ML20154M370
| ML20154M370 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/25/1988 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154M348 | List: |
| References | |
| TASK-3.D.3.4, TASK-TM 50-289-88-07, 50-289-88-7, DL-83-28, NUDOCS 8806010264 | |
| Download: ML20154M370 (32) | |
See also: IR 05000289/1988007
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-U.S. NUCLEAR REGULATORY COMMISSION.
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REGION I
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Docket / Report No. 50-289/88-07
License: OPR-50
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Licensee:
GPU Nuclear Corporation
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P. 0.' Box 480
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Middletown, Pennsylvania 17057
Facility:
Three Mile Island Nuclear Station, Unit 1
Location:
Middletown, Pennsylvania
Dates:
March 6, 1988 - April 9, 1988
Inspectors:
W. Baunack, Project Engineer, Region I (RI)
R. Conte, Senior Resident Inspector
- 0. Johnson, Resident Inspector
T. Moslak, Resident Inspector
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S. Sherbini, Radiation Specialist, RI
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A. Sidpara, Resident Inspector
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C. Woodard, Reactor Engineer, RI
Accompanied
by:
S. Peleschak, Reactor Engineer, RI
- Reportin Inspector
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Approved by:
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CT' Cow (1T1, Chief, Reactof Projects Section No. lA
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Inspection Summary: The NRC staff conducted routine safety inspections during nor-
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mal plant power operation.
Plant operational items reviewed were: loss of "B"
make-up pump; loss of "D" 125 volt a.c. vital bus; and, Safety Issues Management
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System (SIMS) Item No. B-75/85 for Generic Letter 83-28, "Post Trip Review Process."
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Other items reviewed in other functional areas included: reactor d.c. trip breaker
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N . 4 failure; diesel generator exhaust manifold fires, SIMS Item Nos. III.D.3.4.3
5
and M64800 on control room habitability; physical security; Independent On-site
Review Group (IOSRG); and, licensee action on previous inspection findings.
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Inspection Results: Operations activities continued to be accomplished in a safe
manner, operator attention and response to the three operational events was good
in that no major plant transient resulted from these initiating events.
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The IOSRG activities were improved, although the procedure was not being followed
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exactly. A procedure change is in progress to correct this situation,
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licensee action in the areas of radiological controls and engineering support for
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previous inspection findings was adequate.
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No violations of regulatory requirements were identified. Three unresolved issues
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were identified: one involved licensee action to evaluate 125-volt a.c. breaker
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trip sequences; the second concerned licensee action to resolve the problems as-
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sociated with the reactor trip breaker failure; and, the third item involved lic-
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ensee action to modify the 10SRG procedures to properly reflect the actions being
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accomplished by the 10SRG personnel.
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TABLE OF CONTENTS
Pagg
1.0 Introduction and 0verview............................................
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1.1 Licensee Activities.................................
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1.2 NRC Staff Activities..................................
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1
1.3 Persons Contacted...............................................
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2.0 Plant Operations..................
... ..............................
2
2.1 Cri teria/ Scope of Review (NIP 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2
2.2 Ev e n t s ( N I P 9 2 7 0 3 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3
2.2.1
Temporary Loss of "B" Make-Up Pump....................
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2.2.2
Loss of "0" 125-Volt a.c. Vital
Bus.........
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2.3 SIMS Item No. B-75/B-85 - Post-Trip Review Process (NIP 25564)..
4
2.4 Operations Summary. . .
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3.0 Radiological Ccntrols (NIP 83722)....................................
5
3.1 Organization and Qualification..................................
5
3.2 Licensee Actions on Previous Unresolved Items...................
6
3.2.1
(Closed) Inspector Follow-Up Item (289/83-26-03):
Temperature Effects on the Output of the Turbine
Building Sump Monitor.
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3.2.2
(Closed) Inspector Follow-Up Item (289/85-30-03): Rem
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Audit Tracking System.........
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3.2.3
(Closed) Unresolved Item (289/86-12-16): Procedure
Adequacy and Implementation for Shielding
Installation..........................................
7
3.2.4
(Closed) Inspector Follow-Up Item (289/86-13-04):
Radiological Control Department Organization for
TMI-1.................................................
7
3.2.5
(Closed) Violation (289/86-17-10): Failure to Provide
Design Basi s for Radiation Monitor Settings. . . . . . . . . . .
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3.2.6
(Closed) Violation (289/87-09-12): Failure to Survey for
Letdown Prefilter Cubicle Work........................
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3.2.7
(Closed) Violation (289/87-09-13): Failure to Follow
Control Procedures for When Standing Radiation Work
Permit Is Not To Be Used..............................
9
3.2.8
(Closed) Unresolved Item (289/87-09-14): Effectiveness
of Licensee Measures to Assure High Radiation Areas
Remain Properly Posted / Barricaded.....................
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3.3 RIR Review......................................................
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3.4 Radiological Controls
Summary...................................
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4.0 Equipment Operability Review - Maintenance / Surveillance
(NIP 61726/61703)....................................................
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4.1 C ri t e ri a/S c ope o f Rev i ew. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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4.2 Reactor Tri p Brea ker No. 4 Failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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4.3 Operational Test of the Emergency Of esel EG-Y-1B. . . . . . . . . . . . . . .
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4.4 Licensee Actions on Previous Inspection Findings. . . . . . . . . . . . . . . .
14
4.4.1
(0 pen) Inspector Follow-Up Item (289/86-10-02):
Significant Damage to the Diesel-Driven Fire Pump
Building..............................................
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4.4.2
(0 pen) Ure'esolved Item (289/87-06-05): Review of Delta
Pressure Instrument
Performance.......................
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4.5 Equipment Operability Summary...................................
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5.0 Engineering Support (NIP
37700)......................................
15
5.1 SIMS Items......................................................
15
5.1.1
(Closed) SIMS No. I11.0.3.4.2: Control Room
Habitability..........................................
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5.1.2
(Closed) SIMS No. M64800: Technical Specificstion for
Chlorine Detection and (Closed) Unresolved Item
(289/87-11-03)........................................
16
5.2 Licensee Actions on Previous Inspection Findings. . . . . . . . . . . . . . . .
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5.2.1
(Closed) Unresolved Item (289/86-12-14): Minimum Motor
Starting Voltages.....................................
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5.2.2
(Closed) Unresolved Item (289/86-12-12): Design Input
Associated with Emergency Feedwater Pump (EFW) Over-
current Protection....................................
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5.2.3
(Closed) Unresolved Item (289/86-12-15) and
289/87-10-01: Instrumentation Grounding and Shieldin
S t a n d a rd s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . g
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5.2.4
(Closed) Violation (289/87-01-15): Improper Mounting
of Foxboro D/P Transmitters...........................
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5.2.5
(Closed) Unresolved Item (289/86-03-22): Procurement
Deficiencies for Back-Up Instrument Ai r. . . . . . . . . . . . . . . 20
5.2.6
(Closed) Unresolved Item (289/86-12-01): Evaluations
of EFW Pump Recirculation Line Design Change..........
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5.3 E n g i n e e r i n g S u p p o r t S uena ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
5.0 Physical Security Plan Implementation (NIP 71881)....................
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7.0 Safety Assessment - Independent On-Site Safety Review Group
Performance (N,it' 40700)............ 7.i....................
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8.0 Eme rg e ncy P rep a redn e s s ( N Ip 717D7 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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9.0 Exit Interview (N!? 30703)....
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ATTACHMENT
ATTACHMENT 1 - Activities Reviewed
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DETAILS
'1.0
Introduct_ '
10verview
1.1 Licensee , tj- ;tes
During th'
ort period.- the plant operated at full power. As of 8':00
a.m. on April 9,1988, TMI-1 was operating at full power with the' reactor
coolant system (RCS) at normal operating temperature.(579 F average) and
pressure (2155 psig).
1.2 NRC Staff Activities
The purpose of this inspection was to assess licensee activities during
the power operations mode as they related to reactor safety, safeguards,
and radiation protection. Within each area, the inspectors documented
the specific purpose of the area under review, acceptance criteria and
scer of inspection, along with appropriate. findings / conclusions. - The
inspectors made this assessment by observation of licensee activities,
interviews with licensee personnel, measurement of radiation levels, or
independent calculation and selective review of listed applicable docu-
ments. NRC staff inspections are generally conducted in accordance with
NRC Inspection Procedures (NIP's).
These NIP's are noted under the
appropriate section in the Table of Contents to this report.
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Also, the inspector verified proper implementation, on a sampling basis,
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of licensee actions related to.the below-listeo NRC Safety Issue Manage-
ment system (SIMS) item.
The inspectior, approach for the SIMS item was:
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research various licensee and NRC correspondence, including Safety
Evaluation Reports (SER's) to identify key assumptions, commitments,
or other licensee actions to. be taken to resolve the safety issues;
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identify any additional items which need to be verified as deline-
ated in the related NRC Temporary Instruction or other inspection
procedures;
verify proper implementation of the items planned above; and,
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assess licensee performance related to that implementation and re-
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lated to dissemination of the issue and its resolution to licensee
personnel who need to know, such as by procedural upgrading and
training.
1.3 Persons Contacted
During this inspection, the following key licensee parsonnel- provided
substantial information in the development of the inspectors' findings.
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D. Atherholt, Plant Operations. Engineer
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S. Babzcak, Administrator, Sr., Human Resources
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H. Behling, Manager, Radiological Health, TMI-1
R. Barth, Fire Protection Engineer
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-J. Bowmen, Electr.ical Maintenance
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13. Brandt, Plant' Security
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G. Broughton, Operations / Maintenance Director
J. Colitr, Manager, Plant Engineering
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- J. Curry, 10SRG Chairman
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J. Dullinger , Plant Engineering
L. Edwards, Operations _ Quality Assurance Monitor
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K. Garthwaite, Plant Engineering
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R. Germann, Nuclear Safety Assessment Director
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D. Hassler,. Licensing Engineer
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H. Hukill, Vice President and Director, TMI-1
C. Incorvati, Audit Manager
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- R. Knight, Licensing Engineer
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P. Levine, Electrical Engineer
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T. O'Connor, Lead Fire Protection Engineer
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A. Palmer, Manager, Radiological Controls Field Operations
J. Pearce, Plant Materiel
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- M. Ross, Director, Plant Operations-
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J. Schmidt, Radiological Engineer
R. Shaw, Manager, Radiologica1LEngineer, TMI-1.
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H. Shipman, Piant'0perations E..gineer
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D. Shovlin, Plant Materiel Director
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r Smyth, Manager, Licensing
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- r. Snyder, Materiel Assessment Manager
J. Stevens, Corporate Engineer
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R. Warren, 10SRG
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S. Williams, Radiological Engineer
- Denotes attendance at final exit rneeting (see also Section 9).
2.0 Plant Operations
2.1 Criteria / Scope of Review
The resident inspectors periodically inspected the facility to determine
the licensee's compliance with the general operating. requirements of
Section 6 of the Technical Specifications (TS) in the following areas:
review of selected plant parameters foe abnormal trends;
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plant status from a matntenance/ modification viewooint, including
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plant housekeeping and fire protection measures;
control of ongoing and special evoluticNs, including control room
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personnel awareness of these evolutions;
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control of documents, including,logkeeping practices;
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implementation of radiological controls; and,.
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implementation of the. security plan, including access control,
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boundary integrity, and badging p'actices.
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The inspectors focused on the areas listed in Attachment 1.
. Findings.
and conclusions in this functional-area are . addressed below and, in other
functional areas, in other sections of this report.
2.2 Events
2.2.1
Temporary Loss of "B" Make-Up pump
On March.29, 1988, electricians were performing preventive
maintenance on.the "S" 480-volt a.c. bus tie breaker to the
"P" bus (Breaker 1S-12).
Main annunciator alarras C-2-7 "4KV
Engineered Safeguards (ES) motor trip" and C-3-7 "480.V ES
motor trip" were received.
Operators identified MU-P-1B as
being tripped and auxiliary / fuel handling buildings and control
tower ventilation as being shut down.
The control room operators promptly restored make-up and seal
injection using the "A" make-up pump. At essentially the same
time, the electrician' working in the "S" bus room-reported
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accidentally bumping and tripping.the supply breaker.to IC ES
valves Motor Control Center (MCC), which resulted in loss of
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the make-up pump lube oil pumps.
Letdown was re-estabitsbed
at 2.5 gpm and the "B" make-up pump was returned-to service
following hand rotating the pump.
The licensee prepared Plant Incident Report (PIR) No.1-88-01
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to document this occurrence and prescribe ~ corrective actions.
The inspector reviewed the report and concluded that the cause
was attributable to worker activities while performing main-
tenance-activities.
Corrective actions consisted of worker
and crew briefings which re-emphasized'the need to use care
when working with energized equipment that affects plant
operations. The inspector concurred that this event was an
isolated occurrence and not indicative of an overall problem
with worker actions having a negative affect on plant opera-
tions.
Licensee corrective actions were acceptable.
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2.2.2
Loss of "D" 125-Volt a.c. Vital Bus
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On March 31, 1988, the pisnt experienced the loss of the "D"
- 125-volt a.c. vital bus.
This resulted in the control room
receiving multiple plant' alarms on the main annunciator board.
Electricians were in the process of performing maintenance on
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the "D" battery charger; and, durin'g the' accomplishment'of the
maintenance, the d.c..and a.c. input breakers for thes"0"'in-
verter opened which resulted in the de-energizationLof the;"D"
125-volt a.c. bus.
The breakers tripped asca result of high.
voltage testing of. the "D" battery charger. Altho' ugh the
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charger was to be isolated during the test, it was not.
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Thelicensee-issueda-Plant'ilncidentReport(PIR)No.1-88-02
to document the problem and specify corrective actions.
The
.cause was a misinterpretation of procedure PM-E-18, which
called for the "0" battery charger to be re-energized per.
Operations Procedure (0P) 1107-2.
This resulted in'the "D"
battery charger being connected to the "0" inverter, which was-
not the intent of the PM procedure.
The result was that the
high test voltage applied to.the-battery charger tripped the
inverter input breakers.
Communication between maintenance andioperations' personnel.has
been re-emphasized. Additionally, an engineering evaluation
was requested to determine-if the breaker trip / coordination
sequence was proper. At the end of the inspection period, .the
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engineering evaluation was not completed. The licensee will
clari fy procedure PM-E-18 to more correctly specify the re-
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quired electrical alignment. 'This will remain an unresolved
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item (289/88-07-01) pending completion of. the~ engineering re-
view.
The inspector concluded'that licensee corrective actions
for this event-was acceptable.
2.3 S:MS Item No. B-75/B-85 - Post-Trip Review and Data
The inspector conducted a review to verify that a post-trip review pro-
cess was implemented and that post-trip data collection capability was
available as spe;1fied in the licensee response to-Items 1.1 and 1.2 of
Generic Letter 83-28 (Salem ATWS).
The licensee response to this item was contained in letters dated Novem-
ber 8,1983 and February 1, -1984.
These responses were reviewed by NRR
and censidered acceptable as documented in Safety Evaluation Reports
(SER's) dated May 31, 1985 and June 12, 1986.
During previous inspections, the licensee exercised the post-trip review
process for aight reactor trips since restart in October 1985. -The lic-
ensee utilized Administrative Procedure (AP) 1063 as. the primar/ post-
trip review process guideline.
The inspectors monito' red these post-trip /
reviews and verified that the licensee properly implemented the process
in accordance with the procedures.
These reviews were documented in-past
inspection reports.
The inspector discussed various portions of the
procedure and the personnel responsibilities involved with various lic-
ensee operations personnel. De procedure was well understood and was
carried out .in the past with few oroblems.
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The data -collection capabilities via- the transient monitor were reviewed.
All parameters required by the generic letter or as exempted in the NRR
SER have the capability of.being recorded.
This capability was enhanced
by the sequence of events. printout and the alarm printout data, which
was also available.
Strip chart records were available if the computer
was not available at the time of the trip.
The inspector concluded that the requirements of Generic letter 83-28-
for post-trip review were properly implemented at TMI-1.
An unresolved item presently exists (289/86-06-09) concerning the tem-
perature lag for instrumenthtion for the T-SAT monitor. 'This is a sepa-
rate issue that dces not affect 'the required capability of the post-trip
review process.
2.4 Operations Summary
Plant operations continued to be conducted in a safe manner. Operator
response during the "B" MU pump and "D" 125-volt a.c. vital bus incidents
wcs good and licensee corrective actions were adequate.
The problem with
poor- communications between operations and maintenance personnel was
viewed as having the potential to cause other problems with future
evolutions if not properly resolved.
Licensee review of procedures and
proper briefing of personnel prior to the conduct of these evolutions
was viewed by the inspector to bc vital to safe plant operations.
Lic-
ensee corrective action in this area will be monitored periodically in
future inspections.
3.0 Radiological Controls
3.1 Organization and Qualifications
The inspector reviewed the organization of the Unit-1 rsdiological con-
trols staff.
The qualifications of all members of the staff were also
reviewed to ensure that they met the minimum requirements of the Tech-
nical Specifications and applicable standards.
Nearly all positions
identified in the organization were staffed.
The staff members were
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fcund in all cases ta have qualifications at -least up to the required
levels and, in many cases, had much more formal and/or experience than
the minimum required for their positions within the. organization.
Additional review in the radiological controls area focused on licensee
action cn past findings.
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3.2 Licensee Actions on Previous Findings
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3.2.1
(Closed) Inspector Follow-Up Item (289/83-26-03): Temperature
Effects on the Output of the Turbine Building Sump Monitor
This item was opened'in connection with the observed. changes
in the output of the' sump monitor.with changes in.the water
temperature in the sump.
Investigations by the licensee showed
that these temperature-dependent changes were caused by changes
in the gain of the photomultiplier tube (PH) in the detector
assembly.
The detector is a scintillation detector with a.
s.dium iodide (Nal) crystal detector.
The gain changes were
caused by changes in the resistances of the resistors in the
PM tube voltage divider.
The detector was sent to the manu-
facturer.(Harshaw) to fix the problem.
The resistors were
changed to thermistors with negative temperature coefficients.
The licensee performed experiments with the modified detector
in place in the sump in June 1987. These experiments showed
that the temperature effects had been effectively corrected.
However, the data also showed that the output of the detector
changed with changes in sump water level.
These changes were
caused by changes in the amount of shielding provided by the
water in the sump. As the water level dropped, the shielding
effect decreased and the radiation fields in the turbine build-
ing above the sump caused an increase in detector reading.
Discussions with the licensee also showed that the method used
to calibrate.the detector was not representative. The detector
was calibrated by immersion in a 55 gallon drum containing
radioactive water.
The sensitivity of the detector determined
in that manner will be less than that with the detector im-
mersed in the sump water because the source size in.the sump
is larger than that in the drum.
In other words, the detector
in the sump should give a higher count rate for a given acti-
vity concentration than it would for the same concentration
in the barrel.
A review of the above considerations show that, although the
effects of calibration and water level dependence will cause
the detector's output to be different from predicted, both
effects are on the conservative side; that is, both effects
will cause the detector to register a higher than p edicted
reading for a given concentration of activity in the sump.
The trigger setpoint placed on the output of the detector is
used to trip the sump pump and, thus, prevent it from pumping
water from the sump to tSe Industrial Waste Treatment System,
which, in turn, discharges into the Susquehanna River.
The
calculated setpoint was 7600 counts per second (cps),' but the
setpoint used is 195 cps.
The licensee' stated that the main
reason for choosing this low setpoint was that it falls in the
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output ranga of the detector for which calibration data is
available. Use of a higher setpoint would place.the setpoint
in a different output . range, which, in turn, would require
additional calibration data. Based on the above considerations,
the current uncertainties and setpoint are regarded as being
conservative and acceptable.
This item-is'therefore considered
closed.
3.2.2
(Closed) Inspector Follow-Up Item (289/85-30-03): REM Audit
Tracking System
The licensee needed to revise procedure 9100-ADM-1201. The
audits were performed periodically by Radiological Engineering
and the subject of the audits was the radiological controls
program. At the time the item was identified, there was no
mechanism to ensure that all elements of the program were
audited within an audit cycle.
The licensee has modified the audit procedure and incorporated
an audit matrix in procedure 9100-ADM-1201.09, "Internal As-
sessments Procedare." The program _to be audited is divided
into ten elements and the procedure requires that an element
be audited at least every six months by a minimum of two radio-
logical engineers. The complete audit cycle would take five
years.
The inspector stated that an audit of a program element
once per five years appears to be too infrcquent. The licensee
stated that the six-month period between element audits speci-
fied in the procedure allows for a relaxation of' audit efforts
during exceptionally busy periods, such as outages.
The lic-
ensee stated that an' element is normally' audited every calendar
quarter, giving an audit cycle of less than three years.
This
item is closed.
3.2.3
(Closed) Unresolved Item (289/86-12-16): Procedure Adequacy
and Implementation for Shielding Installation
This item was opened in connection with the procedure for in-
sta11ation and removal of temporary shielding'. A procedure
was developed for installation of temporary shielding (9100-
,
ADM-3282.01, "Installation of Temporary Shielding"). A review
l
of this procedure indicated that the weaknesses identified in
connection with this item had been corrected.
3.2.4
(Closed) Inspector Follow-Up Item (289/86-13-04): Radiological
Control Department Organization for TMI_-1
At the time this item was opened, the position of Deputy Field
Operations Manager was included in the actual organization,
but it did not appear in the organization plan. Also, at that
time, certain functions, such as dosimetry, respiratory pro-
l
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tection, and in plant radiological. training, were not part of
the Unit-1 organization. .These functions had been transferred
i
to Unit 1.
- ,
The position of Deputy Manager Field Operations is currently
listed in procedure 9100-A0M-1010.01, "Department Organization
Plan," but it is not currently staffed and does not appear on
the department's organization chart.
The licensee indicated
that_this position will be"filled when the Unit-1 and Unit-2
radiological organizations are merged in the near future. 'The
omission of the position from the organization chart will be
corrected.
3.2.5
(Closed) Violation (289/86-17-10): Failure to Provide Design
Basis for Radiation Monitor Settings
'
This item is related to the selection of setpoints for monitors
RM-G16 through RM-G21 and RM-L1,
RM-G16 through RM-G21 are
ionization chamber area monitors.
Each detector is placed in
a location to monitor the radiation field from a pipe that
penetrates contaihment.
RM-G21 monitors the activity in the
reactor building sump.
The function of the detectors is to
isolate the lines (or sump pump) if the activity within them
exceeds predetermined levels.
This is~ intended to prevent
transfer of radioactivity outside of containment. -The isola-
tion setpoints were chosen on the basis of activities inside
the lines, expressed in uCi/cc.
However, the actual detector
settings were in mR/hr. At the time the violation was issued,
there was no data to show the relationship between the pipe
activities in uCi/cc and the detector setpoints in mR/hr.
The
licensee has since performed a series of calculationslof'ex'
posure rates at each detector location for specified activities
in the respective pipes.
The setpoints were selected for ac-
tivities expected to exist in the pipes for 1 percent failed
fuel conditions.
The calculated exposure rates were then used
to determine the appropriate setpoints. These setpoints have
been incorporated into OP 1101-2.1, "Radiation Monitoring Sys-
tem Setpoints." This item is therefore considered closed.
,
~
[ Closed) Violation (289/87-09-12): Failure to Survey for Let-
3.2.6
down Prefilter Cubicle Work
This violation was issued in conr.eetion with an incident that
occurred in the letdown prefilter room on March 7, 1987. As
a result of the events connected with that incident, two
workers were found to have internal contamination. A prelim-
inary critique of the incident was held by the licensee on
March 7, 1987, and a formal critique was held on' March 10, 1987.
Radiological Incident Report (RIR) was issued by the licensee
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on March 25, 1987 (RIR 87-0192). ~ The critiques and the RIR
were reviewed by the inspector. The corrective actions were
acceptable and the item is closed.
3.2.7
(Closed) Violation (289/87-09-13): Failure to Follow Control
Procedures for When Standing Radiation Work Permit is Not to-
be Used
The violation was issued in connection with an incident that
occurred on March 12,.1987. As a result of the incident, two
workers showed-external and internal contamination. Contamina-
tion occurred when the workers removed a yellow plastic bag
from a splash ring used on a high integrity container used for
spent filters. .The plastic bag did not have a radioactive
material label. A critique was held by the licensee on March
12, 1987, and an RIR was issued on March 12, 1987. The cri-
tique and RIR were reviewed by the inspector.
The corrective
actions were found to be adequate, and this. item is therefore
considered to be closed.
3.2.8
(Closed) Unresolved Item (289/87-09-14): Effectiveness of
Licensee Measures to Assure High Radiation Areas Remain
Properly Posted / Barricaded
High radiation areas are required by' Technical Specifications
to be barric:ded and posted. On.an inspection tour on April
20, 1987, the NRC inspector found the door to the waste
evaporator cubicle to be open with no barricade. The cubicle
was posted as a high radiation area (HRA), but surveys of the-
area at that time showed the radiation fields to be.less (40
mR/hr) than those that would require establishment of a high.
radiation area (100 mR/hr).
The licensee stated that this-
situation has been corrected by moving the posting for the
cubicle outside the door area.
This allows the door to be left
open to facilitate work in the area and still maintain proper
posting and barricading.
The inspector reviewed procedure
9100-ADM-4110.01, "Establishing and Posting Areas." The pro-
cedure only defines a HRA but does not explain what a barricade
is and how to establish such barricades.
The licensee stated ~
that the procedure will be changed to address-this weakness,
i
This item is therefore considered closed.
3.3 RIR Review
The RIRs generated in connection with closecut items 87-09-12 and 87-09-
)
13 discussed abovo were reviewed during this inspection.
In the case
of item 87-09-12, although RIR identified most of the problems that oc-
curred during the incident, it did not clearly identify and isolate the
root causes of the incident. A review of the events by the inspector
shows that two factors were responsible for the incident: the radiologi-
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cal controls technician who covered the job did not attend the pre-job
briefing, and the technician.and his supervisor failed to observe the
requirements of the RWP for the job.
The RWP clearly indicated that
respiratory protection was required for any activity other than visual
inspection.
The activities involved in the incident included movement-
of equipment inside the filter: room, but respiratory protection was not
used. The corrective. actions specified in~the RIR were limited to memos
to various supervisors to observe certain requirements and precautions.
The RIR did not clearly identify the -reasons for not attending. pre-job
briefings and for violating RWP requirements.
It did not, therefore,
propose changes in procedure that would address these deficiencies, nor
did it clearly identify the. persons and organization responsible for the
incident.
The inspector expressed these concerns to the licensee. The
licensee stated that the RIR was weak and.that efforts will be made in
the future to correct this weakness.
As in the case of the above incident, the RIR produced in connection with
item 87-09-13 also did not clearly isolate the root causes of the inci-
dent.
Corrective actions were again limited to memos to various super-
visors alerting them to precautions to take in similar situations.
How-
ever, the persons directly. responsible for the incident, and the proce-
dural violations involved, were not clearly identified. A review of the
incident by the inspector indicated that the root cause of the incident
was the fact that the plastic bag was not labeled.
The licensee stated
that the bag was yellow in color, indicating radioactivity, and that the
workers involved should have known that fact.
However, there was no
discussion in the RIR of why the workers did not respond appropriately
to the color of the bag, if indeed that statement is valid. There was
no clear indication in the RIR of why the bag was not labeled, and who
was responsible for this omission. The licensee stated that contaminated
items in radiologically controlled areas do not-need to be labeled (Pro-
cedure 3000-IMP-4400 01), "Radioactive Material Identification and Hand-
ling." However, the same procedure specifies that the labeling exemption
is for "... radioactive material being worked or otherwise handled by a
radiation worker." Otherwise, the material must be ' labeled to ". . . pro-
vide sufficient information to permit individuals handling or using the
material / containers or working in the vicinity thereof, to take precau-
tions to avoid or minimize exposure." The'RIR did not discuss these
considerations to determine how this procedural requirement applied in
the case in question, and if it applied, why there was no label; or, if
it did not apply, how might the procedure be modified to prevent recur-
rence of similar incidents involving unlabeled contaminated items. The
ir.spector expressed concern about regarding RIR serving the function for
which it wac intended, namely, identifying and correcting root causes.
The licensee acknowledged the inspector's concern and stated that they
planned improvements in preparing RIRs.
This area will be reviewed dur-
ing future inspections.
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3.4' Radiological Controls Summ'ary
l
Licensee actions on various RIR's were generally, weak. The RIR's were
not effective in identifying roct causes of the problems. .The licensee
committed to review the process to ensure that adequate reviews of
radiological' problems are conducted in the future.
Licensee action on
'
the remaining open issues was satisfactory.
4.0 Equipment Operability Review - Maintenance / Surveillance
4.1 Criteria / Scope of Review
The inspectors reviewed selected activities to verify proper implementa-
tion of the applicable portions of the maintenance and surveillance pro-
grams. The inspector used the general criteria listed under the plant
,
operations section of the report.
Specific areas of review are listed
in Attachment 1.
A mare detailed review of equipment operability is
addressed below.
4.2 Reactor Trip Breaker No. 4 Failure
On March 16, 1988, th(' licensee reported that a reactor trip breaker had
failed a portion of the menthly surveillance test.
The licensee was
,
performing Surveillance Procedure (SP) 1303-4 1 on Channel "0" of the
Reactor Protection System (R N ).
~
This serveillance is accomplished.on each of four RPS channels on a
rotating monthly basis. One channel is checked each week; hence, the
entire RPS surveillance is completed monthly. A part _of the surveillance
for Channel
"D" is to check the Nos. 3 and 4 d.c. reactor trip breakers
to verify that the shunt trip and undervoltage (UV) trip coil each func-
tion. The surveillance is accomplished with a test switch which can be
positioned to trip each coil independently.
On a normal reactor trip
signal, both coils actuate.
During the portion of the surveillance that checked the UV coil, it was
observed that when the test switch was positioned to de-energize the coil
.
for the No. 4 reactor trip breaker, the breaker did not trip.
The shunt
trip function was checked; and, it was verified that the shunt trip coil
was operable and, if an actual RPS trip signal was generated, the breaker
would have tripped.
j
The licensee removed the breaker and installed a previously tested spare.
The surveillance test for Channel "0" was completed satisfactorily.
The
technical specifications (TS) allow a_48-hour period to repair a faulty
reactor trip breaker, if only one of the diverse trip features was in-
As the failure did not result in loss of redun-
dancy, no other action by the licensee was required. An evaluation for
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reportability was made and the licensee determined that the event was
not reportable.
The inspector concurred in this evaluation based on a
review of.10 CFR 50.72/73.
The licensee disassembled the failed breaker to' determine the cause of
the failure.
Initial nbservations showed that the trip paddle for the
UV device had become mispositioned with respect to the armature of the
UV trip coil. This prevented the UV armature from moving to actuate the
trip paddle to trip the breaker.
The operation of the trip-shaft was
not affected as the shunt trip paddle was able to complete a breaker trip.
This was verified on three separate occasions.
Subsequent licensee investigation revealed that the clearances between
the end of the UV coil armature and trip paddle prevented the armature
from moving when the UV device was de-energized.
This was possibly due
to manufacturing anomalies with the UV device armature and'the trip
paddle.
The licensee was investigating changes in the preventive main-
tenance (PM) process to ensure that the clearance between the trip paddle
and.UV armature was proper.
Other reactor trip breakers at the facility
had not experienced this problem and the licensee had initially concluded
that this problem was not applicable to other breakers.
The inspector concluded that there was ressenable assurance that the
reactor trip system can function as designed.
This problem appeared to
be unique to the one particular breaker and the surveillance test
1
sequence and PM program used by the licensee should identify any other
breakers with this problem,
The inspectors will continue to follow lic-
ensee' actions to determine what additional PM effort or testing is re-
quired to enhance the reactor trip breaker operability.
-
The licensee has been in contact with the Babcock & Wilcox Owners Group
(BWOG) and General Electric (GE) to determine if any other corrective
action is needed. The licensee expects to receive guidance from B&W
concerning any additional preventiva maintenance that can be accomplished
to more adequately verify that the same condition that existed with' the
No. 4 d.c. reactor trip breaker does not exist with other breakers. This
guidance will be factored into site preventive maintenance procedures
after evaluation.
This item remains unresolved (289/88-07-02) pending
licensee changes to the preventive maintenance program. Additionally,
s
the licencee committed to provide a special report to the NRC staff-on
'
this problem, as a Licensee Event Report (LER) was not mandatory.
4.3 Operational Test of the Emergency Diesel EG-Y-1B
On March 6, 1988 at 4:13 a.m., during the shutdown following a surveil-
lance test, the diese? exhaust manifold caught fire.
The emergency
diesel room was manned during the test and the fire was put out immedi-
ately.
The local and remote speed indication was lost due to the tem-
perature which caused a disconnect of the internal tachometer wires.
The diesel was declared inoperable and diesel 6-Y-1A was tested as re-
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quired by Technical Specifications Section 3.7.2. -The diesel was re-
paired and back in service at 8:15 p.m. the same. day. Continued reactor
operation would have been allowed for 7 days.
The cause of the fire was determined to be a small leak of engine coolant
on the hot exhaust manifold.
There have been small fires in the past
caused by engine oil leakage through an exhaust manifold joint.
The licensee developed a detailed corrective action plan that included:
(1) removal of the exhaust manifold, inspection, cleaning, and reinstal-
lation with new gaskets; (2) replacement of all twelve thermocouples;
(3) repair of the small coolant line; and, (4) repair of the tachometer.
The inspection cf the exhaust system and the mating surfaces did.not
indicate any significant defects.
Following the above-mentinned repairs, the diesel was tested satisfac-
torily.
The inspector witnessed these activities, including job planning,
and reviewed the associated work packages for adequacy and completeness.
The inspector made the following observations.
The licensee initially planned on replacing two thermocouples (Nos.
--
4 and 12).
Further examination identified additional thermocouple
_
problems and the licensee decided to replace all thermocouples.
--
The inspector also noted two different styles of the replacement
thermocouples.
Some were of original construction and some had
smaller diameter metal sheaths.
Both styles have the same part
number and were accepted by licensee Quality Control (QC) inspec-
tors; however, the old style did not have a shelf life limit, while
'
the new style had a six years shelf life.
The licensee prepared
an engineering evaluation supporting the installation and also in-
formed the vendor.
It was the vendor's position that the shelf life
l
on the new style was not applicable and both styles were acceptable
for the application.
However, the vendor did not advise the licen-
see about the apparent difference in construction, as well as the
shelf life issue. Also, it appears that the licensee's receipt
inspection system did not detect the apparent discrepancy.
In re-
sponse to previous firdings (NRC Inspection Report No. 50-289/88-01),.
l
the licensee already has initiated efforts to correct such problems.
The residents office will review the effectiveness of the corrective
actions onc? implemented.
The inspectors discussed the diesel fire issue with licensee's man-
--
agement on March 21, 1988, to assess long-term actions. The licen-
see has developed a long term plan using Kepner Tregce techniques.
The technique is a management tool utilized to scientifically
analyze the generic issues with safety significance and then im-
plement an action plan.
The licensee plans to assess the adequacy
of the corrective actions already taken using this system.
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14
The repair was well pisnned and the licensee did an accepu ole Job in
implementing the shcrt-term actior.s. -The inspector reviewed a total of
seven job tichets.
The job tickets were well prepare.d and all the
planned work was accomplished and the diesel was declared operable on
the same day.
The inspector reviewed SP 1303-4.16 for botn emergency diesels.
The
surveillance data was recorded as required.
On both diesels, the-in-
spector noted that the temperature readings for several cylinders were
outside the recommended range.
However, they were still within the
allowable limits of maximum cylinder teraperature, as well as the maximum
differential temperature between any two cylinders. The rperability of
the diesels was not compromised by tnis temperature anomaly:'however,
the-licensee was in contact with the vendor to establish appropriate
normal operating limits.
The inspector also reviewed the annual surveillance performed per SP
1301-8.2 on both emergency diesels.
The inspector noted that on an older
surveillance (June 1987), some of the data corrections were not initialed.
This situation was corrected for the current annual surveillance.
The inspectors noted substantial efforts of senior licensee managen,en*.
in look?ng ct new ways to more efficiently condt.ct troubleshooting and
in developing a long-tern solution for correcting generic issues.
The
inspector had no furthe questions' regarding this issue.
4.4 Licensee Actions on Previous Inspection Findings
4.4.1
(0 pen) Inspector Follow Item (289/86-10-02): Significant
Damage to the Diesel-Driven Fire Pump Building
This item remained open pending Borated Wate: Storage Tank
(BWST) pipe tunnel plugging and evalut. tion of a fire pump dis-
charge check valve preventive maintenance program.
To prevent flooding.of the auxiliary building through the BW3T
pipe tunnel, the conduit duct banks in the 86T tunnel were
resealed. The inspector reviewed Job Ticket (JT) No.130 on
which this work was completed.
The inspector had no further
questions.
The licensee has completed inspections of eaisting Walworth
check valves in the plant.
The results of these inspections
indicate that there are no problems with other check valves
of this design.
The licensae is currently utilizing a non-destructive exacnina-
tion system (checkmate) that allows check valve performance /
operability determination without disassembly.
The results
of the checkmate examinations are to be verified by comparison
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with the results of'other check valve examination techniques.
After verification, the-licensee will' determine the appropriate
preventive mainterance- frequency for the Walworth check valves.-
.
This item remains open pending determination of the preventive
maintenance frequency of Walworth check valves. The licenses
plans- to determine -the frequency of PM's -by the end of this
year.
4.4.2
-(Open) Unresolved Item (289/87-06-05): Review of Differential /.
Pressure (D/P) Instrument Performance
.
This item remained open pending completion of licensee evalu-
ative actions.
The licensee'had increased the frequency of calibration of the
. main feed pump differential / pressure (0/P) switches from a
refueling basis to quarterly.
The data continued'to be un-
acceptable;
i.e.,
the setpoint continued to drift.
Licensee
personnel are evaluating possible corrective actions; one of
which is replacement of the D/P switches, which appears to.be
the best option.
No plans have been made, as yet, to accomp-
lish this task. This item remains open pending completion of.
licensee action to correct:this situation.
4.5 Equipment Operability Summary
Maintenance and testing activities continue to be accomplished in a safe
.
manner. No forced outages resulted from poor or incorrect maintenance
activities.
Licensee corrective action forfthe problems associated with
the reactor trip breaker are inconclusive, as yet, and any changes in
the maintenance practices for the breakers will .tre examined in future
inspections. The longstanding issues associated with diesel generator
fires, fire service check valve, and feedwater D/P instrument appears
open for an excessive amout.t of time.
This may be due in part to an
inappropriate prioritization of engineering actions.
5.0 Engineering Support
5.1 SIMS Items
5.1.1
(Closed) SIMS No. III.D.3.4.2: Control Room-Habitability
The 10 CFR Appendix A, General Design Criteria 19, "Control
Room," as well as NUREG 0737, Item III.D.3.4, defines the
specific criteria necessary to assure that the control room
is maintained in a safe habitable condition to assure that the
control room operators are adequately protected against the-
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accidental release of toxic 'and radioactive gases and to assure
the plant can be safely operated or shut down under design
basis-accident conditions.
The licensee's design was reviewed by the inspector and was
found to be acceptable.
The inspector had previously .identi-
fied-a few exceptions from NUREG 0737 requirements as docu-
mented in NRC Inspection: Report No. 50-289/87-02. During this
inspection period, the inspector reviewed these exceptions.
The details are os follows.
1
System Design Description (500): TI-670F on the chlorine
--
detection system stated the location of the two chlorine
probes, CE-776-2 and CE-777-2, being below the grade level
in the air intake structure.
The actual location,-however,
is above grade at the 320-foot evaluation.
The necessary.
change notice is in place and the SDD will be revised
accordingly.
--
SDD TI 670F, Section 1.6.6.2.2 stated the location of a
reset button was on Scrtion A of the heating and ventila-
tion (H&V) panel.
The actual location, however, was on
the Section 8 of the H&V panel.
This was the only excep-
tion and it re ri.1ed uncorrected. . The licensee intends
to revise the SDD to incorporate this. exception.
--
Operating Procedure (0P) 1104-19, as well as the Emergency
Procedure (EP) 1203-34 have now been revised to reflect
the current design of the Chlorine Detection System (CDS).
b
The chlorine detection system (CDS) is now fully operational.
The required testing and surveillance are being performed per
established procedures.
The inspector. reviewed the relevant
data and found it to be satisfactory.
The CDS requires con-
tinuous maintenance involving frequent replacement of the
<
chlorine detectors.
The licensee, however, maintains its
operability status by depending upon the routine surveillance
)
and testing, as well as the built-in, self-diagnostic features.
'
The inspector had no other comments on the installation and
)
operation of the CDS.
.
5.1.2
(Closed) SIMS M64800: Technical Specification for Chlorine
Detection and (Closed) Unresolved Item (289/87-11-03)
The safety grade Chlorine Detection System (CDS) was installed
for Cycle 6 startup.
The CDS involved installation of four
chlorine detectors.
Two detectors, CE 766-1 ano CE 777-1, were-
installed at the river water screenhouse and the second set
'
of detectors, CE 776-2 and CE 777-2, were installed at the vir
intake tunnel.
The system is designed so that any chlorine
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release in excess of 5 parts per million (ppm), the system' will
automatically be actuated and the Control Room Ventilation
System (CRVS) placed into a recirculation mode to prevent out-
side chlorine from entering the control room environment and,~
thus' maintain the required control room. habitability.
,
NRC Inspection Report No. 50-289/87-11 addressed the unplanned
periodic actuation of the CRVS_ caused by spurious high chlorine
detector resporse. All the items of concern have.been resolved
as follows.
The inspection report incorr2ctly. noted that the sensi-
--
tivity of the chlorine detectors was having a negative
impact on the system. According to plant engineering,
the electrolyte being used is chlorine-specific and does
not have any effect on the detectura sensitivity.
The
spurious actuation was caused by the direct exposure of
the detecters to the harsh environmental ccnditions.
Through the Change Modification Request (CMR) No. 0820M,
the licensee has corrected this problem by installing a
prot 9ctive umbrella over the detectors.
Since then, the
CDS is operating satisfactorily without any spurious
actuations.
--
In NRC Inspection Report No. 50-289/87-11, the inspector
also had questioned the effectiveness of the weekly pre-
ventive maintenance procedure 1C-145.
The review of this
procedure during the current insnection period does not
indicate any weaknesses.
The root cause of the problem
was corrected by the above-mentioned CMR.
All the open items on control room habitability, as previously
"dentified, have been resolved and verified by the_ inspector
and this item is closed.
5.2 Licensee Actions on Previous Inspection Findings
5.2.1
(Closed) Unresolved Item (289/86-12-14): Minimum Motor Starting
Voltages
NRC Inspection 50-289/86-03 identified concerns regarding the
adequacy of analysis performed to assure that sufficient volt-
age was available to start and operate certain safety-related
motor-operated valves (MOVi ).
An earlier analysis, which was
s
performed in 1979, resulted in the modificatfons of some of
the MOV units to ensure proper operatien.
However, this an-
alysis did not cover MOV's in the EFW and main steas (MS) sys-
tems, since at that time they were not considered safety re-
lated.
Further, the analysis assumed that the unit MCC bus
voltage and the voltage at the motor terminals were the same,
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The following documents which reflect licensee analysis / studies
and actions taken to ensure adequate 1 voltage were reviewed.
GPUN-TDR No. 114, Revision 1, "Adequacy of Station Elec-
--
tric Distribution System Voltages"
GPUN TDR No. 836, Revision 0, "Evaluation of Loading for
--
the Emergency Diesel Generators and Engineering Safeguards
Buses"
GPUN Memorandum, R. J. Hrabak/M. A. Materjorich to G.
S.'
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Saduska, "1986 TMI 230 KV Grid Voltage Study"
The licensee analyses were performed,.using their "DAPPER"
computer program with appropriate verification measurements
of voltage at selected locations, to provide confidence ir the
accura:y of the program. Based upon the analyses'which show
,
reinforcements in the grid transmission system, since the
original analysis, the licensee has increased the minimum
switchyard voltage consid9 red for degraded grid operation from
225 KV to 227 KV.
The min! mum voltages at the MOV terminals
were measured for both motor starting and full load currents.
The resultant voltages were found to be acceptable in accord-
ance with the criteria previously established in TDR No '114.
This item is closed.
5.2.2
(Closed) Unresolved Item (289/86-12-12): JDesign Input Associ-
^
ated with Emergency Feedwater Pump (EFW) Overcurrent Protection
During NRC Inspection 50-289/86-03, concerns were identified
,
regarding the overcurrent relay protection provided for large
safety-related motors. Of particular concern were the EFW
pumps and motors which originally were considered non-safety-
related nuclear components.
Design analysis support for the
EFW pump motor overcurrent protection were considered weak due
to incorrect relay settings and the apparent lack of considera-
tion for'long-term thermal degradation of the motors.
In response, the relay protection was evaluated for the EFW
pump motors and other large safety-related motors including:
Reactor Building (RB) Spray Pump Motors (BS-P-1A,-BS-P-1B)
--
Decay Heat Removal Pump Motors (DH-P-1A, DH-P-18)~
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EFW Pump Motors (EF-P-2A, EF-P-2B)
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Make-Up Pump Motors (MU-P-1A, MU-P-1B, MU-P-1C)
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RB Emergency Cooling Pump Motors (RR-P-1A, RR-P-18)
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These evaluations identified protective relay setting changes
for the spray pumps, decay heat removal' pumps,.and the EFW
pumps.
The inspector confirmed licensee implementation of' the
required changes.
5.2.3
(Closed) Unresolved' Items (289/86-12-15 and 289/87-10-01):
Instrumentation Grounding and Shield Standards
During NRC Inspection.50-289/86-03, concerns were identified
regarding the lack of established formal standard procedures
to cover proper grounding and shielding practices for instru-
mentation and control signal circuits. .NRC Inspection 50-289/
87-10 identified.a specific area of questionable shield
grounding in the Heat Sink Protection System (HSPS) instrumen-
tation.
This inspection confirmed that the licensee has established
and implemented Engineering Standard (ES)-028, Revision 0,
September 25, 1987.
This standard endorses Division of Reactor
Implement and Technology, USAEC, RDT Standard Cl-1T, "Instru-
mentation and Control Equipment Grounding and Shield'r.g Prac-
tices."
Field inspection and review confirmed that the licensee has
also investigated and corrected / resolved questionable ground-
ing/ shielding in the HSPS instrumentation under Field Change
Notice (FCR) No. C038886 and 056084.
These items are closed.
i
5.2.4
(Closed) Violation (289/87-01-15): Improper Mounting of Foxboro
0/P Transmitters
NRC Inspection Report No. 50-289/87-01, examined selected
safety-related components _ including Limitorque MOV's, Foxboro
transmitters, Rosemount transmitters, Target Rock solenoid
valves, level switches, radiation detectors,-temperature de-
tectors, and electrical splices.
This inspection revealed two
Foxboro transmitters mounting using U-bolts on vertical pipe
sections.
Tha 4-bolts were loose on both units. This mounting
is contrary t. _.e qualified seismic mounting specified in the
vendors manual.
The inspector confirmed that the licensee has evaluated the
questionable mounting in accordance with NRC Generic Letter
(GEL) 87-02, "Verification o' Seismic Adequacy of Mechanical
and Electrical Equipment (USI) A-46."
In addition, the licen-
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see is addressing the seismic adequacy of all: safety-related
equipment in order to provide the response required by GEL-
87-02.
This item is closed.
5.2.5
(Closed) Unresolved Item (289/86-03-22): Procurement Deficien-
cies for Back-Up Instrument Air
The inspector reviewed a Quality Deficiency Report (QDR), which
was issued because Purchase Order No. TP-035330 was issued with.
a safety classification of non-important to safety (NITS) and
incorrect dewpoint and filtration requirements.
The inspector verified that a Certificate of Confoer.ance was.
issued by the vendor ( \\ir ~ Products, Inc.).
The inspector also
determined that the quality of air received from the vendor
exceeded the~ required design dewpoint and filtration require-
ments. The purchase order has been revised to ensure adequate
controls on future procurement by classifying the-purchase
order as important to safety (ITS).
The inspector also reviewed the installation documentation for
the modification which installed an air compressor to provide
a permanent source of charging air for the.two-hour back-up
instrument air storage bottles.
The inspector verified that
dewpoints and filtration requirements are being met by this
modi ficatiori .
This item is closed.
5.2.6
(Closed) Unresolved Item (289/86-12-10): Evaluations of EFW
Pump Recirculation Line Design Change
Corrective Maintenance Modification No. CM0515M removed the
. instrument air tubing from the instrument air line downstream
of IA-V-1125 to EF-V-8A.
The justification for removal of the
instrument air line was that it was a potential safety' hazard.
Also, the instrument air supply to EF-V-8A is not credited as
a back-up supply since the instrument air compressors which
feed it are not important to safety ~(NITS).
The work was per-
formed under JT CH-269 per CM0515M.
This item was opened as a PAT inspection finding in NRC In-
spection Report No. 289/86-12 because the prcper safety evalu-
ation could not be located. An adequate safety evaluation was
performed and work was completed as ITS.
The inspector deter-
mined that removal of this line was previously evaluated by
NRC and approved.
This item is closed.
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5.3 Engineering Support Summary -
Licensee action on previously identified areas-in the engineering ' support
area was completed in an acceptable manner.
6.0 Physical Security Plan Implementation
On March 11, 1988, the inspector reviewed licensee's Physical Security Plan
and inspected several areas as mentioned below to assess plan implementation.
The installed search equipment, used by the licensee was found to be operable.
Some redundant equipment was not operable, was covered by the required com-
pensatory measures. The inspection also reviewed equipment test procedures,
test data, and their completeness.
No unacceptable conditions were identified.
The inspector toured the protected area with a security department represen-
tative. The protective fence, gates, locks, etc. were well maintained and
the isolation zones on both sides of the protected area was clean and clear
of any obstacles.
No weakness was observed.
The inspector witnessed the shift change at the Processing Center.
The
security guard officers were found to be very attentive.
The inspector also witnessed an inspection of a vehicle, which was performed
by the security guard.
It was done thoroughly.
The inspector inspected -the CAS and Secondary Alarm Station (SAS), reviewed
'
the relevant documentation, and monitored the operation of various equipment.
The entire operation was accomplished satisfactorily and_both facilities are
well maintained.
The assessment system had adequate clarity. _ The manning
of the security guards was in compliance with the security plan requirements.
The inspector joined the routine patrol with the "Scout" security guard.
This
patrol involves monitoring several equipment areas, fencing, gates, building
roofs, isolation zones, emergency power supply rooms, lighting, etc.
The
patrol activities were documented by the security guard, as required.
'
The entire security crew was found to be professional, well trained, and ex-
perienced.
The overall security functions reviewed were found to be adequate.
7.0 Safety Assessment - (Closed) Violation (289/85-27-09): Independent On-Site
Safety Review Group Performance
An inspection was conducted to evaluate the performance of the Independent
On-Site Safety Review Group (IOSRG) and to verify the corrective and preven-
tive actions taken by the licensee as described in a letter, H. Hukill to
T. Murley, dated February 10, 1986,-in response to a violation related to
10SRG activities identified during NRC Inspection 50-289/85-27.
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-During Inspection 50-289/85-27,'certain requirements associated with the per-
formance of the 10SRG were' identified as not having been performed. ' Also,
it was determined the overall effectiveness of the. group was difficult.to
-assess. 'During this inspection, the 1.icensee's corrective' action which had
been taken, the effectiveness of the corrective action, and overall perform-
ance of the group was evaluated.
An important factor in arriving at any conclusion associated 'with the , perform-
'ance of-the' group is that, since Inspection 50-289/85-07, all 10SRG members,
the Man'ager.of Nuclear Safety and the NSAD Director.to whom the:10SRG reports
have all been replaced with new personnel.
These personnel changes appear
to' have had a significant impact on the implementation of the corrective
action committed.to for the prevention of recurrence of previous ' procedure
adherence problems. As discussed later, however, a few of the previous prob-
l
lems still exist.
In addition to discussions with personnel, the following documentation was
reviewed to determine adherence to Technical Specifications (TS).and admini-
strative requirements.
,
Independent On-Site Safety Review Group Procedure - TMI-1, 6310-ADM-
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1010.01, Revision 5
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Qualification forms for each member of the group
Personnel training records for each member with the exception of the
--
consultant currently part of the group
--
Various 10SRG monthly reports
Various 10SRG bi-monthly reports
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TMI 10SRG work projections fcr 1987
--
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Various 10SRG record of review / investigation forms
'
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Records of TS Change Request reviews
Also, the following documentation associated with evaluations and assessments
,
'
were reviewed,
i
Human Performance Evaluation System Report - Replacement of Expansion
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Joint for RR-P-1B with Wrong Model
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Human Performance Evaluation System Report - Both Emergency Diesels Re-
moved from ES Standby LCO Violation
Containment Integrity, dated October 16, 1987
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Effectiveness Evaluation of GPUN Operating Experience. Review, dated June
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'1987
Evaluation of Shift Scheduling at TMI-1
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Differences Between TMI-1 Administrative Procedure (AP) 1043 and Cor-
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porate Procedure 1504-ADM-3040.01
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Potential for Hydrngen Combustion in the RCS
Failure Rate of RPS System
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Heated Posts
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TMI Saturation Monitor - Time Response
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As a result of the above reviews, many positive findings were identified.
.
However, some negative findings were also made.
In general, the IOSRG re-
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quirements of the TS appeared to be met.
Some implementing procedure non-
adherences were again noted and the corrective action committed as a result
of a previous violation was marginally implemented.
As required by TS, the group is comprised of a Manager - Nuclear Safety and
a staff of three qualified members.
In addition, another engineer skilled
in human performance evaluation has recently been assigned to the group.
'
The documentation considered to be formal evaluations performed by the group
are comprehensive, detailed, and well documented. These evaluations appear
to satisfy the overview review functions required by the TS.
The evaluations
appeared to be effective in that it was noted certain-reviews were brought
to the attention of the president of GPUN. A TS change was being made and
procedure revisions undertaken as a result of 10SRG evaluations.
Site per-
sonnel appear interested in the human performance evaluations being performed,
recommendations was being considered by operations, training had interfaced
with 10SRG, and top level corporate management had requested group evaluations.
A recommendation follow-up system should make the assessment of the group's
effectiveness even easier.
Monthly and bi-monthly reports are also prepared. These reports generally
summarize the significant activities of the 10SRG.
The monthly reports gene-
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rally provide slightly more information than the bi-monthly reports.
These
reports although identified as providing a summary of 10SRG activities do
quite frequently also contain some assessments.
The 10SRG procedure requires only bi-monthly summary reports and formal re-
ports of evaluations and assessments.
Inspection findings show the monthly
and bi-monthly reports do frequently include assessments and that reports
considered to be formal reports are not generally so identified.
For the most
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part,' documentation which serves as a formal report is usually distributed
as a memorandum. Also, the report distribution is not always as required by
TS or the procedure.
Other areas where the practice'is not in accordance with 10SRG procedure is
in the e
of "review records"- and the annual trending'of these records.
The
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use of "review records" was judged to be inappropriate and has been completely
discontinued. ;Also, the 10SRG, procedure describe; a method by which group
findings are resolved with responsible management and that only items which
are not resolved become recommendations.
The IOSRG in its documentation of
assessments and evaluations does not adhere zo this procedural requirement.
This is' discussed later in this section. Although not part of the 10SRG pro-
cedure, weaknesses were noted in that 10SRG recommendations are not:always
clearly identified.
That is, they are sometimes part of a conclusion while
at other times they are clearly noted as recommendations. Also, the results
of recommendations are not maintained, nor is any open item list maintained-
of internal commitments made in monthly ~ and bi-monthly reports.
These issues were discussed in detail with the licensee, particularly since
some of the findings were similar to those identified in NRC Inspection Report
No. 50-289/85-27.
The licensee was aware of the fact that-the 10SRG procedure was not being
fully adhered to.
A complete rewrite of the procedure in draft form had been
prepared and was still in the review process prior to being issued. The draft
procedure, among other things, addresses 10SRG project selection, schedules,
assessment reporting, records, and responsibilities. Along with the require-
ment for procedural compliance, the weaknesses primarily in the areas of
identification of what is an 10SRG assessment, assessment distribution in
order to meet requirements and to be most effective, and the identification
and follow-up to recommendations was discussed in detail with the licensee.
The licensee indicated these matters would be clearly addressed in the review
of the 10SRG procedure.
The licensee further committed to have the revised
procedure issued by August 5, 1988.
This item remains unresolved pending the
issuance of the revised procedure-(289/88-07-03).
,
During the follow-up to the corrective action specified by the licensee, it
was noted that certain of the 10SRG procedure changes which were committed
to in order to avoid further violations were marginally implemented or in-
effective.
For example: (1) The corrective action stated procedural clarifi-
cation would be made so that the use of the word "schedule" would be unam-
biguous. This was accomplished by completely eliminating the word "schedule"
in the revised procedure.
(2) Procedure clarification was specified which
would identify which reports of evaluations and assessments satisfy the TS
requirements. The revised procedure specifies bi-monthly reports are to pro-
vide a summary of reports of evaluations and assessments and formal reports
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of evaluations and assessments.
Currently, monthly reports and bi-monthly
reports are prepared which summarize major 10SRG activities and frequently
contain some assessments.
Few formal reports are issued; however, formal
evaluations are frequently issued, not as formal reports but as memoranda.
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(3)' procedural clarification was committed to, . ensure uniformity and consis-
tency in the documentation and handling of 10SRG recommendations.
The revised
procedure describes a discussion of. assessment findings with responsible
management and the preparation of a memorandum to document agreement reached
in.the development'of corrective or problem solving action. The procedure
specifies only those items which are not resolved may become recommendations.
This method of resolution of 10SRG assessments is not generally used.
Many
recommendations are specified in assessments, both clearly identified as
recommendations or frequently not clearly identified in a conclusion section
of an assessment.
This poor follow-up to committed corrective action and the need for accurate
communication with the NRC was discussed in detail with the licensee.
The
licensee indicated that there was no intent to be anything but fully respon-
sive to their implementation of these corrective actions.
However, with the
significar t changes in personnel that have taken place during the implementa-
tion and follow-up of the corrective action, the desired improvements had not
been fully achieved.
In order to address this situation, the licensee pre--
pared a draft procedure, which will correct this poor implementation of cor-
rective action. The quality of the. revised procedure and its implementation
following its issuance will be closely reviewed by the NRC to fully resolve
this matter.
This section closes outstanding item 289/85-27-09.
8.0 Emergency preparedness - Information Flow During Emergency Exercise
During the past emergency exercises, some communication problems surfaced be-
tween the licensee and the NRC Operations Center.
The licensee's Emergency
Notification System (ENS) and Health Physics Network (HPN) communicators could
not readily provide the requested information by the NRC, as required per 10 CFR 50.72(c)(3).
The problem was the large amount of information, as well
as the type of information.
In order to correct this problem the licensee
was provided additional guidance by NRC Region I.
Subsequently, the licensee
has revised two of their emergency-procedures.
The inspector reviewed these
procedures which reflect the changes consistent with the NRC guidelines.
The
guidelines were provided in a letter dated August 31, 1987, from Thomas T.
Martin, Director, Division of Radiation Safety and Safeguards to Mr. H. D.
Hukill, Director and Vice President of TMI-1.
The communication difficulties were related to the emergency exercises only.
The normal operations are not affected.
The communication during normal
operations is conducted per approved procedures and no inadequacies have been
observed. No further follow-up is necessary,
,
9.0 Exit Interview
The inspectors discussed the inspection scope and finding with licensee man-
agement at a final exit meeting on April 8,1988.
Interim exit meetings were
conducted on March 25, 1988, concerning radiological controls; March 31, 1988,
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26
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concerning several unresolved items; and, on April 5, 1988, concerning IOSRG.
In addition to those marked by cn asterisk in paragraph 1.3, senior licensee
personnel at the final exit meeting included:
--
J. Colitz, Manager, Plant Engineering
--
M. Hukill, Director, TMI-1
The inspection results as discussed at the meeting are summarized in the cover
page of the inspection report.
Licensee representatives did not indicate that
any of the subjects discussed contained proprietary or safeguards information.
Unresolved Items are matters about which more information is required in order
to ascertain whether they are acceptable, violations, or deviations.
Unre-
solved items discussed during the exit meeting are addressed in paragraphs
3.2, 4.3, and 5.2.
Inspector Follow Items
Inspector follow items are matters that necessitate further review and evalu-
ation by the inspectors.
TFese items are used to document, track, and ensure
adequate follow-up on matters of concern to the inspector.
Inspector follow
items are addressed in paragraphs 3.2 and 4.4.
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ATTACHMENT 1
NRC INSPECTION rep 0RT N0. 50-289/88-07
ACTIVITIES REVIEWED-
Plant Operations
--
Control room operations during regular and back shift hours, including fre-
quent observation'of activities in progress and periodic reviews of selected
sections of the shift foreman's log and control room operator's log and
selected sections of other control room daily logs
Areas outside the control room
--
,
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Selected licensee planning meetings
During this inspection period, the inspectors conducted direct inspections during
the following back shift hours.
Day /Date
Time
3/11/88
3:30 - 7:00 a.m.
Maintenance / Surveillance
Job Ticket (JT) CR-771/772 - Diesel generator thermocouple repair
--
JT CR-744 - Diesel generator oil leak repair -
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JT CP-504 - Diesel generator cooling system leak .epair
--
--
JT CP-505 - Diesel generator exhaust manifold flatness check
Surveillance Procedure (SP) 1301-4.1, Revision 42, effective April 13, 1988,-
--
"Weekly Surveillance Checks"
SP 1302-3.10, Revision 1, effective March 15, 198L, Chlorine Detection System
--
Instrumentation Channel Calibration"
--
SP 1303-5.16, Revision 3, effective November 17, 1987, "Chlorine Detection
System Instrumentation Channel Test"
Reactor Coolant System (RCS) Leak Rate
-The inspector selectively reviewed RCS leak rate data for the past inspection
period. The inspector independently calculated certain RCS leak rate data reviewed
l
using licensee input data and a generic NRC "BASIC" computer program "RCSLK9" as
specified in NUREG 1107.
Licensee (L) and NRC (N) data are tabulated'below.
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Attachment 1
2
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TABLE
RCS LEAK RATE DATA
All Values GPM
DATE/ TIME
CORRECTED
OURATION
L
N
N
N
L
G
G
U
U
U
3/14/88
0.4260
0.43
-0.03
0.07
0.0720
1:03 a.m
2 Hours
3/15/88
0.6760
0.68
-0.02
0.08
0.0843
7:58 a.m.
2 Hours
3/28/88
0.2902
0.29
-0.05
0.05
0.0568
2:23 p.m.
2 Hours
3/30/88
0.3161
0.32
0.05
0.15
0.1562
4:11 p.m.
2 Hours
G = Identified gross leakage
L - Licensee calculated
N = NRC calculated
- Declared invalid by licensee due to water addition to make-up tank.
Columns 2 and 3, 5 and 6 correlate + 0.2 gpm in accordance with NUREG 1107.
N is corrected by adding 0.1044 gpm to the NUREG 1107 N due to total purge flow
u
u
through the No. 3 seal from RCP's.