ML20134K595
| ML20134K595 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/07/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20134K589 | List: |
| References | |
| 50-293-96-10, NUDOCS 9702140102 | |
| Download: ML20134K595 (20) | |
See also: IR 05000293/1996010
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Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
License No. DPR-35
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Report No. 96-10
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Docket No. 50-293
Licensee: Boston Edison Company
800 Boylston Street
Boston, Massachusetts 02199
Facility: Pilgrim Nuclear Power Station
inspection Period: November 23,1996 - January 11,1997
Inspectors: R. Laura, Senior Resident inspector
B. Korona, Resident inspector
Approved by: R. Conte, Chief
Reactor Projects Branch No. 5
Division of Reactor Projects
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9702140102 970207
PDR ADOCK 05000293
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EXECUTIVE SUMMARY
Pilgrim Nuclear Power Station
NRC Inspection Report 50-293/96-10
i~his n.soection included aspects of licensee operations, engineering, maintenance, and
plant support. The report covers resident inspection for the period of November 23,1996 '
through January 11,1997.
Operations Extensive preparations, including simulator and classroom training, and l
management oversight led to a well controlled evolution to implement FFWTR. Reactor
operators used excellent self verification and procedural adherence techniques. Some l
procedure changes were required for clarification shortly before the evolution indicating l
that the procedure preparer, reviewer, ORC members and training personnel missed earlier
opportunities to identify and correct these issues. (Section O.4.1)
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After significant changes were made to the problem report process, the training effectively ;
explained the changes including a lower problem reporting threshold, a two level priority
scheme, performing common cause analysis every 6 months and replacing the problem
assessment committee with a corrective action review board. Meaningful exchanges of
information occurred durir'q questions from the training participants. (Section O.6.1)
Maintenance: Maintenance and l&C workers completed the HCU functional testing and
Agastat calibration activities in a competent manner. Indications of overheating in
normally-energized, Agastat relays were evidenced by the bobbin material and washers
becoming brittle. A previous BECo evaluation, largely based on a Wyle lab test report,
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i extended the service life from 10 to 22 years. These Agastat relay service life issues and
generic preventive maintenance implications remain as IFl 96-10-01. Insulation resistance
testing of the "B" core spray pump motor went very smoothly with no problems which
was an improvement over past performance. (Section M.1.1)
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Several lower level adverse equipment / material condition problems identified by the ;
inspector were either went unnoticed or were incorrectly accepted by plant workers and J
members of the BECo management staff. For example, the CRD pump motor air inlet and j
outlet screens were partially clogged with dirt. Also, a steady 1/2 gpm packing leak on the ,
condensate transfer jockey pump went undetected which contributed to radwaste in- I
leakage in the long term. The management tour implementation process yielded mixed
results that was less than fully effective in ensuring the identification and correction of
lower level equipment / material condition issues. (Section M.2.1)
A more rigorous approach of entering TS LCOs during surveillance tests, when required,
better ensures compliance with TS requirements and also allows better consideration of
risk management. Accordingly, Unresolved item 50-293/95-26-01 is closed.
(Section M.3.1)
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Effective reactor fuel receipt inspection training was provided to BECo personnel by a
General Electric representative. The training was thorough and provided not only verbal
direction and a videotaped presentation, but also " hands on" training on the refueling floor.
- Maintenance, reactor engineering, operations, and radiological protection personnel
communicated well to perform the fuelinspections. Discrepancies were appropriately
j identified and dispositioned, which confirmed training effectiveness. (Section M.5.1)
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LER 94-04 and its supplement were closed. The LER provided sufficient information
pursuant to 10 CFR50.73 and NUREG 1022 involving a PCIS actuation during a RCIC
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surveillance test. (Section M.8.1)
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l Enoineerina: Engineering personnel completed an adequate safety evaluation generally
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l bounding the effects of FFWTR of up to 75 degrees. The safety evaluation did not fully 1
discuss two pertinent areas possibly affected by the change including the RIPDs and
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ATWS analyses. In one instance, engineers informally relied on verbal information from the
j vendor which was a poor practice. Subsequently, two vendor letters substantiated the 75
degree FFWTR operation assuring safe plant operations. The Operations Review
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Committee had previously approved the FFWTR safety evaluation and did not identify ,
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these weaknesses. (UNR 96-10-02) Two potential UFSAR updato issues were noted j
regarding core operation in the MELLA region. (UNR 96-10-03) l
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- Plant Suocort
- No unusual or inconsistent increases occurred in the gaseous releases from '
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PNPS to the environment during December 6- 9,1996. The main stack and reactor
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building ventilation 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release totals were less than 1.0% of the TS limit and no
i unusual spikes occurred in the hourly average readings. Positive chemistry personnel I
performance was noted. (Section R.1.1) I
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TABLE OF CONTENTS
1. O P E R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 General Comments ................................. 1
04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1
04.1 Final Feedwater Temperature Reduction (FFWTR) . . . . . . . . . . . . 1
06 Operations Organization and Administration . . . . . . . . . . . . . . . . . . . . . 3
06.1 Major Revisions to Problem Report Program . ........... 3
11. M A I N T E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
M1 Conduct of Maintenance .................................. 4
M1.1 General Comments ................................. 4
M2 Maintenance and Material Condition of Facilities and Equipment ...... 6
M2.1 NRC Plant Tour Results and Evaluation ................... 6
M3 Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . . 8
M3.1 (Closed) Unresolved item (50-293/95-26-01): Safety
Equipment Operability During Surveillance Testing
M5 Maintenance Staff Training and Qualification . . . . . . . . . . . . . . . . . . . . 8
M 5.1 New Reactor Fuel inspection Training and Performance ....... 8
M8 Miscellaneous Maintenance issues .......................... 10
M8.1 (Closed) Licensee Event Report (LER) 94-04: Automatic
Closing of the Reactor Core Isolation Cooling System Turbine
Steam Supply Isolation Valves During Surveillance Testing . . . . 10
lli. ENGINEERING . .............................................. 11
E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . 11
E2.1 Safety Evaluation 3018: FFWTR ...................... 11
I V. PLANT S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 13
R1.1 Gaseous Activity Release Review . . . . . . . . . . . . . . . . . . . . . . 13
V. MANAGEMENT MEETINGS ....................................... 13
X1 Exit M eeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
X3 Management Site Visit Summary ........................... 13
X4 Review of UFSAR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
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REPORT DETAILS
Summarv of Plant Status
Pilgrim Nuclear Power Station (PNPS) began the period operating at approximately
100 percent rated power and remained at or near full power until January 11 when reactor
power was lowered to approximately 65% to implement final feedwater temperature
reduction (FFWTR). Operators commenced returning the unit to full power at the end of
the period.
1. OPERATIONS
01 Conduct of Operations'
01.1 General Comments (71707)
Using Inspection Procedure /1707, the inspector conducted frequent reviews of ongoing ;
plant operations, in general, the conduct of operations was professional and safety
conscious. During tours of the control room, the inspectors discussed any observed
alarms wnb the operators and verified that they were aware of any lit alarms and the
reasons for them. Any anomalies noted during tours were discussed with the nuclear
watch engineer (NWE). For example, the 3-D monicore computer heat balance screen
showed a feedwater subcooling parameter in an " abnormal" state as indicated by the red l
color of the display box. Operators could not readily explain the significance of this
observation. Later, reactor engineers explained that although the heat balance calculation l
inputs into the thermal limit calculations, a bad data point would have been indicated by a '
magenta color box. The red box resulted from higher recirculation flow rates used near the
end-of-cycle operation. The operations department manager included this information in
the operations night notes. Other specific events and noteworthy observations are detailed
in the following sections.
04 Operator Knowledge and Performance
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04.1 Final Feedwater Temoerature Reduction (FFWTR)
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a. Insoection Scope (71707,93702)
A review was performed of the preparations, training and implementation of FFWTR
conducted near the end of this operational cycle to increase cycle efficiency. A reduction
of final feedwater temperature, obtained by securing extraction steam to the first and
second point feedwater heaters, and opening the high pressure heater bypass valve
(MO-3489), adds positive core reactivity to temporarily compensate for reactor fuel
consumption. The inspector observed operator simulator training and reviewed a related
written operator training lesson plan. NRC Region I engineers assisted the inspector in the
technical review of the PNPS licensing, design bases information and the BECo 10 CFR
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Topical headings such as 01, M8. etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual
reports are not expected to address all outhne topics.
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50.59 safety evaluation. During deep back shift inspection on January 11,1997, the
inspector observed operators implement the FFWTR method. The engineering aspects of
the FFWTR are discussed further in Section Ill.E2.1 of this report.
b. Observations and Findinas
The inspector reviewed the written operator training material prepared by operator training
personnel for the operating crews. The training guide included a discussion of the
pertinent portions of the PNPS licensing and design bases. Mentioned in the material was
that BECo implemented FFWTR at PNPS in 1977,1979 and 1983. However, the
extraction steam was only removed from the first point heaters resulting in a feedwater ,
temperature reduction of 30 degrees Fahrenheit. A larger feedwater temperature reduction i
of 60 degrees, not to exceed 75 degrees, was planned this cycle by removing the heating )
from both the first and second point heaters. The training covered the lessons learned
from an event that occurred at another nuclear power plant in November 1995 where the
operators failed to follow the procedure for implementing FFWTR. System engineers
provided input how to best isolate the extraction steam to the first and second point
feedwater heaters and how to open the high pressure heaters' bypass valve. Instructions
were provided to remove the seal-in feature of related feedwater heater valves MO-3109,
3209 and 3489 ia convert them into jog valves. This allowed for slow and controlled
isolation of the first and second point heaters and opening of the bypass valve to control l
the addition of positive core reactivity in smallincrements.
In April 1996 operations support personnel modified (revision 8) Procedure 2.2.152,
Feedwater Heater Extraction Steam and Heater Drains, and added Section 7.5 to provide
instructions to accomplish the FFWTR. After simulator validation in October 1996,
revision 9 was made to procedure 2.2.152. During this inspection period, the operating
crew scheduled to conduct the FFWTR performed the evolution on the simulator. The
inspector witnessed this final simulator dry run noting that the crew members conducted a
thorough review of the procedural steps. A reactor engineer who had done this evolution
in the past offered his insights to the crew. Two more procedural steps needed
clarification. One concern identified by the operators was whether or not the feedwater
temperature referenced was an average temperature of both trains or the lower
temperature of either train. This was especially important when using a feedwater l
temperature limit graph in Attachment 4. A second concern identified by operators j
involved a note that stated not to take controls out of automatic when isolating the steam
to the second point heaters. Operators felt that the note was too restrictive and should I
allow more flexibility when deciding to take manual control of the heater water level
controls. After the training session, a procedure change was made to address the two
aforementioned procedural issues. The inspector determined that the operating crew
performed a thorough verification and validation of procedure 2.2.152.
The operating crew discussed the intent of an exception statement in the front of
procedure 2.4.150, Loss of Feedwater Heating, which stated that the procedure was not
applicable during the process of FFWTR. The inspector later expressed concern to the
operations support team leader that the exception statement was confusing and did not
transition well with procedure 2.2.152. A change was made to the exception statement
which clarified when the procedure would be entered and used when implementing the
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FFWTR method of 2.2.152. The inspector recognized the value of the questioning
attitudes exhibited by the operators during the simulator training in the area of procedural
adequacy; however, the operations support procedure preparer, reviewer, onsite review i
committee (ORC) members and operator training personnel missed opportunities to identify
and correct these procedure quality issues at an earlier stage in the process. The inspector
expressed concern to the plant manager that these procedure quality issues were
consistent with the findings documented in NRC inspection report 96-80 and 96-08, and
questioned whether a feedback loop exists to discuss the lessons learned from a procedure
change quality point of view to improve future performance. I
At the pre-evolutionary briefing (PEB) on January 11,1997, the inspector observed
substantial plant and executive level management presence in the control room. The
operations department manager actively participated in the PEB with the operating crew. l
The nuclear watch engineer (NWE) and crew members pre-selected abort criteria for
possible degradation in condenser vacuum and turbine vibration parameters. A reactivity
manager was assigned to monitor the addition of reactivity as the feedwater temperature ,
was lowered. The FFWTR method was implemented after reactor power was reduced to
66% Extraction steam was isolated from the first and second point feedwater heaters in
a very slow and methodical manner. The reactor operator closing the steam extraction
valves continuously used the prescribed self verification process (i.e., STAR). The high
pressure heater bypass valve was opened. A reduction of approximately 60 degrees in
feedwater temperature was achieved with the FFWTR. After the heaters were isolated,
the reactor was returned to 100% power with no problems. The inspector determined that
operators conducted the FFWTR activities in a well controlled manner with excellent
procedural adherence and self-checking techniques.
c. Conclusions
Extensive preparations, including simulator and classroom training, and management
oversight led to a well-controlled evolution to implement FFWTR. Reactor operators used
excellent self verification and procedural adherence techniques. Some procedure changes
were required for clarification shortly before the evolution indicating that the procedure
preparer, reviewer, ORC members and training personnel missed prior opportunities to
identify and correct these issues.
06 Operations Organization and Administration
06.1 Maior Revisions to Problem Reoort Proaram
a. Inspection Scoce (71707)
The inspector reviewed training provided for a new problem report (PR) program contained
in procedure no.1.3.121.
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b. Observations and Findinos
During the training session, the new problem report process was discussed in detail using
illustrative examples by the plant manager and operations support personnel who facilitated
the training. The previous multilevel PR priority scheme was replaced by a bi-level
significant condition adverse to quality (SCAQ) or condition adverse to quality (CAQ). A
lower problem reporting threshold was established. Shortly after the training, the amount
of problem reports initiated increased approximately three-fold with approximately 20-25
generated per weekday. Detailed root cause analyses will generally be performed for
SCAQs. An apparent cause analysis will be done for CAQs that were caused by
organizational or programmatic failures and/or human error / inappropriate actions. Every
6 months a common cause analysis will be performed to identify any adverse trends.
Finally, the problem assessment committee (PAC) was eliminated and a new corrective
action review board (CARB) was created. The CARB consists of a few select senior
managers designed to review the adequacy of proposed corrective actions and their
implementation for SCAO conditions.
c. Conclusions
After significant changes were made to the problem report process, the provicM training
effectively explained the changes which included a lower problem reporting threshold, a
two level priority scheme, performance of common cause analysis every 6 months and
replacement of the problem assessment committee with a corrective action review board.
Meaningful exchanges of information occurred following questions from the training
participants.
11. MAINTENANCE
M1 Conduct of Maintenance
M 1.1 General Comments
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a. Inspection Scoce (61726. 62707) !
Using inspection procedures 61726 and 62707, the inspector observed portions of
selected maintenance and surveillance activities to verify proper calibration of test
instrumentation, use of approved procedures, performance of the work by qualified
personnel, conformance to limiting conditions for operation, and correct system restoration
following maintenance and/or testing. The following activities wera observed:
- Control rod drive hydraulic control unit (HCU) functional checks for level and pressure
switches
- "B" core spray pump motor winding insulation resistance test
- Calibration of newly installed RPS Agastat time delay relays
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b. Observations and Findinas
Operators prornptly secured the SBO EDG before the end of the two hour operational run
due to the potential fire hazard caused by a leaking fuel oil pump injector tube. Operators
closely followed the procedure when securing the engine. A small coolant leak due to a
broken bolt in the coolant return header support identified by the inspector is further
discussed in Section M2.1 of this report.
Instrumentation and controls (l&C) technicians competently calibrated new RPS Agastat i
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relays and performed HCU level and pressure functional checks. Workers initiated prs and
promptly notified supervision of emergent problems. Activities were properly stopped at
calibration transition points to facilitate shift turnovers. The normally-energized RPS
Agastat relay vendor recommended service life of 10 years was extended by a WYLE lab
test report no. 48687-REL-1.0 to 22 years which was subsequently accepted by BECo
engineering in PR 94.0236. Inspections performed during this period revealed signs of
overheating in the plastic relay bobbin material and end washers / spacers. The inspector
observed the overheated matenal became brittle and crumbled easily. The extension of the
10 year qualification life to 22 years and potential generic implications for the preventive
maintenance program constitutes an inspector follow-item (IFl 96-10-01).
Insulation resistance testing of the 4160 volt "B" core spray pump motor went smoothly
with no problems which was an improvement from previous observations in an earlier
inspection period. The operator wore a complete suit of protective clothing during breaker
racking operations. Section 11 M1.1 of NRC inspection report 96-06 documented several
difficulties encountered previously while testing the "B" RHR pump motor windings. At
that time, the wrong megger box was brought to the work site and operators had to
manually rack the breaker due to a broken test box and degraded connections.
c. Conclusions
Maintenance and l&C workers completed the HCU functional testing and Agastat
calibration activities in a competent manner. Indications of overheating in normally-
energized, Agastat relays were evidenced by the bobbin material and washers becoming
brittle. A previous BECo evaluation, primarily based on a Wyle test report, extended the
service life of these relays from 10 to 22 years. These Agastat relay service life issues
and generic preventive maintenance implications constitute IFl 96-10-01. Insulation
resistance testing of the "B" core spray pump motor went very smoothly with no problems
which was an improvement over past performance.
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M2 Maintenance and Material Condition of Facilities and Equipment ,
M2.1 NRC Plant Tour Results and Evaluation
a. Inspection Scope (71707.62707)
Periodic field inspections were performed in selected plant areas, including the reactor
building quadrant rooms, to assess the overall plant material condition. This review also
included the results of plant tours made by the NRC Region i Administrator, Mr. Hubert
Miller, on December 9,1996 and NRC Chairman, Shirley Jackson, on December 19,1996.
b. Observations and Findinos
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Progress was made during this period in the preservation coating of the control rod drive ;
(CRD) quadrant room and in the outer auxiliary bay. However, the inspector identified i
equipment deficiencies in both areas that were not self-identified by plant workers or !
members of the BECo management staff. Three adverse conditions were noted in the CRD i
quadrant room. The CRD pump motor air inlet and outlet screens were partially clogged j
with dirt. Also, a yellow catch-containment installed under valve 301-40A was completely :
full of water due to valve packing leakage. The yellow catch-containment was readily I
visible from nearby locations. Lastly, the lower level CRD floor drain cover was missing 3
which created the possibility of foreign material entering into the floor drain system. The l
inspector noted an NWE in the reactor building and showed the NWE the above adverse ;
conditions for corrective actions. A few days later, the inspector verified that work j
request tags (WRT) were hung to address the CRD motor screens and the missing floor j
drain cover. The catch-containment under valve 301-40A had been drained of water and a l
tygon hose installed to allow the water leakage to drain to a nearby floor drain. Upon !
closer inspection, the inspector observed that water leakage from valve 301-40A was
leaking on the floor; apparently when the catch containment was modified to install the
drain hose, the catch containment was not sufficiently repositioned. The inspector
reported this condition to the NWE and determined that increased attention-to-detail by
radiological control technicians should have avoided this condition.
In the outer auxiliary bay area, the inspector observed packing leakage (approx.1/2 gpm)
from the condensate transfer jockey pump (P-111) was properly collected and directed to
be processed as radwaste in-leakage. The morning station report tracked radwaste in-
leakage as a performance indicator which averages approximately 30 gpm. The inspector
noted that no WRT was hanging indicating that P-111 shaft packing leakage was not
entered into the work control system for corrective action. Control room operators initially
thought that the adverse condition was entered into the system. After a detailed
equipment history review, the day watch engineer informed the inspector that no WRT
existed for P-111 shaft leakage, but a PR and a WRT were initiated to obtain corrective
actions based on the inspector's concern. Further review determined that a maintenance
request (MR) was written on March 29,1994 for the same problem but was later cancelled
to engineering service request (ESR) 04-33. ESR 94-33 was written to replace the pump
shaft packing with mechanical seals. The ESR was closed on November 29,1994 with
instructions to the maintenance staff to replace the packing with better material. A new
MR was never written to perform the work. The inspector noted that the P-111 shaft
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leakage of approximately 1/2 gpm correlates to 21,600 gallons / month which contributed
significantly to the radwaste in-leakage over a long period of time.
Some of the preservation coatings in the intake structure, performed as part of the plant
material condition upgrade program, showed signs of degradation. Significant packing l
leakage (several gpm) was noted coming from the "A" salt service water (SSW) pump
shaft which cascaded down to lower levels in the intake structure and ultimately into the
water bays. Later in the inspection period, maintenance personnel adjusted the "A" SSW i
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pump shaft packing to reduce the leakage. As mentioned in Section ll M1.1 of this report,
the (SBO) emergency diesel generator (EDG) was secured due to fuel oil leakage from high
pressure tubing located between a fuel injector pump and injector. This leak was identified
and closely monitored by the operators. Also during the SBO EDG run, the inspector
identified a small coolant leak resulting from a broken bolt used to secure the cooling water
return header support. The operators initiated a WRT to obtain corrective actions.
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in the aggregate, these adverse conditions, combined with previously identified issues in ,
1996 NRC inspection reports, indicate that lower level equipment problems exist that were ;
either accepted or unidentified by plant workers and management personnel. The inspector
reviewed the 1996 management observation tour program results contained in electronic
messages. Of the sample reviewed, the inspector noted that the quality of the department
level manager results varied greatly, with very few performed in the last quarter of 1996.
Evidence existed of executive level management participation as indicated by numerous
tour reports. The management observation tour program expectations stem from various
memoranda and electronic messages from the senior vice president nuclear dating back to
September 6,1994. Department level managers were expected to spend 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in the
plant each week and provide a write-up with an overall assessment.
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Late in this inspection period, the licensee issued procedure IOTWI.002, Performance
Evaluation Program, dated January 7,1997. The new procedure provided guidance for
management tours including review and trending of the tour results and the issuance of a
quarterly trend report. The inspector noted that IOTWI.002 did not contain the
expectations for the department level managers. The independent oversight team leader
acknowledged the concern and informed the inspector that the management expectations
would be issued soon.
c. Conclusions
Several adverse equipment / material condition problems identified by the inspector were
either unidentified or incorrectly accepted by plant workers and members of the BECo
management staff. For example, the CRD pump motor air inlet and cutlet screens were
partially clogged with dirt. Also, a steady 1/2 gpm packing leak on the condensate transfer
jockey pump went undetected and contributed to radwaste in-leakage over the long term
due to a work tracking oversight. The management tour implementation process yielded
mixed results and was less than fully effective in identifying and correcting lower-level
adverse equipment / material condition issues. A revised process was initiated.
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M3 Maintenance Procedures and Documentation j
M3.1 (Closed) Unresolved item (50-293/95-26-01): Safetv Eauioment Operability Durina
Surveillance Testina l
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a. Inspection Scooe (61726)
A review was performed to assess BECo's evaluation and actions taken to address a
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surveillance test program weakness that allowed rendering safety equipment inoperable
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during surveillance testing and not entering the applicable limiting condition for operation
(LCO).
b. Observations and Findinas
Section 3.2.1 of NRC inspection report 50-293/95-26 documented that the evaluation of
Request For Information (RFI)90-175 appeared outdated, with improper interpretations of
NRC regulations, based on the newer information contained in NRC Generic Letter 91-18.
Subsequently, operations management reviewed NRC Generic Letter 91-18, Section 6.4,
and related surveillance test information from other nuclear power plants. On November
15,1996, the operations department manager issued operations section standing order 96-
07, Revision O, which implemented a new and more rigorous approach for surveillance
tests affecting LCOs. Operations support personnel prepared a detailed listing of LCO-
related surveillance procedures that coded which ones require entry into a TS LCO. Based
on standing order 96-07 and related program procedure changes, operators routinely enter
LCOs during surveillance tests which render safety related equipment inoperable. The
standing order was intended to remain in effect until revisions are made to the individual
surveillance procedures. The inspector verified operators are using this standing order.
c. Conclusions
A more rigorous approach of entering TS LCOs during surveillance tests, when required,
better ensures compliance with TS requirements and also allows better consideration of
risk management. Accordingly, Unresolved item 50-293/95-26-01 is closed.
MS Maintenance Staff Training and Qualification
M 5.1 New Reactor Fuel Insoection Trainina and Performance (62707)
a. Insoection Scope
The inspector attended new reactor fuelinspection training provided to maintenance
technicians and supervisors, reviewed applicable station procedures, and observed the fuel
inspection activity. The inspector also observed interdepartmental communication during
the fuel inspections to determine whether they were effective to complete the activity in a
quality manner.
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b. Observations and Findinas
Fuel inspection training was performed by General Electric (GE) in early November. The
training was attended by the maintenance technicians who performed the inspections,
maintenance supervisors and radiological protection technicians. The training included a
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videotaped presentation of fuel fabrication and inspection as well as " hands on" training on
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the refueling floor. After the taped presentation, the GE representative further elaborated
on the inspection techniques.
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The " hands-on" training was also conducted by the GE representative. The inspector
noted that technicians who had not performed the inspections before were grouped
together for this training so that appropriate time and discussion was facilitated. The
- inspector verified that the training encompassed the inspections required by PNPS
procedure 4.2, inspection and Channeling of Nuclear Fuel. This training allowed not only
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training on the actual fuel inspection, but also an opportunity for the technicians to unload
the fuel from the storage containers and move it with overhead cranes to the unloading
station, inspection stand, and spent fuel pool.
i The inspector observed portions of the inspections of the 208 fuel assemblies during this
period and noted the inspections were performed in accordance with procedures 4.2; 4.1,
Receiving and Handling of Unirradiated Fuel Assemblies; and 4.0, SNM Inventory and
Transfer Control. Careful transportation of the fuel was observed including slow crane
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movement and positioning into the fuelinspection stand. Technicians were aware of the
fuel movement and guided the bundles and channels into the stand, ensuring no
unintentional contact was made that could potentially damage the fuel rods. Maintenance
technicians were observed thoroughly inspecting the fuel in accordance with procedure. In
,
addition, although it was only required by procedure to inspect the lower tie plate from the
bottom of the bundles before they were placed in the inspection stand, this inspection was
also performed after the bundles were channeled and before they were placed in the spent
fuel pool. Maintenance technicians were observed carefully channeling the inspected fuel
bundles before transportation to the spent fuel pool. Good coordination was observed
between maintenance and the operator on the refuel bridge for the final transportation of
the fuel asse.nblies to the spent fuel pool storage racks.
Maintenance personnel appropriately performed the dimensional inspections and identified
a nonconformance in one of the bundles. Nonconformance report (NCR)96-045 and PR
96.0577 were written to document fuel rod-to-fuel rod spacing gage sticking on one
bundle. GE was contacted for guidance and the bundle was placed back in its shipping
< container until it was inspected again per GE instruction. The subsequent inspection was
successfully performed and the fuel assembly was stored in the spent fuel pool.
RP technicians were observed performing the required surveys and smears during the fuel
inspection activity. The maintenance supervisor on the refuel floor and reactor engineer on
the refuel bridge communicated well and maintained proper special nuclear material
inventory and transfer control in accordance with procedure 4.0. Material balance area
(MBA) forms were appropriately completed to document fuel bundle movement from the
storage crates into the spent fuel pool storage racks.
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c. Conclusions
Effective reactor fuel receipt inspection training was provided to BECo personnel by a
General Electric representative. The training was thorough and provided not only verbal
direction and a videotaped presentation, but also " hands on" training on the refueling floor.
Maintenance, reactor engineering, operations, and radiological protection personnel
communicated well to perform the fuel in.spections. Discrepancies were appropriately
identified and dispositioned, which confirmed training effectiveness.
M8 Miscellaneous Maintenance issues (92902,92700)
M8.1 (Closed) Licensee Event Report (LER) 94-04: Automatic Closina of the Reactor Core
Isolation Coolina System Turbine Steam Sucolv Isolation Valves Durina Surveillance
Testina
The inspector reviewed LER 94-04, submitted to the NRC on September 2,1994, and
Supplement 1 to this LER, submitted on December 29,1994, to verify accuracy,
description of cause, previous similar occurrences, and effectiveness of corrective actions.
The LERs were also reviewed with respect to the requirements of 10 CFR 50.73 and the
guidance provided in NUREG 1022 and its supplements.
LER 94-04 reported an August 3,1994 automatic primary containment isolation control
system (PCIS) Group 5 actuation during the performance of a reactor core isolation cooling
(RCIC) system quarterly surveillance test. Initial investigation of the PCIS actuation signal
determined the direct cause was the failure of the governor control valve to respond to
control system demand due to valve binding. Further investigation revealed the fulcrum
dowel pins in the valve were misaligned, causing the valve binding. The supplement to
LER 94-04 documented the results of BECo's root cause analysis of the improper alignment
of the RCIC turbine steam governor control valve fulcrum alignment pins. I
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A detailed description of the event and BECo's troubleshooting activities was documented l
in NRC inspection report (IR) 50-293/94-18, Section 3.1. Unresolved item (UNR) 94-18-02
was opened pending BECo's evaluation of a discovered RCIC lube oil design issue and
completion of the root cause determination for the governor valve failure. An NRC safety
inspection of BECo's root cause analysis and corrective actions to address the entrapment
of air in the RCIC lubricating oil system was documented in NRC IR 95-02.
UNR 94-18-02 was subsequently closed in NRC IR 95-15, Section 4.3. l
The inspector verified that corrective actions that were planned when the unresolved item
was closed were completed. Specifically, procedure 3.M.4-78, RCIC Turbine 5-Year
Preventive Maintenance Inspection, was revised to include guidance on the alignment of
the dowel pins during governor control valve reassembly. The drawing specified for
revision in the LER and IR 95-15 was retired. The drawings which replaced it were verified
to be revised to include the dowel pins.
LER 94-04 and its supplement accurately documented the event, troubleshooting activities,
corrective actions taken and planned, root cause analysis results, the request for
enforcement discretion while RCIC was inoperable longer than the allowed outage time in
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BECo technical specifications, and similarity to previous events. The supplement was
l submitted in a timely manner after the root cause analysis was completed. The LERs
properly addressed reporting criteria and corrective actions were completed. Therefore,
Licensee Event Report 94-0A and its supplement are closed.
Ill. ENGINEERING
E2 Engineering Support of Facilities and Equipment
E2.1 Safety Evaluation 3018: FFWTR
a. Insoection Scooe (37551)
In Section 104.1 of this report, the operational aspects of implementing FFWTR were
discussed. The inspector reviewed the PNPS updated final safety analysis report (UFSAR)
to identify any pertinent limits or related information that could be adversely affected by
FFWTR. UFSAR Section 14.4, Abnormal Operational Transients, notes that FFWTR is a
viable method to extend full power operation of the core by lowering feed-water inlet
temperature and that the effects of the temperature decrease must be evaluated to
determine any required adjustment to the operating limit minimum critical power ratio
(OLMCPR) for the cycle. The PNPS core operating limits report (COLR), Revision 11D,
contains MCPR operating limits with FFWTR in Table 3.3-3 for temperature reductions up
to 75 degrees Fahrenheit. The basis for the COLR FFWTR information was a cycle-specific
General Electric (GE)' Analysis, WHB: 96-035 dated October 24,1996, which also credited
previous GE analyses for several cycle independent evaluations. Using the GE analyses,
BECo completed 10 CFR 50.59 safety evaluation 3018, dated November 26,1996, which
concluded no unreviewed safety question (USO) was involved with FFWTR of up to 75
degrees.
b. Observations and Findinas
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UFSAR Section 3.3.6.10, impact of increased Core Flow and FFWTR on Reactor Internal
Components, states that reduced feed-water temperatures (analysis was done with a 1
reduction of approximately 43 degrees Fahrenheit as compared to 75 degrees in SE 3018) l
increases the overall pressure differential across the reactor components in the high steam
environments, such as the top guide, upper shroud, shroud head and steam dryer. The
loads for these components are limiting at the reduced feed-water temperature condition.
The inspector noted that Section "D", Affected FSAR Section, of safety evaluation 3018
listed UFSAR Section 14.4, but not UFSAR Section 3.3.6.10. The cycle-dependent GE
evaluation stated that evaluation of the reactor internal pressure differences (RIPDs) was
cycle independent; the assumptions and conclusions of which remained valid for Cycle 11.
The inspector noted that SE 3018 appeared inconsistent with UFSAR Section 3.3.6.10
because SE 3018 allowed FFWTR up to 75 degrees whereas UFSAR Section 3.3.6.10
- assumes approximately 43 degree reduction.
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contained in UFSAR Section 3.3.6.10 but chose not to document the basis for accepting
the 75 degree FFWTR on LOCA RIPDs in writing because General Electric personnel stated
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over the telephone that the increase on the RIPDs was very small relative to the margin
that existed in the design and that the change was Nended for only this operating cycle.
The inspector considered reliance on verbal information from a vendor in lieu of providing a
written basis in SE 3018 for LOCA RIPDs was informal and a poor practice. Further, the
statement in SE 3018 that the changes incurred by FFWTR did not affect any cycle- l
independent assumptions did not appear to be correct. Engineering personnel held further
discussions with GE who provided a letter dated January 16,1997 substantiating the prior
verbal information provided to BECo.
Also, the inspector noted that UFSAR table Q7-1 in the SE 2842 package, initial
Conditions For ATWS Analyses, listed feedwater temperature as 367 degrees (normal
value). The iaspector questioned engineering personnel how FFWTR of up to 75 degrees i
affected the ATWS analyses since SE 3018 did not discuss this aspect. Engineering !
personnel recontacted General Electric who issued another letter, dated January 21,1997,
which indicated that the long-term phenomena of ATWS was dependent on initial core
power and not iriitial feedwater temperature. These two engineering issues (i.e., RIPDs
and ATWS) associated with SE 3018 constitute Unresolved item 96-10-02 pending further
NRC review.
Further, the inspector questioned licensing personnel about two UFSAR update issues. l
UFSAR section 14.4.2, Operating Flexibility Options referenced the core operating
flexibility option no.1 which involved the extended load line limit (ELLL) rather than the
current maximum ELLL (MELLL). The regulatory affairs department manager informed the
inspector that the UFSAR update for MELLL should address MELLL in UFSAR Section
14.4.2. The inspector obtained a copy of SE 2842, approved by ORC on December 1, l
1994, which did not update the aforementioned section. Also, the inspector questioned
why the previous UFSAR update submitted in 1996 pursuant to 10 CFR 50.71(e) did not
include SE 2842. Licensing personnel informed the inspector that UFSAR changes were
not updated until the modification was made and then all related drawings changes were
made as part of modification close-out. On January 15,1997, Beco completed a
regulatory relations group self assessment (96-4) that identified several opportunities for
improvement with updating the UFSAR. Fcr example, the current process of waiting until
modification close-out before updating the UFSAR does not provide timely support to the ;
organization. The inspector expressed concern that the existing UFSAR update process
has the potential to result in violations of 10 CFR 50.71(e). These UFSAR update issues
constitute unresolved item 96-10-03 pending further NRC review.
c. Conclusions
Engineering personnel completed an adequate safety evaluation generally bounding the
effects of FFWTR of up to 75 degrees. The safety evaluation did not fully discuss two
pertinent areas possibly affected by the change including the RIPDs and ATWS analyses,
in one instance, engineers informally relied on verbal information from the vendor which
was a poor practice. Subsequently, two vendor letters substantiated the 75 degree
FFWTR operation assuring safe plant operations. The Operations Review Committee had
previously approved the FFWTR safety evaluation and did not identify these weaknesses.
(UNR 96-10-02) Two potential UFSAR update issues were noted regarding core operation
in the MELLA region. (UNR 96-10-03)
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IV. PLANT SUPPORT
R1 Radiological Protection and Chemistry (RP&C) Controls
R1.1 Gaseous Activity Release Review
a. Insoection Scope (71750)
A review was performed of main stack, offgas and reactor building ventilation gaseous
release data from December 6 - 9,1996 to verify compliance with technical specification
(TS) limits. During this time period, the reactor operated at or near full power,
b. Observations and Findinas
The inspector obtained an EPIC plant computer printout for the hourly averages of gaseous
activity for the reactor building ventilation release rate (microcuries/sec.), offgas release
rate (MR/HR) and main stack release rate (microcuries/sec.). Chemistry personnel routinely
add the hourly average readings each day to develop a total 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release number which
was compared to TS limits. Inspector review of the hourly average readings detected no
unusual or inconsistent increases in the gaseous release data for 12/6-9/96. In all
instances, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> total release data for main stack and reactor building ventilation
was much less than 1.0% of the TS release limits.
c. Conclusions
No unusual or inconsistent increases occurred in the gaseous releases from PNPS to the
environment during December 6 - 9,1996. The main stack and reactor building ventilation
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release totals were less than 1.0% of the TS limit and no unusual spikes occurred
in the hourly average readings. Positive chemistry performance was noted.
V. MANAGEMENT MEETINGS
X1 Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management at the
conclusion of the inspection on February 6,1997. The licensee acknowledged the findings
presented.
X3 Management Site Visit Summary
On December 9 and 10, Mr. Hubert Miller, NRC Region i Regional Administrator visited the
site to meet with the resident inspectors, interview several members of the licensee staff
and tour the plant. Mr. Richard Conte, Region i DRP Branch 5 Chief, visited the site on
December 11 to provide routine oversight activities of the resident inspectors and to
interface with BECo managers. On December 19 NRC Chairman Shirley Jackson visited
the site for a plant tour and meeting with senior BECo managers. Mr. Hubert Miller
accompanied the NRC Chairman during her visit on December 19.
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- X4 Review of UFSAR Commitments
a
I A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR
1
description highlighted the need for additional verification that licensees were complying
with Updated Final Safety Analysis Report (UFSAR) commitments. For an indeterminate
time period, all reactor inspections will provide additional attention to UFSAR commitments
and their incorporation into plant practices and procedures. While performing inspections
discussed in this report, inspectors reviewed the applicable portions of the UFSAR. Several
UFSAR related issues were identified and documented in Section til E2.1 of this report.
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4
l INSPECTION PROCEDURES USED
>
IP 37551: Onsite Engineering ;
IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing
'
Problems
.i IP 61726: Surveillance Observation
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IP 62707: Maintenance Observation
i IP 71707: Plant Operations
IP 71750: Plant Support Activities
i IP 82301: Evaluation of Exercises for Power Reactors
, IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor :
! Facilities
i IP 92901: Followup - Operations
1
IP 92902: Followup - Maintenance .
IP 92903: Followup - Engineering l
4
IP 92904: Followup - Plant Support
l lP 93702: Prompt Onsite Response to Events at Operating Power Reactors
4
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ITEMS OPENED, CLOSED, AND UPDATED
]
1
Ooened L
,
IFl 96-10-01 Agastat Relay Service Life Extension and Generic Implications
1 UNR 96-10-02 SE 3018 (FFWTR) Weaknesses
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! UNR 96-10-03 UFSAR Update Concerns For SE 2842 (MELLA)
f
Closed ;
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UNR 95-26-01 Entering TS LCOs During Surveillance Tests
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LER 94-04 Automatic Closing of the Reactor Core Isolation Cooling System
i Turbine Steam Supply isolation Valves Due to High Steam Flow Signal
I During Surveillance Testing
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LIST OF ACRONYMS USED
ALARA As Low As is Reasonably Achievable
APRMs Average Power Range Monitors
BECo Boston Edison Company :
CFR Code of Federal Regulations
CRD Control Rod Drive
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CS Core Spray i
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EPIC Emergency and Plant Information Computer
ESF Engineered Safety Feature
gpm gallons per minute l
l&C Instrumentation and Controls
IFl Inspection Follow-Up Item
IR inspection Report j
LER Licensee Event Report
MG Motor Generator
MR Maintenance Request
NCV Non-Cited Violation
NOV Notice of Violation
NRC Nuclear Regulatory Commission
NRR Office of Nuclear Reactor Regulation
NWE Nuclear Watch Engineer
PNPS Pilgrim Nuclear Power Station
PR Problem Report
RP Radiological Protection
SALP Systematic Assessment of Licenseo Performance
SE Safety Evaluation l
SNM Special Nuclear Material l
SRO Senior Reactor Operator ;
TS Technical Specification ;
UFSAR Updated Final Safety Analysis Report '
WWM Work Week Manager
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