ML20153A858
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| ML20153A858 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/11/1988 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153A853 | List: |
| References | |
| TASK-2.E.4.2, TASK-TM 50-293-87-57, IEB-79-08, IEB-79-8, IEIN-87-030, IEIN-87-30, NUDOCS 8803210189 | |
| Download: ML20153A858 (30) | |
See also: IR 05000293/1987057
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V. S. NUCLEAR REGULATORY COMMISSION-
REGION I
. Docket / Report No. 50-293/87-57
Licensee:
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_ Boston Edison Company
800 Boylston Street
Boston, Massachusetts 02199
Facility: Pilgrim Nuclear Power Station
Location: Plymouth, Massachusetts
Dates: December 7, 1987 - January 19, 1988
Inspectors: C. Warren, Senior Resident Inspector
J. Lyash, Resident Inspector
T. Kim, Resident Inspector
Approved By: 1/ h cI8
A. Randy Bt6 ugh, Chie,f Date
Reactor Projects Section No. 3B
Areas : Inspected: Routine res. dent inspection of plant operations, radiation
protection, physical security, plant events, maintenance, surveillance, outage
activities, and reports to the NRC. The inspection consisted of 350 hours of
direct inspection. Principal licensee management representatives contacted are
listed in Attachment I. Observations made by the NRC Region I, Regional
Administrator during a tour on December 8,1987 are documented in Attachment
II of this report. A copy of Attachment II was provided to licensee management
for followup.
'Results:
Violation: Repeated occurrences of locked high radiation area doors being left
open and unattended were identified by the licensee. Problems with high radia-
tion-area access control have been previously identified and were the subject
of . violations during inspections 50-293/87-03 and 50-293/87-11'. Corrective
actions taken in response to these findings have not prevented their
recurrence. (Section 3.b, VIO 87-57-01)
Unresolved Item: The licensee identified that two reactor vessel level gauges .
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were incorrectly installed. A licensee investigation is currently ongoing to
determine the cause and to assess the adequacy of post installation test.
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(Section 4.d, UNR 87-57-02)
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8803210189 880311
PDR ADOCK 05000293
Q DCD
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Inspection Results-(Continued)
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Concerns:
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- 1. The ~ licensee. experienced safety rele+.ed equipment malfunctions upon
receiving ~ a spurious reactor- scram ' signal 'on January 17, 1988. -(Section
4.d)
2. Inadequa'te procedures and planning of - surveillance tests resulted in un-
necessary engineered safety feature actuations. (Section 3.a)
3. Poor preplanning and control of maintenance was noted during an-electrical
relay replacement. A similar problem was the subject of a violation dur-
ing inspection 50-293/87-50. (Section 4.c)
4. Weak identification and tracking of lifted leads and jumpers led to a
water spill in the.high pressure coolant injection system room 'during the
integrated leak rate test. (Section 6.0)
5. The prelube' pump for the "B" emergency diesel generator failed to. restart
during a surveillance test. An identical failure occurred during a loss
of offsite power event on November '12,1987.. Licensee followup appeared
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adequate but the failure root cause has not been identified. (Section
3.b)
6. The inspectors evaluated the erosion of construction dirt into wetlands
area. The inspector's independent survey of the area, and the licensee's
analyses indicate that the level of activity does not represent a health
or safety concern. However, the material should not be allowed to erode.
(Section 3.c)
Strengths:
1. The licensee's preparation and execution of the reactor vessel hydrostatic
test was well organized and controlled. (Section 5.0)
2. .The licensee's response to a January 17, 1988 reactor scram signal and
subsequent equipment malfunctions was prompt, thorough and effective.
(Section 4.d)
3. Using non-nuclear steam for testing of high pressure coola1t injection
system and reactor core isolation cooling system enabled the licensee to
discover problems which may not have been easily identifiable using
nuclear steam due to radiological conditions. (Section 3.b)
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TABLE OF CONTENTS
Page
1. . Summary of Facility' Activities ........................ 1
2. Followup on Previous Inspection Findings .............. 1
3. Routine Periodic Inspections .......................... 4
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a. Surveillance. Testing
b. ' Radiation Protection and Chemistry _
c. Fire Protection
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4. Review of Plant Events ................................ 15
a. . Spurious Isolations of RHR Shutdown Cooling System
b. Reactor Water. Cleanup System-Spurious' Isolation
c. -Engineered Safety Feature Actuations Due to a
Failed Logic Relay .
'd, Spurious Reactor Protection System Actuation
5. Review of Reactor Vessel Hydrostatic Test Procedure
: a nd - Te s t ' Re s u l t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
6. Integrated Leak-Rate Testing .......................... 21
7. Licensee Nuclear Organization Managenant
Realignment ......................................... 23
8. Management Meetings ................................... 24
Attachment I - Persons Contacted
Attachment AI - Regional Ad:ninistrator's Tour Observations
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DETAILS
1.0 Summary of Facility Activities
The plant was shutdown on April 12, 1986 for unscheduled maintenance. On
July 25,1986, Boston Edison announced that the outage would be extended
to include refueling and ' completion of certain modifications. The reactor
core was .defueled on February 13, 1987. The licensee completed fuel re-
load on October 14, 1987. Reinstallation of the reactor vessel internal
components and the.vessei head was also subsequently completed.
During this report period, the licensee performed the reactor' vessel
hydrostatic test and the primary containme- . integrated leak rate -test
(ILRT) as described in Sections 5.0 and 6.0. On December 9, 1987, Pilgrim
Station conducted a partial participation emergency preparedness exercise.
On December 14, 1987 the licensee announced as part of a planned manage-
ment realignment, the appointment of eight managers to key management
positions in the licensee nuclear organization at Pilgrim Station. The
details of the management realignment are described in Section 7.0.
NRC inspection activities during the report period included: 1) observa-
tion of the licensee's annual emergency preparedness exercise on
December 9, 1987, 2) NRC Reactor Operator Licensing examinations were
administered to eight candidates on the week of December 7, 1987, 3) ob-
servation of the primary containment ILRT and review of the test results
during the week of December 21, 1987. The results of these inspections
are documented in inspection reports 50-293/87-54, 50-293/87-56, and
50-293/87-58. In addition, representatives of the NRC's Office of Inves-
tigation were onsite December 3, December 7, and December 8, 1987 to
interview onsite security personnel. On December 8, 1987, the NRC
Regional Administrator for Region I, Mr. William T. Russell, toured the
plant with the resident inspectors. On January 7, 198,8,
Dr. Thomas E. Murley, Director of the Office of Nuclear Reactor Regulation
(NRR) and other NRC representatives toured the plant with the resident
inspectors.
2.0 Followup on Previous Inspection Findinjs
(Closed) Unresolved Item 82-E4-02 - Discrepancies in the Licensee's
Response to IE Bulletin 79-08
Previous reviews of this item are documented in the inspection reports
50-293/82-30, 50-293/83-01, 50-293/83-14, and 50-293/84-26. IE Bulletin
(IEB) 79-08 ano the TMI Action Plan Item II.E.4.2 required licensees to
review the containment isolation initiation cesign and procedures to
ensure proper initiation of containment isolation, upon receipt of an
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automatic containment isolation signal. The licensee provided the
results of their review in letters dated April 25, and August 21, 1979.
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The licensee stated that the RBCCW supply and return lines, instrument air
line, RHR to spent fuel pool cooling tie line, and torus make ' up line
would be manually isolated and that station procedures would specify the
requirements for manual isolation if a containment isolation signal was
received. Thit was documented as acceptable by NRC:NRR in letters to the
licensee cated December 18, 1979 and April 3, 1980. However, an inspector
identified that manual isolation of these lines with qualified valves is
not possible. Any valve which is used for primary containment isolation
must meet Seismic Class I (FSAR section 12.2) and applicable 10 CFR 50,
Appendix J, containment leakage testing criteria. Further, if manual
operation of a valve is required to effect containment isolation, the
isolation point for the valve must also be accessible under those condi-
tions which make its use necessary.
In response to the inspector's questions, the licensee re-evaluated their
response to the IEB 79-08 and TMI Action Plan Item II.E.4.2, and concluded
that isolation of these lines is assured by the .use of Seismic Class I
check valves. The licensee also agreed that isolation for the RBCCW
supply .line, instrument air line, RHR to spent fuel pool cooling tie line,
and torus makeup line cannot be performed by manual valve closure. The
RBCCW return line from the drywell can meet the isolation valve criteria
with MOV-4002 which is seismic class I, local leak rate tested and can be
closed by a control switch located in the main control room. The licensee
subsequently submitted a supplemental response to IE Bulletin 79-08 and
TMI Action Plan Item II.E.4.2 on October 24, 1984 correcting the previous
response. The inspector reviewed the supp'emental response and verified
that the contents were consistent with the conclusions drawn from the
licensee's re-evaluation and the FSAR. Both RBCCW supply line and instru-
ment air line are considered Class C lines in Section 7.3 of the FSAR
since they penetrate containment but have no interaction with the primary
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containment free space or the reactor vessel. According to the original
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design criteria, a single check valve is provided to attain isolation for
a Class C line. These check valves are seismic class I and local leak
rate tested. The inspector reviewed the results of local leak rate test
data for these check valves which were performed on June 12 and July 26,
1987 and found no discrepancies. The torus makeup line is identified as
Class B in Section 7.3 of the FSAR. The torus makeup line is non-essen-
tial and ties the condensate transfer system into the RHR test line, which
penetrate primary containment and ends below the torus water level. Ft e
water-sealed Class B lines such as the torus makeup system, the original
plant design bases allow one isolation valve in addition to the water seal
to meet isolation requirements. Also, the Safety Evaluation by the NRR on
Appendix J Review indicate that Type C testing is not required for valves
in lines which terminate below the level of the suppression pool. As for
the RHR to spent fuel pool line, the licensee revised the operating pro-
cedures 2.2.85, Fuel Pool Cooling and Filtering System, prohibiting the
use of the RHR to spent fuel pool lines except in cold shutdown. The
inspector had no further questions. This item is closed.
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(Closed) Inspector Follow Item (IFI 87-27-02) - Cracking of Surge Ring
Brackets in Large GE Motors
v On July ' 2, 1987, IE Information Notice 87-30, Cracking of Surge Ring
Brackets in large GE motors, was issued. . The purpose of the notice was
to alert recipients of a potential for failure of. surge ring brackets and
cracking of felt blocks in l a rge , vertical electric motors' manufactured
by General Electric Co. Felt blocks are used in large electric motors to
. keep the windings separated where they loop back at the end of the . stator.
The blocks are attached to a surge ring that is held in place by L-shaped
surge ring brackets welded to the surge ring and bolted to the motor cas-
ir g. . Failure of these surge ring brackets and cracking of the felt blocks
allows movement and wear of the end-turns, leading to a reduction in
insulation resistance and possible - motor failure. In addition, broken
pieces of the surge ring bracket may enter the space between . the stator
and the rotor, resulting in electrical or mechanical motor degradation.
Following an investigation to determine the applicability of the subject
notice to the Pilgrim Station, the licensee found that RHR, core spray,
and recirculation pump motors were potentially . af fected. RHR and core
spray pump' motors were overhauled on site by GE under contract with the
licensee in 1986. The surge ring brackets were not inspected during the
overhaul. However, small cracks were found on the "A" and "C" RHR pump
motor winding felt blocks. The amount of cracking found was dispositioned
by GE to be acceptable and a normal phenomenon found in form-wound motors.
On July 27 through August 5, 1987, GE performed a surge ring bracket
inspection of the RHR and recirculation pump motors using a boroscope with
the motors in place. The inspection of the RHR motors (A thru D) revealed
absence of cracks on the surge ring brackets. During the inspection of
the "B" recirculation pump motor, it was noted that the recirc motor surge
ring bracket construction is of the bolt and stud design, whereas the RHR
and core spray motor brackets are of the L-shaped design. The L-shaped
design configuration is known to have the potential of cracking, accoroing
to the IE Notice 87-30 and the GE letter to the licensee dated
July 14, 1987.
During the week of October 26,1987, "B" core spray pump motor was dis-
assembled and the surge ring brackets inspected by G.E. Due to the geo-
metry of the core spray pump motor internals, there is limited access for
the bore scope, therefore, this inspection could not be accomplished with-
out partial disassembly of the estor. It was verified that the design had
12 brackets per surge ring and two surge rings for the top end turn assem-
bly and two surge rings for the bottom end turn assembly. None of the
brackets had indications of cracking. The licensee scheduled the inspec-
tion of tne "A" core spray pump motor during the next outage because of
scheduling conflicts. The licensee indicated that based on the inspection
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results of the RHR and "B" core spray pump motors, postponement of the "A"
core spray pump motor inspection is justified. The licensee also added
that the number of operating hours and starts are similar between the A
and B core spray pump motors since both core spray systems' testing and
surveillance requirements are similar. The inspector had no further
questions. This item is closed.
(Closed) Unresolved Item 87-45-05 - Failure to Issue Licensee Event
Reports
In inspection report 50-293/87-45 the NRC identified three engineered
safety feature actuations which appeared to be reportable under 10 CFR
50.73 but had not been reported by the licensee. The licensee reviewed
the three actuations, agreed that they should have been reported and
agreed to issue License Event. Reports (LER) to document the occurrences.
In addition the licensee agreed to perform a review of previous actua-
tions to determine if any additional reports were needed.
During this inspection period the licensee's compliance section conducted
a review of all Failure and Malfunction Reports (F&MR) issued from April
1986 through the present. This review identified four F&MRs that fit the
description of an ESF actuation under the current BECo interpretation of
NUREG 1022. The licensee will submit LERs to document the following ESF
actuations at a later date.
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4/28/87 Initiation signal to both Emergency Diesel Generators (EDG)
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6/7/87 Actuation of Reactor Building Isolation and Standby Gas
Treatment System start signal
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9/17/87 Auto start of "A" EDG
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10/6/87 Reactor Water Cleanup and Shutdown Cooling System Isolation
These LERs will be reviewed upon issue as part of the normal resident
inspection program. The inspector has reviewed the licensee's actions in
addressing open item 87-45-05 and is satisfied that those actions were
thorough and timely. This item is closed.
3.0 Routine Periodic Inspections
The inspectors routinely toured the f acility during normal and backshif t
hours to assess general plant and equipment conditions, housekeeping, and
adherence to fire protection, security and radiological control measures.
Inspections were conducted between 10:00 p.m. and 6:00 a.m. on January 17,
18, and 19, 1988 for a total of four hours and during the weekends of
December 12, 19, 27, 1987 and January 3, 9, 17, 1988 for a total of 17
hours. Ongoing work activities were monitored to verify that thty were
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being conducted in accordance with approved administrative and technical
procedures, and that proper communications with the control room staff had
been established. The inspector observed valve, instrument and electrical
equipment lineups in the field to ensure that they were consistent with
system operability requirements and operating procedures.
During tours of the control room the inspectors verified proper staffing,
access control and operator attentiveness. Adherence to procedures and
limiting conditions for operations was evaluated. The inspectors examined
equipment lineup and operability, instrument traces and status of control
room annunciators. Various control room logs and other available licensee
documentation were reviewed.
The ir.spector observed and reviewed outage, maintenance and problem inves-
tigation activities to verify compifance with regulations, procedures,
codes and standards. Involvement of QA/QC, safety tag use, personnel
qualifications, fire protection precautions, retest requirements, and
reportability were assessed.
The inspector observed tests to verify performance in accordance with
approved procedures and LCO's, collection of valid test results, removal
and restoration of equipment, and deficiency review and resolution.
Radiological controls were observed on a routine basis during the report-
ing period. Standard industry radiological work practices, conformance
to radiological control procedures and 10 CFR Part 20 requirements were
observed. Independent surveys of radiological boundaries and random
surveys of nonradiological points throughout the facility were taken by
the inspector.
Checks were made to determine whether security conditions met regulatory
requirements, the physical security plan, and approved procedures. Those
checks included security staffing, protected and vital area barriers,
personnel identification, access control, badging, and compensatory
measures when required.
a. Surveillance Testing
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Diesel Generator Prelube Pump Failure
On December 13, 1987, the prelube pump for the "B" emergency
diesel generator (EDG) failed to restart on demand during a
routine surveillance test. Upon disassembly it was identified
that a small piece tf metal had become lodged between the pump
rotor and idler gear. The interference from the metal caused
the pump motor breaker to trip on pump start. An identical
failure occurred during a loss of offsite power event on
November 12, 1987. In that case the failure caused a lengthy
delay in returning an idle diesel to service. While not
required for diesei operation, the prelube system reduces EDG
bearing wear during equipment start.
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In response to the failures, the licensee drained and. inspected
the lube oil sump, and disassembled and inspected the lube oil
filters, strainers and heater. The lube oil heater _ was found
to have failed in the energized mode resulting in significant
carbon deposits in the heater and filter. No appreciable
deposits were found in the. lube oil sump. In addition, a piece
of filter element ' packaging material was found in the lube oil
filter housing. No foreign material which could have- contrib-
uted to' the prelube pump failure, however, was found. The pump
was replaced and the diesel was returned to service. No addi-
tional failures occurred during the inspection period. The two
pumps which failed had in-sequence serial numbers. Licensee
Quality Control personnel performed magnetic particle and dye-
penetrant testing of the internals of a third in-sequence pump
in the warehouse. No flaws were noted. The licensee is pursu-
ing the root cause of the failures in cooperation with the pump
vendor, Viking Pump. The licensee stated at the exit interview
that the "A" EDG prelube pump and lube oil heater would be
inspected during the next "A" diesel outage. The inspector will
continue to monitor licensee followup to this problem.
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Steam Testing of the High Pressure Coolant Injection and Reactor
Core Isolation Cooling Systems
The licensee completed full pressure steam testing of the High
Pressure Coolant Injection (HPCI) and Reactor Core Isolation
Cooling (RCIC) system turbines by utilizing te.nporary oil fired
auxiliary boilers as a source of non-nuclear steam. The full
pressure steam testing is part of a post-maintenance and system
operability check. Both HPCI and RCIC systems were overhauled
during the current outage. Utilizing temporary test procedures
TP 87-198 and TP 87-199, the HPCI/RCIC testing included turbine
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overspeed trip, pump full flow capacity and operation from the
alternate shutdown panels. Also during the test, the suction
path was changed from the condensate storage tank to the torus
and back.
During the testing, several problems were identified by the
licensee in both HPCI and RCIC syste:ns. In HPCI, problems with
the governor control system were noted including a minor oil
leak in the servo-motor. Steam leaks at gauges and turbine
drain line were also discovered. In RCIC, the licensee dis-
l covered a previously installed blank flange in the turbine steam
leak off line which caused steam leaks. A few problems were
also noted on the RCIC governor control system. The licensee is
in the process of dispositioning these items. The inspector
noted that using non-nuclear steam for the testing enabled the
- licensee to discover problems which may not have been easily
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identifiable using nuclear steam due to the radiological condi-
l tions. The inspector will review the results of the tests and ! dispositioning of the problems identified during the tests. l
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Incorrect Installation of Fire Dampers
On December 17, 1987, during performance of a. routine surveil-
lance test the licensee' inadvertently actuated two fire dampers.
One of the dampers failed to fully close - due to- interference
with a hook used to secure it in the~ open position. When the
fusible link was energized, the metal damper retaining strap
should. have fallen away allowing full closure. The hook attach-
ing the strap to _the fusible link was oriented -with the open
side toward the damper. The damper caught on the hook and re-
mained partially open. Upon discovery the licensee .immediately
stationed fire watches at all areas containing suspect' dampers.
Inspections were promptly conducted and it was identified that
all of the installed hooks were oriented in this manner. The
hooks were repositioned so that the open side faces away from
the damper. Three dampers were inaccessible and compensatory
measures-remain in place pending inspection.
The dampers were originally supplied to the licensee without the
hooks. A revision to the plant design change (PDC) package
added the hooks to facilitate surveillance testing. Installa-
tion instructions contained in the PDC specified hook orienta-
tion with the open side toward the damper. The vendor data
sheet supplied by Air Balance Inc. also showed the hoox instal-
led in this manner.
Licensee event report (LER) 87-020-00 was issued describing the
problem and corrective actions taken. The LER states that pre-
. liminary licensee assessment of the issue determined that it did
not. meet the reporting threshold of 10 CFR Part 21. The inspec-
tor discussed the Part 21 reportability with the licensee's
Nuclear Engineering Department (NED). NED personnel stated that
the failure mechanism wn created by the licensee when the hook
was added. .In addition the presence of mitigating factors such
as fire detection and suppression, and control of combustible
materials support the conclusion that a substantial safety
hazard did not exist. The licensee also feels that LER 87-020-
00 contains sufficient information to clearly define the
problem. The inspector had no further questions in this area.
The inspector examined two dampers in the cable spreading room
to verify that the hooks had been reoriented. Both hooks had
been modified, however, neither of the dampers had locking rings
installed at the hook to retaining strap connection as required
by the installation instructions in the PDC. The licensee
reviewed the function of the locking rings and concluded that
they were not required. A change to the PDC was initiated to
delete the ring. The inspector had no further questions.
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b. Radiation Protection and Chemistry
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Locked High Radiation Area Access Control
During the period covered by inspection report 87-57, four
instances occurred in which the licensee failed-to' properly con-
trol access to areas that had been designated as locked high
radiation areas. In three of these cases, doors to. locked high
radiation areas were found closed but not locked and in the
fourth case a door into a locked high radiation area was found
to not be on the list of doors that were being controlled under
the locked high radiation area door procedure.
On December 15, 1987, a contract painter failed to check that
the door to the locked high radiation area he was exiting was
properly latched. The unlatched door was identified during the
next routine check of high radiation area doors. Licensee par-
sonnel immediately latched the door and initiated a radiological
occurrence report (ROR) to document the occurrence and track all
actions taken during the investigation. Surveys of the area
showed no dose rates greater than 1000 millirems per hour
(MR/hr). Interviews with the individual involved determined
that the procedures and requirements were well understood and
that the HP technician had informed them of their responsibil-
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ities prior to entry into the area.
On December 27, 1987, and again on January 8,1988, instances
similar to the one described above took place. In both cases
the licensee initiated RORs and took steps to determine: 1)who
had been in the area, 2) were they aware of the procedure, and
3) had they been properly briefed. prior to entry into the areas
involved. In both of these cases the root cause has been deter-
mined as personnel error.
4 In one instance the licensee identified that one of the multiple s
doors into an area classified as a locked high radiation area
was not on the list of doors to be checked on a routine basis.
The door was immediately checked and found to be locked. Records
have been audited to determine if any unauthorized entry into
the area had occurred and no instances were identified. The
door has been placed on the list and is now routinely checked.
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The inspector reviewed licensee actions as a result of these
instances and is satisfied that in all cases, the immediate and
followup actions were timely and complete. Surveys taken were
comprehensive and conducted almost immediately af ter discovery
of unlocked areas. Dose calculations were performed and
dosimetry read in all cases. Involvement by senior HP and plant
management was evident in all instances.
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Inadequate control of locked high radiation areas .has been an
area of longstanding NRC concern. Notices of Violation have
been issued in the past, during inspectior,# 50-293/87-03,
50-293/87-11', . and 50-293/87-19 - which addressed these concerns.
'In regard to these violations the licensee instituted corrective
actions which have been successful in addressing segments of the-
-problem but have not been successful-in preventing recurrence of
events involving high radiation area door control.
The inspector has independently reviewed the licensee's program
for control of high radiation areas and high radiation area key
control and has found them adequate. Although the programs
themselves are adequate and personnel have been trained on those
programs, instances still occur where locked high radiation
areas are not adequately controlled.
Based on review of these four instances coupled with the review
of Unresolved Item 87-50-08, the inspector determined that the
licensee actions in response to these previous findings have not
prevented ' recurrence, Failure to comply with the requirements
of Technical Specification 6.11 and Implementing Procedure
G.1-012 is an apparent violation of NRC requirements as docu-
mented in Appendix A of the cover letter to this report
(87-57-01). Licensee response to Appendix A should include
those measures taken to insure that corrective actions are
effective and lasting.
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Contaminated Clothing Offsite
On December 17, 1987, at 7:26 p.m. hours a Bechtel pipefitter
who-was exiting the reactor building, set off a whole body por-
tal monitor alarm. The portal monitor indicated contamination
of his chest area and left hand. The health p5ysics technician
on duty at the access point removed the individual from the por-
tal monitor and began performing a survey using a RM-14 with DT
260 probe. The HP technician identified; 1) contamination on
the individual's left hand, 1-2 thousand dpm per 100 square
centimeters (K OPM), which was removed by washing, 2) contamina-
tion on the shirt in both the chest (80K OPM) and lower stomach
area (1K DPM). The shirt contamination was removed by tape (80K
OPM) and washing with soap and water (1K DPM). The employee,
now wearing an undershirt and trousers, was then sent to clear
the portal monitor which again alarmed and indicated contamina-
tion in the chest area. The HP technician again surveyed the
individual and identified contamination on the undershirt in the
chest area (70K DPM). The individual was then sent into the
portal monitor bare chested and was cleared. The individual was
given his outer shirt, which was still wet from decontamination
and cleared through portal monitor. At this point, the indi-
vidual removed the wet shirt, put on his jacket, cleared the
portal monitor again, and left for his home.
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Upon returning to work December 18, 1987, the individual was
given a whole body count to determine if any internal contamin-
ation had occurred. The whole body count showed no internal'
contamination. After completion of the whole body count the
individual was interviewed to determine how he had been contam- '
inated, where the occurrence took place and how long he was
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contaminated prior to detection, to calculate skin dose received.
The interview revealed that the individual had been contaminated
when he disconnected a partially pressurized service air hose
and . depressurized it. The interview also revealed that the
individual used the portal monitor at the 91 ft, elevation of
the reactor building, received an alarm, did not call for HP
assistance but instead tried to decontaminate himself prior to
proceeding to the reactor building access. Station procedures
require that an individual who finds himself contaminated is to
call health physics for assistance. The individual stated that
he was aware of this requirement. During the interview the
individual expressed concern about whether his heavy winter
jack ' could have shielded the contamination on his shirt and
undershirt from detection by the portal monitors. To demon- ,
strate that this could not happen, a HP supervisor placed
plastic bags, which contained the contamination removed from his
shirt, inside the coat and attempted to exit through two por-
tals. The portal monitors alarmed on each attempt. The indi-
vidual appeared satisfied with the demonstration put his jacket
back on, with the plastic bags removed and attempted to leave
the reactor building. An alarm was actuated on the portal
monitor and contamination was indicated on the left arm. The
on duty HP technician removed the individual from the portal
monitor and identified 3K DPM contamination on the upper right
sleeve (outside) of the jacket even though the jacket had not
been worn into the reactor building. At _this juncture the indi-
vidual expressed concern over whether the shirt that he had worn
the previous day could still be contaminated. The licensee had
a HP technician accompany the individual to his home. The
individual's shirt was found to be contaminated, was bagged and
returned to the site. Surveys of the individual's home and
vehicle identified no further contamination.
Efforts to determine how the contaminated shirt was worn through
the portal monitors without setting of an alarm yielded positive
results. The individual stated that he had purposely kept him-
self away from the portal monitor in an attempt to keep his wet
shirt away from his skin. The licensee taped the plastic bags,
with the contamination in them, back onto the snirt and an HP
supervisor attempted to pass through the portal monitors by
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mimicking the' body posture used by the individual _when he cicared
the monitor. The HP supervisor was able to pass through six
different monitors without setting off an alarm. The HP super-
visor then used the portal monitors in the correct manner and
all six monitors alarmed proving that the equipment was func-
tional.
The licensee has evaluated the occurrence to identify the root
causes ~ and -immediately implemented corrective action. This
occurrence was _ caused by one sequence of events that involved
two distinct personnel = errors. The primary cause involved the
failure of the HP technician to perform an adequate survey of
'
the contaminated individual's clothing when the portal monitor
alarm was received. The second problem involved the f ailure to
properly use the installed portal monitors at the reactor build-
ing access.
In addition to personnel interviews.to identify the sequence of
events the licensee also reviewed procedural adequacy, personnel-
training and portal monitor calibration and performance. These
F- reviews verified that training was adequate and portal monitor
performance'was as designed. Procedures for control of contam-
inated individuals at the reactor building access did not spec-
ifically require that all articles of clothing require a 100%
frisk prior to this occurrence. Instructions have been posted
at the reactor building access which now clarify the procedure
to be followed when an individual is found to be contaminated.
The portal monitors in use at Pilgrim do not presently have a
switch at chest level. which must be actuated to start the moni-
toring process. Lack of this feature allowed the individual
.,
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wearing a contaminated shirt to lean away from the machine suf-
ficiently to clear the monitor without any alarm. The licensee
has determined thats the manufacture of the portal monitor now
produces a chest high switch for the installed model and will
install them in the future.
Calculations have been performed by the licensee to determine
the radiation dose received by the individual and the amount of
,
radioactive material that was released from the site on the con-
taminated shirt. The results of these calculations show that
the individual received a localized radiation dose to the skin
of 260 Mrem, which is below the federal limits for skin exposure,
and that the amount of radioactive material on the individuals
clothing was 0.2 microcuries which meets the federal criteria as
an exempt quantity of Co-60. The inspector is satisfied with
the licensee's analysis and corrective actions and has no
further questions.
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Allegation of I'mproper Disposal of Radioactively Contaminated
Shrubs . ( RI-87-A-0107)
On August 31 and September 11, 1987, the NRC resident office at
t
Pilgrim received. allegations that radioactively contaminated
shrubs had been removed from the site and improperly disposed.
.The alleged improper disposal occurred on ' July 23, August 26 ar.d
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August 28,1987. During this time period the licensee removed a
large number of shrubs from various areas of the site, including
those - planted near the old. administration building and the
switchyard. The shrubs were removed to facilitate site con-
struction activities _and to alleviate certain security' concerns.
Upon receipt of .he first allegation on August 31, 1987 the NRC
requested that the licensee perform -an evaluation and provide
,
the results .for review. In addition an independent NRC review
was initiated.
Resident and specialist inspectors reviewed the licensee's con-
clusions. The licensee evaluated material release records and
interviewed personnel regarding removal of shrubs -during the
!.. week of July 20,1987. Several truckloads of shrubs that 'were
transported offsite during the midnight shif t on July 24 were
! examined in de tail. Because trace amounts of Cobalt-60 had pre-
i- viously been found in soil onsite, some .of the shrubs had the
soil removed from the roots prior to release. Each shrub was
hand surveyed and found to meet established offsite release
c ri te ri a . They .were transported first to the licensee's shore-
front area and later to a dump site on licensee property. The
licensee concluded that the shrubs had been adequately surveyed
and that no radioactive material had been improperly released.
The resident inspectors reviewed the licensee's program for con-
trol of release of material from the site. This area was also
evaluated by NRC specialist inspectors during inspection 50-293/
87-19. Both inspections concluded that appropriate surveys and
release liniits have been established and implemented. Resident
and specialist inspectors examined licensee release records for
the dates in question to verify that vehicles leaving the pro-
tected area had been properly surveyed. No discrepancies were
identified. An NRC resident inspector accompanied by a licensee
representative collected four samples of the shrubs which had
been deposited in the dump site discussed above. Each of the
four samples consisted of root, branch and foliage clippings
from a number of different shrubs. The samples were indepen-
dently analyzed by the NRC. Three of the samples indicated no
contamination. One sample indicated only trace levels of Cobalt
-60. Measurements showed that the amount of C0-60 present in
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this sample was about 2% of the average radioactivity typically
found in soil due to naturally occurring isotopes.
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The . licensee's program for release of material from the site
appears adequate. Appropriate survey . techniques and' release
limits-have been established. Review of -records confirmed that
the program is being implemented. Samples of the shrubs col-
.lected by the NRC showed' zero or negligible contamination and
pose ' no health and safety concern. Based on the above this
allegat'an is considered closed. NRC Region I staff provided
status briefings concerning this allegation to. Senator Kennedy's
staff. and to the Massachusetts Department of Public Health.
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Allegation of Airborne Radioactivity in the Trash Compaction
Facility (RI-87-A-0120)
On October 5, 1987, the resident office received an anonymous
allegation that personnel working at the sort table in the trash
compaction ; facility .(TCF) were being routinely exposed to air-
borne radioactive contamination. The alleger stated that the
two filter systems designed- to treat exhaust air from the sort -
table prior to discharge into the room were not functioning, and
that the filter differential pressure alarm circuits had been
disabled.
On October 7 and 8, 1987, NRC specialist inspectors toured the
TCF and examined the design and condition of the equipment. The
sort table is used to separate contaminated materials for com-
paction and disposal. Potentially contamitated air is exhausted
from the table, passed through two filters operating in parallel,.
and released into the room. Airborne radiation levels in the
room are measured by means of a separate air monitor which is
operated whenever the sorting table is used. The alarm is
<
typically set at 3 X 10 -10 (3E-10) microcuries per cubic cen-
timeter (cc). In addition the filters are surveyed daily and
changed if contact dose rates exceed 2mR per hour. The inspec-
tors also examined the trash compaction unit in the area and
found that similar controls had been applied. Based on the
above, no immediate health and safety concerns were indicated.
On January 15, 1988, the resident inspectors toured the TCF,
examined equipment operation and interviewed licensee and con-
tractor personnel involved in ongoing work activities. A radia-
tion work permit specifying protective clothing, health physics
coverage, and use of a continuous air monitor is in place to -
control work at the sort table. Personnel involved stated that
trash bags were surveyed prior to sorting and rejected if radia-
, tion levels exceeded Smr/hr, if they contained liquid, or if any l powdery material was present. The health physics technician on
auty stated that filter radiation levels are monitored daily.
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Workers and health physics personnel also' stated that filter:
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-differential 1 pressure (dp) instruments are monitored to detect
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filter plugging, however no one had.been clearly assigned this
responsibility and no' dp -limit was. established. The inspector-
observed the operation of the table and noted that .the '"filter
restricted" alarm ' actuated - for one of the two filters. The
alarm actuated ' for the filter displaying - the -lower differential
pressure.- When questioned workers stated ' that much of the -
monitoring and alarm circuitry for the table was not functional,
and that - the filter alarm was. not reliable. The . table was
originally part of a larger processing system and much .of the
disconnected circuitry was intended to perform functions which
are no longer needed. The f aspector verified that current
filter dp readings are consistent with the manufactures name
plate data.
;It appears that the general-process applied, incibding insoec-
tion ana. survey of trash bags' prior to sorting, daily filter
surveys . and continuous air monitoring would preclude airborne
radioactivity problems. Based on the above this : all.egation is
closed. However, th'e inspector noted that no work instructions
existed-describing the controls applied and equipment monitoring
requirements. When discussed with licensee radiation protection
management they promptly committed to review the situation and-
issue appropriate guidance. This was confirmed during the
inspector's exit interview. -
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Erosion of Construction Dirt'into Wetland
On January 15, 1988, at 5:45 p.m. the licensee made an ENS
notificatfor in accordance with 10 CFR 50.72 (b)(2)(vi) which
requires the licensee to inform the NRC of an event or situation
related to health and safety of public for.which a news release
.was made or notification of another government agency has been
made. During routine environmental monitorin'g, the . licensee
observed erosion from a pile of construction dirt into an adja-
cent licensee controlled wetland. The Plymouth Conservation
Commission and the Massachusetts Department of Public Health
were notified and the press release was made by the licensee.
Also on January 16, 1988 two representatives from the Plymouth
Conservation Commission toured the area.
In the last several years during onsite excavation for plant
modifications, dirt, asphalt and concrete containing low le"els
of contamination were stored in a fenced in storage area outside
the protected area on the licensee's property. The licensee
estimated that the storage area contains 110,000 cubic feet of
material. Before removal from the protected area, samples of
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material were obtained and isotopic analyses was performed by
the licensee. The activity found was - reasonably uniform at1
levels of 10(1E-6) and 10(1E-7) microcuries of Cobalt-60 and
Cesium-137.per gram. Sampling and- storage of this material was
previously reviewed during inspection 50-293/87-18. On
January 21, 1988 the inspector toured the area, accompanied by.
a licensee health physics technician, and performed a survey of
'the storage area and found no detectable radiation above .back-
ground levels. During the tour the inspector noted that bales
of hay had been put around the perimeter of the fence which
borders wetlands area to prevent further erosion . of material.
The fenced in storage area was secured with a locked gate. The
inspector's survey of the area and review of licensee's analyses
indicate that the level of activity does not represent a health
or safety concern. However, the inspector raised a concern to
the licensee management that the material should not be -allowed-
to erode. The inspectors will continue to monitor the licensee
actions in formulating long term solution to properly dispose of
the material.
c. Fire Protection
On January 17,1988, at 4:55 a.m. the control room received a report
from a security guard of smoke coming from a contractor lavatory
trailer, which is located adjacent to. the Bechtel warehouse inside'
the protected area fence. The onshift fire brigade chief was dis-
patchtd to the scene and confirmed smoke and smoldering in the area.
The fire brigade was immediately dispatched and fire was extinguished
using a portable dry chemical extinguisher and a hose from a nearby
hydrant house. Electrical maintenance was called to shut off the
power to the trailer. By 5:30 ~ a.m. , .the fire brigade members had
cleared the scene and a continuous fire watch was posted ii, the area.
' The cause of the fire was believed to be overheating of an overhead
heating unit for the trailer. No personnel injury . occurred. The
inspector toured the scene with a licensee fire protection engineer
on January 18, 1988. Minor damage to a small area of the ceiling in
the trailer was observed. The Plymouth Fire Department was notified
by the licensee in the morning of January 18, 1988.
4.0 Review of Plant Events
The inspectors followed up on events occurring during the period to deter-
mine if licensee response was thorough and effective. Independent reviews
of the events were conducted to verify the accuracy and completeness of
licensee information.
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a. Spurious Isolations of RHR Shutdown Cooling System
On December 7,1987, at 2:28 p.m. , an inadvertent isolation of both
inboard and outboard containment isolation valves on the RHR shutdown
cooling suction line occurred. Preparation for the reactor vessel
hydrostatic test was in progress. As part of the hydrostatic test
procedure, a technician was installing an electrical jumper in the
primary containment isolation system logic panel C-941 to bypass the
reactor coolant system (RCS) high pressure interlock on the inboard
isolation valve. When the termination screws were loosened to in-
stall the jumper, the leads lost contact and caused a false high
pressure isolation signal. RHR was in its shutdown cooling mode when
the isolation signal was generated, and the shutdown cooling suction
valves (MOV 1001-47, 1000-50) automatically closed as designed.
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Coincident with the closure of the valves, the "A" and "C" RHR pumps
tripped automatically to protect the pumps from loss of adequate
suction. The licensee determined the actuation was due to a person-
nel error. The licensee revised Procedure 2.1.8.1, Class I System
Hydrostatic Test, to caution the I&C technician of potential isola-
tion of RHR shutdown cooling system while installing the jumper.
On December 8, 1987, at 9:45 p.m. , the inboard isolation valve (MOV
1001-50) on the RHR shutdown cooling suction line automatically
closed. The automatic isolation occurred when the plant reached
100 psig during pressurization for performance of the class I hydro-
static test. The outboard isolation valve (MOV 1001-47) was already
closed to form a pressure boundary for the test. The licensee's
investigation determined that the cause of the isolation was that
Procedure 2.1.8.1 did not identify all the jumpers necess4ry to
b3 pass the RCS high pressure interlock on the inboard isolation
valve.
As immediate corrective action, the licensee halted the pressuriza-
tion of RCS and reviewed the logic prints. The licensee revised
Procedure 2.1.8.1 to reflect the need to install an additional jumper
in panel C-942. In reviewing this event along with other similar
events documented in previous inspection reports, the inspector noted
that inadequate planning and inadequate procedures appear to be a
common root cause for several ESF actuations which occurred on
September 17, september 22, October 15 and October 24, 1987. The
inspector expressed this concern at the exit meeting with licensee
management. The licensee informed the inspector that the Technical
Group is in the process of developing generic guidance for isolating
or jumpering an electrical component which may cause inadvertent
safety system actuations. The inspector will continue to monitor the
effectiveness of licensee's corrective action to prevent further ESF
actuations due to inadequate planning and inadequate procedures.
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b. Reactor Water Cleanup System Spurious Isolation
On December 17, 1987, at 11:05 a.m., the inboard primary containment
isolation valve on the reactor water cleanup (RWCV) system suction
line automatically isolated. I&C technicians conducting a routine
surveillance of the RWCL' high area temperature isolation logic inad-
vertently grounded a lead which had been lif ted during the test.
Grounding the lead resulted in a blown logic power fuse and isolation
of the valve (MOV 1201-2). Following investigation by the control
room supervisor, the fuse was replaced and the isolation was reset.
The licensee's investigation concluded that the root cause is a per-
sonnel error. The licensee informed the inspector that the proced-
ure, 8.M.2-1.2.2, Reactor Water Cleanup Area High Temperature, will
be revised to provide cautions to the control room operators and the
I&C technicians. Also, an effort is ongoing to review recent ESF
actuations caused by personnel error to formulate appropriate
corrective actions.
c. Engineered Safety Feature Actuations Due to a Failed Logic Relay
On January 6, 1988, at 2: 50 p.m. , the coil of primary containment
isolation system (PCIS) electrical relay 16A-K57 failed, creating a
f aul t and resulting in blown logic power fuses. The aeenergization
of this portion of the PCIS logic caused a partial primary contain-
ment isolation along with a reactor building isolation and start of
the "8" Standby Gas Treatment system (SBGT). The licensee notified
the NRC at 5:12 p.m. via ENS. The failed relay was a GE type CR120A
relay. The licensee has experienced several failures of this type of
relay in the last few years. The licensee's evaluation of this nigh
failure rate and corrective actions to address it are described in
the inspection report 50-293/87-50.
On January 7, 1988, the inspector reviewed maintenance request (MR)
88-9 which had been initiated to investigate the cause of the above
mentioned ESF actuations and to replace the blown fuse and the faulty
relay. The inspector noted that the relay replacement was performed
using only procedure 3.M.1-11, Routine Maintenance. This procedure
contains general guidance and its stated use is for performing main- '
tenance activities which are not complicated or critical enough to
require detailed written procedures. In this case, no step-by-step
instruction was initiated to control the sequence of work, to control
and tag lifted leads and jumpers, and to ensure verification and
independent verification of system restoration. A similar problem
involving lack of a sufficiently detailed controllir g procedure and
the appropriate reviews during an electric relay replacement on
November 24, 1997 was the subject of a violation as documented in the
inspection report 50-293/87-50, The licensee informed the inspector
that the corrective actions to address the violation are being
formulated and will be submitted to the NRC.
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d. Spurious Reactor Protection System Actuation
On January 17, 1988, at 1:13 a.m., a spurious reactor scram signal
was generated during the performance of a reactor level instrument
calibration. The full scram signai on low water level was received
due to a disturbance in the reactor water instrument line when an I&C
technician was valving a level instrument (LI-263-59A) back sin ser-
vice. The Rosemount level transmitters (LT-263-57 A&B) which initi-
ated the. scram signal are on. the same ~ instrument rack. The licen-
see's preliminary investigation indicated that the root cause of the
event is attributed to a combination of personnel error and inade-
quate procedure. The investigation also identified that' the level
instruments (LI-263-59 A&B) were incorrectly installed in that the
sensing lines were reversed. The new Barton level instruments
(LI-263-59 A&B) were recently installed during this outage and would
only be used for local indication during a shutdown from outside the
control room. The licensee is currently reviewing the plant design
change (PDC 85-07) records and post-installation test data to deter-
mine the cause. Surveillance test records are also being reviewed
by the licensee. This item -is unresolvea pending the completion of
the licensee investigation (87-57-02).
Upon receiving the spurious scram the control rnom staff noted that
scram discharge i' :trument volume (SDIV) vent valve CV302-238 primary
containment vent and purge valves A050448 and A05035B and one of two
redundant secondary containment isolation dampers in each line did
not close. In addition the "B" standby gas treatment system (SGTS)
did not start. Based on the initiating event, these components
should have actuated. The licensee notified the NRC of the failures
via ENS at 5:00 a.m. on , January 17, 1988.
l The control room staff conducted an immediate critique with available
I&C personnel, and documented observations for management followup.
Later on January 17, the licensee inspected the physical condition
l of the SDIV vent and drain valves and noted paint on the stem of
CV302-238. The paint was removed and the va'.ve successfully stroke
timed. The licensee htid a second critique with management repre-
sentatives on the morning of January 18, 1988 to assess the situa-
tion. Subsequently, a walkdown of involved isolation logic components
was performed to verify relay contact configuration cnd to identify
any jumpers or lif ted leads. This walkdown was performed to the
extent possible without disturbing components. No discrepancies were
noted. Early on January 19, the licensee performed a test in which a
reactor scram was intentionally initiated. The same equipment failed
to actuate as during the January 17 scram. Based on this licensee
management stopped all work on the affected components. A task force
composed of members from the technical staf f, systems group, I&C and
operations was designateo to investigate tne incident. This team
reviewed available information and developed an action plan.
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Walkdowns of the air system piping and components supplying motive
air to SDIV vent valve CV302-238 were performed to verify ' hat the
as built configuration is in accordance with design doc 's and
that components are in good physical condition. No disteepancies
were identified. Valves CV302-238 and CV302-228 are supplied air by
the same solenoid operated valves. The licensee cecnergized these
solenoid valves and observed that CV302-22B closeo while CV302-233
did not. This indicates a mechanical problem with the valve or
operator. The licensee was identifying replacement parts and pre-
paring to disassemble the valve by the close of the inspection
period. The inspectors will continue to monitor licensee followup
to this failure.
Licensee review of logic drawings confirmed that the remaining equip-
ment which had not properly actuated shared common isolation logic
components. A series of survefilance tests was performed to allow
,
monitoring of key relay ac%ations. A single contact on a General
Electric (GE) HFA relay was determined to be misfunctioning. The
contact is required to close when an isolation s1gnal is received,
actuating the affected equipment. However, contact resistance
remained high with the contact closed. The relay was replaced and
the system successfully tested. The licensee contacted GE to coor-
dinate disassembly and inspection of tt.e relay. Dissassembly had not
begun by t he. close of the inspection period. The inspector will
continue to monitor licen ee investigation of this failure.
The inspector exprev.ed concern that ^hree separate equipment mal-
functions had occurred during the inadvertent actuation. This may
reflect weakness in the surveillance and post-work tect program.
However, the licensee's response to the a c'.ua ti on and subsequent
malfunctians was prompt, thorough and effec ve. Control room aper-
ators quickly recognized each of the failures. They held a critique
on the same shift with insclved personnel. Critique observations -
were clearly documented and provided to nanagement. An additional
critique with managercent present established pri( rities. Action was
taken to freeze equipment until a r. investigation pin could be
developed and implemented. Followup was well cocrdinated and in-
volved representatives of several portions of the organization. In
this case licensee commitmert to determining and correcting the
problem root cause was evident.
'
Review of _ Reactor Vessel Hydrastatic Test Procedure c,d Test Results
Duriew - 1spection period the licensee comp.eted the reactor vessel
; : est. Several reactor vessel instrument nozzles were repaired
outage, prompting performance of a hydrostatic test rather
n leakage test. The reactor vessel reached minimum test
. a all inspections were completed on December 9, 1987. Only
i -- age associated with mechanical connections, such as flanges and
ni ting was identified. The reactor vessel was depressurized on
Deco -
12, 1987 af ter completioc of excess flow check valve testing.
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The_ inspector reviewed the licensee's hydrostatic test procedure to verify
that appropriate prerequisites, precautions and instructions had 5een
: included. A sample of valve lineups was reviewed to datermine the ade-
quacy of established test boundaries. Completed valve lineups were~also
examined. -Control of temporary electrical and mechanical jumpers was
evaluated to ensure proper documentation and restoration. The-inspector
observed installed pressure instrumentation and verified appropriate range
and calibration status. The adeouacy of staffing - to support test per-
formance was periodically verified. The inspector reviewed test results
and discussed them with engineering, operations, and quality control
oersonnel to ensure that test changes were properly processed, adequate
inspections were conducted, and that inspection results were promptly
dispositioned.
\
The licensee's preparation for and execution of the test was generally
well organized and controlled. Procedures for test performance and con-
duct of visual inspections were clear and comprehensive. A detailed
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Quality Control (QC) work instruction was developed specifying components-
and piping requiring inspection. Inspection assignments were broken down
by location, elevation and component. This QC instruction also included
a series of piping diagrams depicting the test boundaries which were
utilized to assist in inspection performance and documentation. The
licensee's Technical Engineering Section, Quality Control staff and
Nuclear Engineering Department each reviewed test boundary adequacy. In-
spection results were well documented, and maintenance requests were
promptly initiated to correct identified leakage.
The licensee experienced two shutdown cooling isolations during implemen-
tation of the test procedure. These isolations are discussed in detail in
section 4.a of this report. During the test the licensee identified leak-
age past the seal ring at the stuffing box to pump casing joint on both
racirculation pumps. Leakage flow was estimated to be one to two gallons
per minute for each pump. The leakage wet the pump casings and portions
of the suction piping, and acceptable inspections could not be completed
.
in these areas. The licensee stated that similar leakage on at least one
of the pumps was noted during the last outage. That leak sealed as system
,-
temperature increased during startup. The licensee believes that the
leakage observcd during the recent test will also stop as temperature is
increasM, and no pump repairs are planned. The licanses stated at the
ins.nector's exit interview that the pump casings and suction piping will
be reinspected during start;p.
.
The inspector noted that the test procedure did not contain valve lineups
for manual instrument isolation valves within the test boundary. Many
instruments and a significant portion of instrument piping has been
replaced this outage. 'lisuai inspections were performed of clas; I piping
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downstream of these v31ves. The inspector questioned the basis for licen-
see confidence in instrument line isolation valve positions during the
test. The licensee pointed out that hydrostatic :esting of these lines
was not required dur'ng this outage. In addition excess flow check valve
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testing was conducted immediately cf ter completion of the hydrostatic test
with the system still pressurized. Successful completion of the check
valve testing requires proper alignment of the manual isolation valves,
and provides assurance that the piping was pressurized during the visual
inspections. Tre licensee however, agreed that the intent of the test had
been to pressurize and inspect this piping and that the current procedure
does not adequately assure the correct valve alignment. Licensee manage-
ment stated that the procedure would be revised to address this weakness.
6.0 Integrated Leak Rate Testing
On December 21, 1987, the licensee began performance of the primary con- $
tainment integrated leak rate test (ILRT). The containment was pressur-
ized with air to the full test pressure of 45 pounds per square inch and
maintained at this pressure for 24 hours. The 24 hour test period started
at 10:15 p.m. on December 21, 1987. A regional specialist inspector was
onsite during the ILRT to review the adequacy of the test procedure and to
observe the conduct of the test. The preliminary licensee test results
indicated a succe .sful test, with measured leakage slightly greater than
20 percent of the allowable leakage. A primary contributor to the ob-
cerved leakage was identified as a drywell pressure transmitter piping cap
which had not been fully tightened. Upon completion of the specialist
inspector's review of the ILRT results, inspection report 50-293/87-58
will be issued documenting the inspectors findings.
While preparing for the primary containment integrated leak rate test
(ILRT) the licensee observed that several torus temperature and moisture
elements were not functioning properly. Troubleshooting identified cir-
cuit faults at a torus electrical penetration assembly. The licensee
removed the penetration assembly protective cover inside the torus and
found that it was filled with water. The penetration is installed ver-
tically through the top of the torus. On both the inboard and out: 3rd
sides of the penetration a metal frame is attached on which 28 terminal
boards are mounted. Cables passing through the pe ietration, and supplying
instrumentation in the torus also landed on trese terminal boards. A
protective cover is bolted over the f rame n.u terminal boards on both
sides of the penetration. Design drawings specify that cover joints are
to be sealed with s'licone tape. The licensee stated that the protective
cover had not been oroperly sealed, allcwing water intrusion and buildup.
The water caused significant corrosion of the cable connectors, terminal
boards and metal framework. This corrosion and water buildup resulted in
the observed electrical circuit faults. Licensee inspection of the other
torus electrical penetration identified similar conditions. Temporary
repairs of the temperature and moisture elements were made to allow ILRT
performance. Cables for communications, lighting, and torus to drywell
vacuum breaker indication also run through the penetration. The penetra-
tion is not considered by the licensee to require environmental qualifica-
tion but is designated as a "Q" component. The licensee is evaluating the
root cause of the water intrusion and is developing a temporary procedure
_
to control repair and testing of the penetration. The inspectors will
^
continue to monitor licensee followup and corrective actions.
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The licensee informed the inspector that penetration repairs would not be
completed until after ILRT performance. The inspector questioned the
.effect of the planned repairs on the penetration leak tightness, and the
ability to perform adequate leakage test after the planned rework. The
licensee stated thr.t the work would not affect penetration leakage but
that-adequate testing could _ be performed after work completion. Based on
available drawings however, the licensee could not demonstrate adequate
testability. In response to NRC concern the licensee obtained the needed
drawings from the vendor and verified that the penetration was completely
testable. The inspector had no further questions.
During the ILRT, the licensee identified a water leak in the high pressure
coolant injection (HPCI) turbine room. It was determined that the in-
creasing pressure in the torus air space caused the suppression pool water
to back up through the HPCI turbine exhaust line and throgh the drain
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piping, overflowing the HPCI gland seal condenser onto the HPCI room
floor. The turbine exhaust line disenarges to the torus through a check
valve .and a locked open stop-check valve. To prevent any condensation
from collecting in the turbine exhaust line downstream of the check valve,
a drain piping drains any condensation to the HPCI gland seal condenser
through a drain pot. Two solenoid operated drain valves on the drain pot
close automatically on a HPCI (Group IV) isolation signal. This is to
provide the isolation from the torus to the gland seal condenser. The
licensee's investigation determined that leads had been lifted in the HPCI
isolation interlock logic circuit since October 30, 1987 in support of the
HPCI steam testing . utilizing temporary oil-fired auxiliary boilers. With
the HPCI isolation signal bypassed, the drain valves remained open as the
- drain pot was filleJ with the suppression pool water. The licensee sub-
sequently relanded the leads in the HPCI isolation interlock logic circuit
and the drain valves closed.
After reviewing the ILRT procedure, HPCI test procedure and interviewing
licensee personnel, the inspector concluded that licensee review of the
active maintenance requests prior to the ILRT was not thorough in that the
lifted leads controlled by the MR 87-663 were not identified. The LtR tags
were attached on the HPCI isolation logic circuit inside a logic panel and
thus the tags were not identified during a system walkdown prior to the
ILRT. The dra:n valve positions were verified by the light indications on
the control room panel 903 as prescribed in the ILRT procedure.
The inspector also determined that the maintenance <. quest above may not
be an adequate method of identifying and tracking jumpers and lifted .
leads, especially for a long term application and for components which
could affect other ongoing maintenance or surveillance. Station proce-
dures do not require temporsy modification controls ;ur jumpers and
lif ted leads which are controllea by active maintenance requests. The
inspector discussed these findings at the exit interview with licensee
, management. The licensee informed the inspector that a lifted leads and
jumper log will be kept in the control roon to aid the operators in con-
trolling lif ted leads and jumpers.
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7.0 Licensee Nuclear Organization Management Realignment
On December 14, and on December 31, 1987, the Boston. Edison Co. announced,
.as . part of a planned realignment occurring over the next several weeks,
the appointment of the following managers to -key management positions in-
the licensee nuclear organization at Pilgrim Station.
--
Mr. Kenneth L. Highfill was named to assume the new position of
Station Director. In this capacity, Mr. Highfill will oversee day
to day operation of the Pilgrim Station including plant operations,
planning and outage, nuclear training, plant- support' functions, and
administrative services. Mr. Highfill will report directly to Mrb '
Ralph G. Bird, Senior Vice President-Nuclear.
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Mr. Robert J. Barrett was named the new Plant Manager. Mr. Barrett ,
will report to Mr. dighfill, the Station Director. 1
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Mr. Roy Anderson, currently Deputy Outage Manager, was named to
assume the new position of Planning and Outage Manager. Mr. Anderson
will report to Mr. Highfill, the .itation Director.
--
Mr. Ed Kraf t was named to assume the new position of Plant Support
Manager. In this capacity, Mr. Kraft will oversee radiological,
sacurity, industrial safety and fire protection, and other station
support functions. Mr. Kraft will report to Mr. Highfill, the
Station Director.
--
Mr. Donald Gillespie, currently Director of Planning and Restart, was
appointed to the position of Quality -Assurance Department Manager.
Mr. Gillespie will assume the position after completing his Senior
Reactor Operator' training. The Quality Assurance Department Manager
reports to Mr. J. E. Howard, Vice President-Engineering.
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Mr. Frank Famulari, currently Operations Quality Control Group
Leader, was named to assume the newly created posi tion of Deputy
Quality Assurance Department Mananer. Mr. Famulari will report to
Mr. Gillespie, and be acting Department Manager until Mr. Gillespie
assumes the position after completing the Senior Reactor Operator
training.
--
Mr. John F. Alexander was named to assume the position of Operations
Section Manager. Mr. Alexander will report to Mr. Barrett, the Plant
Manager.
--
Mr. Donald J. Long was named Security Section Manager. Mr. Long will
report to Mr. Kraft, the Plant Support Manager.
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8.0 Management Meetings
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At periodic intervals during the course of the inspection period, meetings
were held with senior facility management to discuss the inspection scope
and preliminary findings of the resident inspectors. On January 26, 1988,
the inspectors conducted a final inspection exit interview to formally
present inspection findings.
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Attachment I to Inspection Report 50-293/87-57
Persons Contacted ,
* R. Bird, Senior Vice President - Nuclear
* K. Highfill, Station Director
K. Roberts, Plant Manager
R. Barrett, Deputy Plant Manager
R. Anderson, Planning and Outage Manager
E. Kraft, Plant Support Manager
F. Famulari, Deputy Quality Assurance Manager
D. Swanson, Nuclear Engineering Department Manager
J. Alexander, Operations Manager
N. Brosee, Maintenance Manager
J. Jens, Radiological Protection Manager
J. Seery, Technical Manager
R. Grazio, Fie!d Engineering Manager
P. Mastrangelo, Chief Operating Engineer
,
R. Sherry, Chief Maintenance Engineer
N. Gannon, Chief Radiological Engineer
D. Long, Security Manager
F. Woznick, Fire Protection Manager
*Sonior licensee representatives present at the exit meeting.
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- ATTACHMENT.1II
January 6, 1988
- MEMORANDUM'FOR: Ken = Roberts
. Plant Manager
FROM: . Clay Warren- -
- Senior Resident Inspector - Pilgrim
SUBJECT: FACILITY TOUR FINDINGS, DECEMBER 8,1987.
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- The items' on - the~ attachment were noted during the - facility tour on
December 8,.1987. Please-contact the Resident inspector Office when your staff
is ready to discuss the evaluation of-the-items and the status'of any actions
- taken. Please note. the items and the facility response will be. addressed in a
'
routine inspection. report.
Thank you-for your time and attention to these matters.
Sincerely,
i
Clay C. Warren
Senior Resident Inspector ,
Attachment: ;
As stated ,
cc w/ attachment: c
,
R. Blough
- . W. Kane
W. Russell
- J._Wiggins
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ATTACHMENT
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Numerous motors appear to have failed grease seals caused by overgreasing
without first removing grease drains. This condition causes a buildup of
grease and dirt in the cooling airflow path and in extreme cases grease
in the motor windings. (SBGT fans and SLC pumps)
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Nuts and bolts were noted 10ying inside an electrical cabinet in the RCIC
room.
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Multiple cases of open junction boxes, terminal boxes and conduit pulled
away from terminal boxes were noted.
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Motor heaters for the "B" RHR pump appear to have overheated causing the
insulation on the heaters to melt.
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HPCI room cooler drip pan is full of paint scrappings which could lead to
drain clogging.
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Standby Liquid Control system relief valves have boric acid c rystal
buildup which could alter setpoints.
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Painting effort should be more closely controlled to prevent painting
inappropriate surfaces, i.e., linkages, valve packing glands, trip
throttle valves, limit switches, etc.
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Numerous instances of scaffolding materials, i.e., nails and wood chips,
laying on floors. This material could migrate to drain systems and cause
pump or valva damage. S:af folding was also noted attached to permanent
equipment such as piping and conduit.
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Valve 1001-36A niotor operator conduit had melted plastic cover.
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