire Protect w. Seview for Unit 2.....-.....
1 3.0 Maintenance' ard Suri:eillance (61726/62703)
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3.1 Surveillance Observations..............-
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3.2 Maintenance Observations................
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3.3 Reactor Building Containment Isolation Valve Failure 7'
a 4.0 Radiological Controls (71707)..............
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5'. 0 Engineering and Technical Support (37700)...-
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5.1-Diesel Generator Air Leak.... -...,......
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J 6.0 Licensee Action 'on Previous Inspection Findings _(NIP '92703):
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6.1 (Closed) Unresolved Item-(50-289/87-06-05) Calibration
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Frequency'of Feedwater D.P. $ witches..........
6.2: (Closed) Unresolved Item (50-289//88-17-03) Review of
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. Licensee Action for NI-11/12 Anomaly.......,
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6.3 -(Closed) Unresolved Item (50-289/88-22-01) Licensee L!
Evaluate Administrative Pror.edure-for Signature '
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Verification Sheet...
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'6.4- (Closed) Unresolved Item (50-789/88-22-02) Incomplete i
QA/QC Witness Signoff Sheets._.............
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ic-7.0 Management Meeting s (30703).................
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- * Denotes inspection modules performed
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DETAlLS l
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3.0 Introduction and Overview 1.1 Licensee Activities ~
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r At J.e be ning'of this reporting period, TMI-1 continued to be
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11 'ted a reactor power of 94 percent due to the."B" Once-Through-
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.Ste, 4erator. (OTSG) high level-at 92 percent. On June 22, 1990,-
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snsee performed a special test to determine the correlation-i
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p between OTSG 1evel and reactor power.
Reactor power was slowly
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raised toward 100 percent power to determine the maximum OTSG level
that could be attained. prior to affecting feedwater preheating.
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l Following completion of the test, the plant was returned to the pre-
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L test conditions of 94 percent reactor power and "A/B" OTSG levels of
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83 percent and 92 percent, respectively.
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Work continued in Unit'2 preparing the unit for post-defueling moni-f tored storage (PDMS).
This included work on-installation of shield-a
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ing on the reactor vessel work platform, decontamination and disas-
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sembling of defueling tools and general housekeeping in the reactor i
building. Modifications continued on the systent that will be used to
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evaporate the accident generated water.
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In light of the reduced work scope at Unit 2, the licensee has com-t bined the radiological controls departments associated with both u
units into one department.
The radiological controls program at TMI t
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is now administered under one manager.
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1.2 NRC Staff Activities
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j This inspection assessed the adequacy of licensee activities in-y cluding reactor safety, safeguards and radiation protection.
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inspectors made this assessment by reviewing.information on a sampl-i
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ing basis, through actual observation of licensee activities, inter-f views with personnel, or independent calculation and selective review of applicable documents.
Inspections were accomplished on both nor-
mal and back shift hours.
NRC inspections are generally conducted in accordence with NRC
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Inspection Procedures (NIPS).
These NIPS are noted under the appro-priate section in the Table of Contents to this report.
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Back shift inspections were accomplished during th 'ollowing j
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Day /Date-Time
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Sunday, May 19, 1990 8:00 a.m. - 17:00 p.m.
Monday, June 4, 1990 9:00 p.m. - h:00 p.m.
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Tuesday, June 5, 1990 10:50 p.m.
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Sunday, June 10. 1990 7.:30 s.m. - 10:30 a.m.
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Tuesday, June.19, 1990 10:00 p.m.
11:00 p.m.
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Tuesday,' June 26, 1990 8:30 p.m. - 10:30 p.m.
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l'3. Persons Contacted l
D. Atherholt, Operations Engineer
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G. Broughton, Operations / Maintenance Director
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- J. Byrne, Manager, TMI-2 Licensing
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G, Giangi,. Manager, Corp. Emergency Preparedness
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R. Harper, Manager, Plant Material
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C. Hartman, Manager, Plant Engineering i
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D. Hassler, Licensing Engineer
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L H. Hukill, Vice President and Director
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- G. Kuehn,-Site Operations Director
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R. Knight,' Licensing Engineer
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'M. Nelson, Manager, Safety Review
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- R. Rogan, Director, TMI-2 Licensing and Nuclear
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Safety
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- M. Ross, Plant Operations Director
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g, T. Seaver, QA Auditor
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H, Shipman, Operations Engineer
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E Schrull, Licensing Engineer
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G. Simonetti, Manager Emergency Preparedness
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R. Skillman, Director, Plant Engineering
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- P. Snyder, Manager, Plant Materiel Assessment -
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- C. Smyth, Manager, Licensing
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- R, Wells, Licensing Engineer
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- Denotes attendance at final exit meeting (see SEction 9.0) on June' 26, 1990.
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2.0 Plant Operations I
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2.1 Facility Inspection
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The resident inspectors routinely inspected the facility to determine
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the licensee's compliance with the general operating requirements of Section 6 of Technical Specifications (TS) in the following areas:
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review of selected plant parameters for abnormal trends;
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plant status from a maintenance / modification viewpoint,
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including pir.nt housekeeping and fire protection measures;
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l; control of. ongoing and special evolutions, including
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control' room personnel awareness of these evolutions;
_ control of documents, including log keeping practices; l
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implementation of radiological controls; and,.
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e l-implementation of the security plan,. including access
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control, boundary integrity, and badging practices.
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E In general, the inspector determined that the licensee, from a house-~
keeping and fire protection perspective, maintained Unit lin good condition.
Observations noted by the licensee's managers on back-shift tours appear to have positive effects on the_ licensee's ability l
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to quickly identify potential problems in the plant. Overall,-
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management attention toward plait safety was noted.
Details of the i
observations are noted below.
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2.2 Steam Generator Level Verification Test
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Background / Scope of Inspection f
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Following plant startup from the recently completed 8R outage, the
. licensee experienced difficulty obtaining full reactor power.
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was primarily due to fouling of the secondary heat transfer surfaces on both once-through-steam generators (OTSGsi.
The OTSGs are oper-
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ated at a variable-level which is power dept.ndent and the OTSG can
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only be_ operated up to approximately 92 precent level on the operat-ing range level instrumentation system.
Prior to June 21, the li-censee.had maintained approximately 8? percent level in the "A" OTSG
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and 91 percent in the "B" OTSG. The load on the individual OTSGs had
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been re-ratioed via ICS controls to obtain a variable difference be-t tween the'"A" and "B" T-cold temperatures of approximately 4.5
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degrees F.
This allowed more energy to be extracted from the "A"
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OTSG due to more fouling in-the "B" OTSG tube support plates.
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.SG 1evel had been limited to 92 percent for several reasons.
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ICS system' settings allow a maximum of 95 percent level prior to restricting feedwater flow. At approximately. 97.5 percent level, the water level in the downcomer is at a level where feedwater heating could be reduced or potentially stopped, due to reduced steam flow through the aspirating steam ports.
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On June 22, 1990, the licensee conducted a special test in accordance l
with Special Temporary Procedure STP-1-90-0039 to attempt to verify
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the actual maximum OTSG level at which the plant could be operated,
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without affecting OTSG operating parameters such as feedwater heat-ing, OTSG pressure, and superheat margin.
The licensee also com-pleted a safety evaluation (SE), #000224-009, Rev.1, which evaluated the affects of operating the OTSG outside previously acceptable oper-ating limits.
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The inspector reviewed the SE and concluded test it adequately ad-
L dressed the identified concerns associated with the test to be ac-i complished. Additionally, the inspector reviewed the STP and dis-
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cussed the. implications of conducting the test'with licensee person-L nel. The inspector also attended an operating crew briefing by the L
plant operations director.
The test was witnessed by the inspectors.
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Findings t
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The safety evaluation reviewed by the inspectors adequately addressed potential safety concerns for operating at OTSG-levels in excess of the presently accepted 92 percent. The licensee calculated that L
actual downcomer level was lower than indicated by.approximately 14
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percent.
This was primarily due to dynamic affects of the feedwater
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L flow direction in the downcomer.
The licensee evaluation concluded L
that OTSG 1evel could be increased above previously accepted limits with little or no affect on OTSG performance, The STP used for the test contained appropriate limits and precau-
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tions,.along with a very slow (3 percent per hour) power increase rate, such that operator action would be taken in a timely manner to.
preclude adverse plant effects. A limit of'10 degrees F decrease on lower downcomer temperature was selected as the end point for the test. OTSG tube to shell delta temperature was monitored to prevent exceeding the. limit of 60 degrees F.
The inspector did note that the test procedure wording for conducting the second phase was not pre-cise regarding the initial conditions required to_be reestablished, after completing the first phase of the test._ The licensee also
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recognized this problem and reconvened the plant review group and reviewed the applicable section of the procedure. Based on the pur-i-
pose and this review, the plant operating group established the ini-
.tial conditions that were required.
No changes to the existing pro -
cedure were required.
The licensee acknowledged the inspector's concern about the weakness in the procedure and stated that more precise wording would be used in subsequent test procedures.
The test was conducted in a controlled and safe manner.
The '.nspec-tors verified that appropriate limits and precautions were observed.
The licensee raised power to 100 percent without any detrimental affects on OTSG operation.
This was accomplished with no delta T-cold adjustment. Additionally, with delta T-cold adjusted to a positive 4 degrees F, to drive the "B" OTSG to a higher level, 100 percent power was also attained with a slightly higher "B" OTSG 1evel. The "A" OTSG 1evels remained at less than 90 percent. At the conclusion of the test, the plant was returned to pretest power
. level of approximately 94 percent with delta T-cold adjusted to a
negative 4 degrees F to maintain a lower "B" OTSG 1evel.
ICS OTSG 1evel limits at 95 percent were also restored.
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Conclusion
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The. inspectors had no safety concerns about the performance of the
test..The evolution was adequately controlled and documented.
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results are being evaluated by the licensec for potential impact on i
plant operations at higher OTSG 1evels.
'l 2.3 Fire protection Review for Unit 2
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On May 24, 1990, the resident inspector accompanied on NRC regional fire protection specialist and representatives of the licensee's fire
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Votection engineering'and industrial safety departments on a tour of
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the Unit 2 Reactor Building. The purpose of the tour was to assess the 'verall. housekeeping, industrial safety conditions, and fire preveition measures for the reactor building.
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In preparation'for entry into the reactor building (RB), the inspectors reviewed the current radiological surveys and radiation work permits and received _a briefing by Radiological Control
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Department peaonnel on measures to be taken while in the RB to
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minimize dose. Upon entering the RB, the inspector evaluated conditions on the ground level (305' elevation), operating floor
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-(347': elevation), atop the "B" D-ring, and on the shielded-work platform (SWP).
From this tour, the inspector observed that aside
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from the SWP, the remaining areas were congested and cluttered with hoses, cables and long handled tools.
Several randomly placed aerosol paint cans and a one gallon unmarked plastic bottle, containing an unknown liquid, indicated that the use of transient combustible materials was not rigorously controlled in the.RB.
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addition to the flammable liquids that were present, the inspectors
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observed excessive quantities of plastic bagging and sheeting throughout the RB.
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While crossing the ground level, the inspection team experienced an
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unsafe condition in that while a load of considerable weight was, being lowered from the operating floor (347' el.) to the ground level (305' el.), the team was not notified of the ongoing task and was
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permitted to walk beneath the load. The individuals performing the
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load transfer did eventually notice the presence of the inspection
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team in.the' load path and directed the team away from the descending load. The team concluded that this situation could have been averted if an individual was stationed near the load's landing to direct the
crane operator of interferences in making the tr.nsfer.
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trial safety engineer immediately discussed this problem with the
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supervisor. The inspector noted that the industrial safety engin-
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eer's discussion corrected the safety hazard.
The inspector had no
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further concerns.
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Through examining the condition of several portable fire extin-
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.guishers and. hose reels, the inspectors determined that the equipment l
was maintained in good condition, monthly inspections were performed,
as required, and the equipment was readily accessible.
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Based on these observations and experiences,-the inspectors concluded
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that the. licensee has not aggressively controlled the housekeeping
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and. industrial. safety conditions in the RB, The licensee is imple-menting.a satisfactory' radiological controls program by conscien-t
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tiously conducting frequent surveys to identify changes in radiolo-l
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gical conditions and by prescribing the appropriate contamination control measures, respiratory protection equipment and dosimetry.;
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3.0 Maintenance and Surveillance Observations i
3.1-Surveillance Observations
On a~ sampling basis, the inspector reviewed selected surveillance-
activities to ensure that specific programmatic elements described
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below were being met. Details of this review are documented in the following sections:
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The inspector cbserved performance of the following surveillance tests to determine that: -the test conformed to technical specifica-
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tion-(TS) requirements; administrative approvals and tagouts were
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obtained before initiating the surveillance; testing vas accomplished
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by qualified personnel in accordance with an approver'. procedure; test
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-instrumentation was calibrated; limiting conditions for operations were met; test data was accurate and complete; re:ioval and restora-tion of the affected components was properly accomplished; test results met technical specifications and prceedural requirements;
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deficiencies noted were reviewed and appropriately resolved; and the
surveillance was completed at the required frequency.-
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This inspection included review of the following= procedures:
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Surveillance Procedure (SP) 1303-3.1, " Control Rod Movements"
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Surveillance Procedure (SP) 1302-3.2, " Strong Motion l
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Accelerometer Calibration" p
Surveillance Procedure (SP) 1302-17.6, "RM-A-5/15 Calibration"
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Review of surveillance activities found that the licensee continues to properly implement the surveillance program.
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3.2 Maintenance Observations i
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The' inspector observed portions of selected maintenance activities to
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determine that the work was-conducted in accordance with approved
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procedures, regulatory guides,. technical specifications, and industry l
codes or standards.. The following items were. considered during this l
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L review: limiting conditions for operation were met while components
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or systems.were removed. from service; required a61nistrative ap--
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provals were obtained prior to' initiat,ing the. work; activities were
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e accomplished using approved procedures and QC hold points were estab-
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L 11shed where required; functional testing was performed prior to l
declaring the particular component (s) operable; activities were ac-
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complished by qualified personnel; radiological controls were imple -
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ments; fire protection controls were implemented; and the equipment
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was verified to be properly returned to service:
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These observations included the following maintenance activities:
River Water Pump House Cooling Coil preventive maintenance, per
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Job Order JO 00024098
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Decay Heat Pump, DH-P-1A preventive maintenance, per J0s 25957'
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and 25958 on June 26, 1990 The inspectors reviewed maintenaace activities and found that the
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licensee continues to perform maintenance evaluations adequately.
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3.3 ' Reactor Building Containment Isolation Valve Failure.
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On May 23, 1990, isactor building containment inboard isolation valve j
for the letdown line (MU-V-2A) failed to properly' stroke while per-forming a radiation monitoring system surveillance test.
This iso-lation valve is a four inch limitorque globe valve.
This valve, which is normally open, receives a close signal on indication of high
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radiation on the letdown line.
However, during surveillance testing the valve receives a partial closure signal to test this function.
During this surveillance, the isoittion valve failed to partially p
stroke.and tripped the associated breaker. Manual attempts to shut
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the valve from the control room also resulted in no valve motion and tripping of the valve motor circuit breaker.
Initial megger and
bridge resistance readings taken from the motor control center indi-cated that one of the electrical phases of the motor winding was
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grounded.
To further evaluate.the ground, a reactor building entry was made.- Megger and bridge resistance checks indicated that the ground was in the motor leads.
The licensee decided to replace the g'
motor rather' than to repair the motor lead:, in place.
An inspecticn of the motor by the licensee in the hot machine shop revealed a break-
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down of the motor lead insulation.
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L MOVATS~ testing was performed on the valve with the new motor.
In order to-limit thermal cycling of the letdown coolers, the licensee
performed partial stroke testing of the valve. The post maintenance
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testing. requirements were reviewed by the plant review group and
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N concluded that a partial stroke test was acceptable.
Review of the
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MOVATS test by the plot engineering group. concluded that the valve r[,
was operable, u
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The inspector discussed'and reviewed the licensee's documentation
F with plant engineering to insure that the licensee had properly repaired and tested the valve.
In addition, the inspector visually.
inspected the damaged motor to confirm the licensee's characteriza-b-
tion of the failure mechanism.- The inspector.concurreo M th the
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licensee's conclusion that a phase to ground fault was caused by degradation _of the motor lead insulation.
It appeared that the de-
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gradation was caused by inadvertent pinching of the motor leads be-
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twe0n the motor casing and the cover plate.
Visual inspection of.the
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motoa winding and rotor assembly revealed no additional damage. -With
.respoct to the inadvertent pinching of the motor leads, the licensee e
is.riviewing the applicable limitorque valve procedures to ensure the problem is properly addressed in.the procedure, Whether the problem-o
was (aused by a poor work procedure or a poor maintenance practice by an. individual, maintenance caused the valve failure.
Discussion with the licensee indicated that the correct approach to ensure that fur-ther valves are disassembled and assembled with respect to the elec-
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trical connections is being done.
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Review of the MOVATS testing of the valve with the new motor compared
k to previous base line data from prior testing indicated no unusual
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characteristics.
% ed on the data review, the inspector concluded that the valve had been properly returned to service.
4.0 Radiological Co'.trols
The inspectors observed routine activities associated with radiological l
control practices. These included dosimetry control, exiting radiological control areas, routine surveys, and area radiation monitor operation.
Conditions of step-off pads, disposal of protective clothing and personnel
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frisking were observed on a sampling basis, Noted activities reviewed or-witnessed, indicated that the radiological control program continued to be properly implemented.
5.0 Engineering Support 5.1 Emergency Dies 5' Generator Air Leak
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On May 21, 1990, at 7:00 p.m., the control room operator received an Emergency Diesel Generator (EOG) 1A Blocked alarm.
Subsequent investigation found an airline in the local engine mounted instrument panel (EMIP) had ruptured, causing a pressure switch to sense a low pressure condition.
The staring air receivers pressure dropped
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approximately ten pounds to 230 psig.
The starting air compressor
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I was able to maintain starting air pressure within its required band;
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therefore, the diesel was.still operable during the tube failure.
i While locating the leak, the operators isolated the starting air from the "A" EDG by closing EG-V-15A.
This_ rendered the EDG unable to automatically start from a valid start signal; however, an operator.
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was stationed.at the valve in case of an emergency.
The air leak was located and isolated, and EG-V-15A was reopened and independently verified open.
The EDG was inoperable for approximately seven
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minutes.
The licensee did not test run the "B" EDG due to the short duration the "A" EDG was inoperable.
h This is the thiru-rupture of the air _ line tubing over the past 13
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months.
Plant' engineering reviewed this event and documented the results of the evaluation in Engineering Evaluation Report (EER)
90-120M.
The installed ruptered tubing was 44-PP type tubing (1/4" tubing with 0.040" wall thickness with a working and burst-pressures
of 300 psi and 1200 psi respectively at 75 degrees F).
At 100 degrees F, the allowable working tressure drops to approximately 200
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L psi.
The diesel generator air system typically does not reach 100 degrees F.
Therefore, the applicat:on of this style tubing met the minimum requirements for this application.
The vendors' technical
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manual recommends 44-NSR type tubing (1/4" tubing with 0.050" wall thickness which has a working pressure of 600 psi). This type tubing maintains > a working pressure in excess of 300 psi up to 225 degrees i
F.
L Based on this evaluation, the licensee concluded that'the use of
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44-PP type tubing was marginally acceptable; however, 44-NSR type tubing was more appropriate' for this application.
The air start tubing has been replaced with the vendor recommended tubing in both EDGs.
Further research by plant engineering regarding the use of the 44-PP type tubing in the EMIP indicated that a modification was approved in 1985 allowing use of this tubing.
It is not clear that PP tubing was used in this application prior to this time, t
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The licensee reviewed and documented the potential for common mode R
failure of both EDGs due to air system tube failure in accordance with Administrative Procedure 1044, Event Review and Reporting Requirements..The emergency diesels were considered-operable based on the following information:
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sufficient air pressure has remained to start the diesel following a tube failure, and (2) tube
failure does not affect diesel operation once the diesel has started.
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The licensee is processing a change modification request (CMR076) to replace the plastic style tubing (44-NSR) with a more rigid reliable metal tubing, as a long term corrective action.
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the inspector concluded that the actions taken were appropriate. The i
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L plant engineering resolution to be untimely, and the use of 44-PP
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fication in 1985.
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Open Items y
6.1 = -(Closed) Unresolved Item (50-289/87-06'-05) Licen see l' valuate k
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l Calibration Frequency for Feedwater Delta Pressure Switches
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This item concerned the calibration of the delta pressure (dp)
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instrumentation used to generate the signal for auto-initiation of f
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HSPS upon loss of the main feedwater pumps.
The swi:ches in question
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L (4 total) measure dp across the main feedwater pumps. The instru-
ment calibration data revealed excessive drift in the 50 psig set-
point. 'The licensee increased the frequency of calibration to t
quarterly, but the setpoint of the. instrument continued to drif t.
The'
licensee subsequently reevaluated the setpoint and reset the trip
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point to 75 125 psig. A 50.59 evaluation was a ccomplished to docu-'-
-ment the analysis for this setpoint change.
Tre licensee plans to
repl&ce the instruments and upgrade the instrurrentation loops in the
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9R outage.
The inspector reviewed tM above referenced sefety evaluation
'No. 000424-006 and found that it was complete and. adequately evalu-ated the affects of the setooint change on ssstem operations.
No
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safety concerns were identi'ied.
The system now functions within the allowed setpoint tolerance os evidenced by recently completed cali -
bration check performed in Itarch of 1990.
The licensees' present
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action has resolved the cal bration concern, and action to upgrade the instrumentation in the rext outage will result in additional i
system reliability.
This item is closed.
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6.2 (Closed) Unresolved Item (50-289/88-17-03) Review of
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Licensee Action to Correct NI-11/12 Indication Anomaly l
.
.During startup from the 7R outage, the inspectors reviewed the oper-
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ation of the newly installed post accident neutron flux monitors,
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NI-11/12.- These were installed per Regulatory Guide 1.97.
NI-12 had operated erratically during initial plant startup.
Subsequent
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troubleshooting by.the licensee identified a problem with the signal-
pre-amplifier in the instrument circuit.
This problem was corrected.
Subsequent startup and shutdowns that have occurred since the pre-amp was repaired have confirmed the operability of NI-11/12.
The inspec-tor reviewed recent calibration data to verify that NI-11/12 are operating properly.
No problems were noted.
This item is closed.
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f b.3 (Closed) Unresolved-Item (50-289/88-22-01)- Licensee to
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Evaluate Administrative Procedures.o Incorporate Signature
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Jerification Sheet
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The licensee revised. Technical Functions Procedure SS003, Conduct of-
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Test to incorporate a requirement that all personnel conducting or
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initia11ng steps in test procedures, provide a-name to correspond to.
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the initial. This will provide for proper identification of ~
personnel actually performing the' test, The inspectorfreviewed the procedure' change, Revision 4 dated December 18, 1989, and concluded.
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that this procedure change resolved the inspectors concerns.
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item is closed.
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p 6.4 (Closed) Unresolved Item (50-289/88-22-02) Incomplete: QA/QC
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L Witness Signoff Sheets
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i This item concerned a test procedure TP 300/0.1 completed for the rewire of the ICS/NNI modules.
The test had.been monitored by QC_
personnel.
The test results were approved prior.to sign off by the
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i QC inspector.
The licensee now requires that QA/QC sign offs be x
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of completed.for monitored tests prior to acceptance by the-Startup and
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Test Manager.
In some instances, the QA/QC sign-off'on-the test.
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procedure may be' delayed until resolution of test comments but must
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be obtained prior to final test procedure approval.
The inspector-
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reviewed several completed test procedures from the recently com-l pleted 8R outage to verify that QA/QC sign off sheets.had been signed
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prior to final test procedure-approval.
No deficiencies were noted
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and this item is closed.
7.0 Management Meetings
- The inspector discussed the inspection scope and findings with licensee
,.
f management weekly and at a final meeting on June 26, 1990. Those person-nel.. marked by an. asterisk in paragraph 1;3 were present at the final i
m3n.agement meeting.
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