ML20129A286
| ML20129A286 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/04/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20129A121 | List: |
| References | |
| 50-302-96-09, 50-302-96-9, NUDOCS 9610220088 | |
| Download: ML20129A286 (26) | |
See also: IR 05000302/1996009
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U.S. NUCLEAR REGULATORY COMMISSION
REGION 2
Docket No:
50-302
License No:
Report No:
50-302/96-09
Licensee:
Florida Power Corporation
Facility:
Crystal River 3 Nuclear Station
Location:
15760 West Power Line Street
- Crystal River. FL 34428-6708
Dates:
August 11, 1996 - September 7. 1996
Inspectors:
R. Butcher Senior Resident Inspector
T. Cooper, Resident Inspector
S. Cahill, Resider,t Inspector
W. Bearden. Reactor Inspector, paragraph 08.1
B. Crowley, Reactor Inspector, paragraphs E8.1, E8.2,
R8.1, X1.1
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R. Gibbs, Reactor Inspector, paragraph 08.1
E. Girard, Reactor Inspector, paragraphs E8.3, E8.4
Approved by:
K. Landis, Chief, Projects Branch 3
Division of Reactor Projects
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9610220088 961004
ADOCK 05000302
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EXECUTIVE SUMMARY
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Crystal River 3 Nuclear Station
NRC Inspection Report 50-302/96-09
This integrated inspection included aspects of licensee operations
engineering, maintenance, and plant support.
The report covers a four week
period of resident ins)ection; in addition, it includes the results of
announced inspections )y four regional reactor inspectors.
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Ooerations
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Several power reductions occurred as a result of secondary side equipment
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problems. This reflects a continuing trend of poor unit performance due to
secondary side equipment problems as previously noted in Inspection Report
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50-302/96-05. (paragraphs 02.1 - 02.4)
The Employee Concerns Program was reviewed, and found to be effective and
adequate.for the Crystal River site. (paragraph 08.1)
Maintenance
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A Violation (50-302/96-09-01) was identified for failure to follow a
procedure, resulting in the inadvertent initiation of the control room
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emergency ventilation system. (paragraph M1.1)
Enoineerina
. A Violation (50-302/96-09-04) was identified for failure to update plant
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operating curves to reflect a power uprate that occurred in 1981. (paragraph
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E8.1)
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A Violation (50-302/96-09-05) was identified for failure to incorporate a
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design change in the operation of Makeup System valve MUV-64 into operations
procedures. (paragraph E8.2)
A Violation (50-302/96-09-06) was identified for three examples of design
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control errors. The errors included incorrect design inputs for the battery
charger and transformer modifications, and failure to update a drawing and
-the ISI program following a modification 12 years ago. (paragraph E8.3)
A Violation (50-302/96-09-07) was identified for failing to take timely
corrective action to modify the Emergency Feedwater Initiation and Control
system. (paragraph E8.4)
Plant Sucoort
'A Violation (50-302/96-09-02) was identified for failure to follow security
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procedures, resulting in unescorted visitor personnel within the protected
area. (paragraph S1.1)
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A Violation (50-302/96-09-03) was identified for failure to perform a 10 CFR 50.59 safety evaluation for changes to operating limits described in the Final
Safety Analysis Report for controlling dissolved hydrogen concentration.
(paragraph R8.1)
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Report Details
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Summary of Plant Status
The unit began the inspection period with the output breakers closed and
the unit at 100 % power. The following evolutions occurred this
inspection period:
- On August 14. 1996 at 7:28 3.m.. reactor power was reduced to 94 %
due to flashing occurring in t1e hotwell following the closing of HDV-
61. HDV-61 is the drain valve from FWHE-3A to FWHE-2A.
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- On_ August 17, 1996 at 7:37 a.m.
reactor power increase commenced
following the repair of HDV-61.
Reactor power was returned to 100 % at
7:00 p.m.
- On August 20, 1996 at 10:00 p.m.
reactor )ower was reduced to
approximately 60 % due to the loss of the B C)P.
See paragraph 02.1 for
more details.
- On August 22, 1996 at 1:18 a.m., a reactor power increase was
initiated after repairs to the B CDP.
Reactor power reached 100 % at
11:00 a.m. on August 22, 1996.
- On August 26, 1996 at 1:11 a.m. while at 100 %. a aower reduction was
commenced to remove CWP-1A from service to support scleduled
maintenance.
Reactor power of 85 % was reached at 3:35 a.m.
.On August 30. 1996 at 9:36 a.m. while at 85 % reactor power, a
Jower
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reduction was initiated due to fan belt breakage on the generator
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duct cooling fan.
Reactor power was stabilized at 40 % at 2:04 p.m.
- Following repairs to both CWP-1A and the generator bus duct cooling
fan, reactor power was returned to 100 % at 9:55 a.m. on September 1.
1996.
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- At 1:40 p.m. on September 2. 1996, a power reduction from 100 % power
was initiated due to low turbine lube oil pressure. The main generator
output breakers were opened at 9:00 p.m. on September 2. 1996, with
reactor power maintained at approximately 8 % of rated power.
- At 5:00 p.m. on September 3. 1996 the decision was made to place the
plant in Mode 3.
At 6:00 p.m. the plant entered Mode 2 (less than or
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equal to 5 % reactor power).
At 6:20 p.m. the )lant entered Mode 3
(shutdown with reactor coolant system greater tlan or equal to 280
degrees F).
- Due to a problem identified with the turbine lube oil piping, the
decision was made to continue cooldown of the plant.
The plant entered
Mode 4 (280 degrees F > Tavg > 200 degrees F) at 9:05 p.m. on September
4. 1996 and established Decay Heat Operation at 4:30 a.m. on September
,
5. 1996. - Mode 5 (Tavg less than or equal to 200 degrees F) was entered
at 11:38 p.m. on September 5, 1996.
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L. Operations
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Conduct of Operations
01.1 General Comments (71707)
01.2 Plant Shutdown
During the plant shutdown on September 2, 1996, at 10:00 p.m., while at
a reactor power of ap]roximately 22%, secondary cycle samaling pump SSP-
4D tripped, allowing Jack flow from the sample valve to t1e condenser.
This resulted in substaritial air leakage into the condenser and an
increase in hotwell oxygen.
Chemistry personnel became aware that air
was being drawn into the condenser through the sample valve and closed
SSV-25.
PR 96-351 was written to document this problem and recommended
an evaluation be performed to add an in-line check valve that would
prevent similar events.
During the loss of vacuum in the condenser as noted above, operators
manually started a second vacuum pump -started a second gland water
pump, and increased the rate of power decrease.
This rapid power
reduction resulted in a pressurizer out surge and RCS pressure reduction
to approximately 2048 psig.
TS 3.4.1, RCS Pressure. Taperature, and Flow Departure From Nucleate
Boiling (DNB) Limits, Condition A, requires with one or more RCS DNB
parameters not within limits, restore the parameters to within limit
within two hours.
SR 3.4.1.1 requires loop pressure be equal to or
greater than 2061.6 psig with four RCPs operating.
During the evolution
noted above, the RCS pressure was below 2061.6 psig for approximately
two minutes, resulting in no violation of TS.
The licensee issued PR 96-354. RCS Pressure Less Than DNB Limit, to
address this issue on September 3, 1996.
The root cause analysis
determined one of the primary causes was that management expectations
concerning maintenance of primary plant parameters during events such as
this, were not clearly understood.
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The licensee has developed a corrective action plan to address this
cause.
Interpretations of TS requirements during plant transients are
being developed.
An OSB entry is being developed detailing the results
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of the event investigation and management expectations concerning
maintenance of primary parameters during plant transients.
The licensee
is revising the licensed operator training to address these management
expectations. The developed corrective actions for this event are being
tracked by the licensee and no further actions by the NRC are required.
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02
Operational Status of Facilities a .iquipment
02.1 Loss of the B Condensate Pumo (717)
On August 20, 1996 at 10:00 p.m. with the plant at 100 % rated power,
the control room received a Condensate Pump B Uncoupled alarm followed
by a B Hotwell Level High annunciator alarm.
Operators verified the B
CDP was uncoupled and entered AP-510. Rapid Power Reduction, and reduced
reactor power to ap3roximately 60 %.
During the power reduction, the B
CDP was placed on t1e spare controller and demand was increased with no
response.
Investigation showed that the control cabinet for the B CDP
had no power and therefore the electromagnetic coupling was not
functioning.
The condensate pumps are located on the 95 foot elevation
of the Turbine Building on the northeast and northwest corners of the
Each pump assembly is set into a deep pit enclosure
located below the 95* elevation.
The' pump and motor shafts are not mechanically connected.
They are
coupled via an electromagnetic coupling device, which allows the speed
of the aump to be varied while the motor turns at a constant speed of
1150 RPi. As the output of the pump controller is increased, an
increasing voltage is applied to the electromagnetic coils.
This causes
an increase in the magnetic field strength between the motor and pump
shaft, causing the pump shaft to begin to follow the motor rotation.
There are three condensate pump coupling control cabinets located on the
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95 foot elevation of the Turbine Building on the north end of the
condensers.
The two outside cabinets are designated as the normal
controls for its respective condensate pump.
The center cabinet is a
spare coupling controller and may be selected for use on either
condensate pump. A select switch, mounted on the spare controller
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cabinet, is used to select A or B condensate pump to the spare
controller.
Technicians replaced blown fuses in the B aump and spare control
cabinets.
The available evidence for the alown fuses was inconclusive,
with the most probable cause being a ground on the brushes or slip rings
caused by build up of carbon dust from the brushes. The pum) was tagged
out, the brushes replaced and the slip ring area cleaned.
T1e B CDP was
run with no load and the pump was swapped to the spare controller and
back to normal to verify that both controllers were functional.
PR 96-
327. Loss of CDP-1B coupling causes power reduction from 100 % to 62 %.
was written to document this problem.
A reactor power increase commenced at 1:18 a.m. on August 22, 1996.
02.2 Circulatina Water Pumo 1A Reoairs (71707)
On August 26. 1996 at 1:11 a.m. while at 100 % reactor power, a power
reduction was initiated to remove CWP-1A from service for scheduled
maintenance.
Reactor power was stabilized at 85 % at 3:35 a.m.
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Following completion of the 10 R refueling outage, the licensee observed
degradation in the discharge )iping of the circulating water pumps. An
interim maintenance plan has Jeen implemented to periodically remove one
of the CWPs from service'and provide a temporary repair utilizing
Belzona.
N.3 Generator Bus Duct Coolina Fan (71707)
A' drive belt on generator bus duct cooling fan TBF-1 was evaluated and
found to need replacement.
If the generator bus duct cooler is lost,
power must be reduced so that the generator current-is below 10,000
amps. A power reduction was initiated at 9:36 a.m. on August 30, 1996,
and reactor poner was stabilized at 40 % (313 MWe), with generator
current at 8.770 amps at 2:04 p.m. on August 30, 1996.
After replacement of the belts, the generator bus duct cooler fan was
placed back in service at 8:00 a.m. on August 31, 1996. After a run-in
period and tension test for the new belts, a power increase was
initiated at 3:00 p.m. on August 31, 1996.
Since CWP-1A had been
returned to service also (see paragraph 02.2), reactor power was
returned to 100 % at 9:55 a.m. on September 1, 1996.
02.4 Decreased Turbine Lube Oil Pressure (71707)
At 4:00 p.m. on August 30, 1996. Crystal River 3 received an auto-start
of the main turbine ac Jowered back-up bearing oil pump (TBP-2) and the
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high pressure seal oil Jack-up pum) (TBP-8) on low lube oil pressure.
' eis auto-start of the ac powered Jack-up oil pump occurs at 10-12 psig.
At the time, the unit was at approximately 40% power, in the process of
reducing power to replace one failed and one degraded fan belt on the
main generator bus duct cooling fan (see paragraph 02.3).
TBP-8 was
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returned to standby, but TBP-2 restarted when a shutdown was attempted,
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Main lube oil pressure remained at approximately 18 psig with TBP-2
operating.
This is the normal operating lube oil pressure.
The main turbine lube oil system consists of the main oil pump (TBP-1),
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the AC turbine bearing oil pump (TBP-2), the DC turbine bearing oil pump
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(TBP-3), the high pressure seal oil backup pump (TBP-8), the bearing
lift oil pump (TBP-6), the lube oil ejector and the main turbine lube
oil reservoir (LOT-2), with its associated heaters, coolers, and va)or
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extractor.
Under normal full power operating conditions oil from t1e
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main turbine lube oil reservoir is routed to the individual bearings on
the main turbine by the lube oil injector.
The main oil Jump, which is
attached to the shaft of the main turbine, )rovides a hig1 pressure
supply of oil that is used to operate the tirust bearing trip device and
serves as a backup source of oil to the seal oil system.
The main oil
pump and lube oil ejector are dependent on each other. The main oil
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pump depends on the lube oil ejector for its suction supply and the
ejector uses the discharge of the main oil pump as its motive force.
The ac powered turbine bearing oil pu p (TBP-2) is located on and takes
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suction from the main turbine lube 01
reservoir.
TBP-2 is interlocked
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to auto start any time bearing lube oil pressure drops to below 10-12
psig. TBP-3 is interlocked to autostart any time bearing lube oil
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pressure drops to below 8-10 psig as sensed by TB-253-PS.
After discussions with Westinghouse, the licensee issued STI 96-029 with
guidance for operations. This guidance included the following:
- Maintain lube oil temperatures as low in the normal band as possible.
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To facilitate this, both SCPs were placed in service,
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If turbine lube oil pressure degrades below 14 psig while TBP-2 is in
o]eration, the turbine should be removed from service via a controlled
slutdown.
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Increase the frequency of monitoring lube oil pressure.
After a review of the data, engineering recommended that the plant
continue to run TBP-2 continuously and return the plant to full power
following repair of the bus duct cooling fan.
The plans were to
schedule an outage to repair the low lube oil pressure problem. At 9:55
a.m., on September 1, 1996, the plant was returned to full power
operation.
At 6:30 p.m., it was noted that turbine lube oil pressure had continued
to degrade, at a rate of 0.8 psi / day. At 9:00 a.m. on September 2,
1996, a meeting was held to discuss the troubleshooting plan for the low
lube oil pressure concerns. At 1:40 p.m. on Seatember 2, 1996, the
decision was made to decrease power and place t1e turbine on turning
gear, in order to troubleshoot and make repairs.
The main generator
output breakers were opened at 9:00 p.m. on September 2, 1996 and
reactor power was maintained at approximately 8% of rated power.
At 4:55 a.m. on September 3, 1996, when the maintenance technicians
began removing the hand hole cover over LOV-471, oil came out.
The
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licensee removed the adjacent manway cover and saw oil spraying inside
the tank from the area of check valve LOV-471.
Upon further
investigation, it appears that the line downstream of LOV-471 has a long
longitudinal split. The licensee )roceeded to go on decay heat and
stayed in Mode 5.
At the end of tais inspection period, the lube oil
tank was drained in order to make repairs.
08
Hiscellaneous Operations Issues
08.1 Emoloyee Concerns Proaram
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Insoection Stone (40001. 40500)
An inspection was conducted to evaluate the effectiveness of the
licensee's Employee Concerns Program. The inspection included a program
review, employee interviews, and documentation reviews.
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The inspectors reviewed the licensee's program as cutlined in Nuclear
Operations Department procedure N00-36, Employee Concerns Program,
Revision 6 and. interviewed the Employee Concerns Representative who
administers the program. The interview with the Employee Concerns
Representative focused on his qualifications for the position, as well
as a review of the program and data related to the concern
receipt / closure rate, number and age of open concerns, and outage time
frames.
The inspectors interviewed 28 employees from various levels (i.e.
managers and technicians) and disciplines, including:
engineering,
operations, maintenance, chemistry, health physics, security and
training personnel. The selection was random, except the selection was
made by departments in order to obtain a representative sample from the
various work disciplines. This interview sample size was consistent
with other inspections of this nature.
Personnel interviewed were asked
if they would report safety concerns to their su
whether they were knowledgeable of the licensee'pervisor or management. j
s concerns program and
how to use the program, whether the licensee's facilities were
adequately accessible, and whether they felt uncomfortable or knew
someone that had been badgered for reporting safety concerns.
Personnel
who stated they had used or had knowledge of someone who had used the
concern program were asked about timeliness and adequacy of concern
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resolution.
Eleven closed employee concern files from the 1995/1996 time frame were
reviewed to determine the adequacy of the licensee's investigation and
corrective actions.
Files selected included both substantiated and
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unsubstantiated concerns. Specifically the inspectors reviewed the
files to determine if overviews and summaries of activities related to
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the concerns (i.e. , priorit , investigations, and communications) were
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adequate to address the emp oyees' safety concerns.
Additionally the
inspectors evaluated each f le to determine if concerns were clearly
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identified, and if closeout letters to the concerned employee adequately
described the concern, the extent of the licensee's review, whether the
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employee's concern was substantiated or not substantiated, and any
planned corrective actions.
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The inspectors also conducted a review of the anonymous precursor cards
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(e.g. low level corrective action documents written by persons unknown)
which had been written in the last six months at the site.
A listing of
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these cards, which included a description of the problem and the
corrective action taken, was provided by the licensee and reviewed by
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the inspectors.
In addition, data concerning the number of anonymous
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precursor cards written versus the total number of precursor cards
written during 1994 through August of 1996 was provided and reviewed.
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Observations and Findinas
The Employee Concerns Program is small and relatively informal, staffed
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with only one individual.
The number of concerns received in the
program is also small, and the safety significance of those concerns is
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limited. The number of concerns received was as follows: 1993 - 8
concerns from exit interviews and 34 from working employees (42 total):
1994 --74 concerns from exit interviews and 35 from working employees
(109 total): 1995 - 8 concerns from exit interviews and 6 from working.
employees (14 total): 1996 - 40 concerns from exit interviews and 10
from working employees (50 total). The high number of concerns in 1994
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was attributed to some company downsizing and a refueling outage. The
low number of concerns in 1995 was attributed to management development
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and support of the Precursor Card Program. The high number of concerns
in 1996 was attributed to a refueling outage.
Also, timeliness of
concern resolution was adequate, with only 1 concern open from 1994 and
12 open from 1996.
During the inspection,' the number of anonymous precursor cards raised
some concern on the part of the inspectors.
Further investigation of-
this area resulted in resolution of this concern as follows:
Review of
the anonymous precursor cards written in the last six months determined
that there were a total of 110 written.
Review of a listing of these
cards, however, revealed that only five of these cards had any safety
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significance, and the NRC had already investigated the problems
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identified on two of those five cards. Additionally, further comparison
of the number of anonymous cards to the total number of cards written
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determined that the percentage of anonymous cards was very small,
averaging less than four percent (1994 - 45 of 719, 1995 - 83 of 2886,
and 1996 - 117 of 3959).
Proaram Strenaths:
All employees interviewed stated they had confidence in the ability of
their management to safely operate the plant.
Essentially all personnel interviewed expressed confidence in management
and a receptiveness on the part of management regarding identification
of safety concerns.
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All personnel interviewed stated that they would raise safety issues
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through their management, the corrective action program, the Employee
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Concerns Program or the NRC.
All personnel interviewed stated that they would resolve safety issues
through their line management and the licensee's corrective action
program prior to involving the Employee Concerns Program or the NRC.
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Most personnel were familiar with the existence of the licensee's
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concerns program and felt that accessibility to the program was
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acceptable. Only a small percentage of the people interviewed had used
the program.
The site employee concern representative provided adequate physical
3rotection of the files, and all records related to investigations were
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cept in locked storage with restricted access to protect the
individuals' identities.
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In most cases reviewed, the closecut letter to the concerned employee
was completed.within four months from receipt of the concern, The
inspectors' review of the files indicated that the quality and
timeliness of the licensee's reviews of concerns, investigations, and
followup with concerned employees was adequate.
Procram Weaknesses:
The percentage of concerns in the security area was unusually high. The
licensee provided information which indicated recognition of this
problem, including a self assessment of the security area.
Resolution
of many of the issues identified in the assessment was not completed,
however, and continued management attention to this area is warranted.
The inspectors were informed that SBI Security has a separate concerns
program, which is managed from SBI's corporate office.
SBI is the only
contractor which has a significant number of personnel on site.
Although SBI personnel frequently use the licensee's program, SBI's
corporate program is available for use by SBI personnel working at
Crystal River and is not audited by the licensee.
When requested by the
inspectors, the licensee stated that they were not knowledgeable of the
extent of use, if any, of the SBI corporate program by site SBI
personnel.
Concerns are closed out based on actions which will be taken in the
future to resolve a concern without any followup to verify completion of
those corrective actions (i.e., corrective action recommendations are
not tracked through implementation),
The Employee Concerns Program relies on the corrective action program
for resolution of some concerns, and yet the Employee Concerns Program
and the corrective action program are not tied together. Although
technical reviews of the concerns were effectively performed such that
concerns were fully investigated, closure packages did not always
include documentation to demonstrate that the employee's concern had
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been adequately addressed.
However. when requested by the inspectors
the Employee Concerns Representative was able to produce documentation
such as precursor cards to show that selected deficiencies had been
documented.
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Some of the individuals who had used the Employee Concerns Program were
not satisfied with the results.
Some of this can be attributed to a
lack of followup with the concerned individual to assure they fully
understood all corrective actions taken.
Others were not satisfied with
the timeliness of concern resolution.
This perception, whether factual
or not, has a tendency to su) press the effectiveness of an employee
concerns program _and should 3e addressed by the licensee.
Although letters to concerned individuals do address whether or not any
corrective actions will be taken, the employee concerns packages and
letters to the individuals do not reflect whether or not the concern is
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substantiated or not substantiated,
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Nuclear Operations Department procedure NOD-36 requires the concerned
individual be notified if a concern resolution will take longer than
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30 days. There is no documentation included in the concern packages
which reflects this action.
The letter to the concerned individual in one file did not accurately
reflect the actual corrective action, which had resulted from the
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licensee's review of the concern. Corrective actions described in the
letter would not have been adequate to fully resolve the concern.
However, the inspectors noted that actual corrective actions completed
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by the licensee were adequate. The inspectors were informed that the
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concerned individual would be notified that additional corrective
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actions had occurred.
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Conclusions
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The inspectors concluded that the licensee's program was effective in
handling and resolving employee safety concerns.
Employees who were
interviewed knew about the licensee's concerns program and would use it
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if they needed to, but most were generally satisfied with their
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supervisors' rece]tiveness to resolving safety concerns without the
intervention of t1e concerns arogram. Accessibility to the program was
thought to be acceptable to t1e people interviewed. The inspection
determined that the technical issues in the )rogram were being
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adequately resolved. The inspectors noted t1e willingness of licensee
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employees and management to identify and resolve safety issues through
the normal chain of command and site corrective action programs.
The
inspectors concluded that, even though the Employee Concerns Program is
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a small and informal program it is adequate.
IL. Mahltenance
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Conduct of Maintenance
M1.1
Inadvertent Actuation of Control Room Ventilation in the Recirculation
Mode (62707)
On August 13. 1996, during performance of PT-366. Toxic Gas Detection
System Calibration, an iriadvertent actuation of the toxic gas system
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occurred. While performing step 4.7.1. the contract technician was
instructed to ) lace the SAMPLE /ZERO switch to the ZERO Josition.
By
error, the tecinician placed the main power switch to tie OFF position.
This caused the monitor to deenergize and tripped the control complex
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ventilation system to the emergency mode. The technician, realizing
that something was wrong, placed the power switch back to the ON
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position.
In response, the operations SSOD had the technicians stop the SP, and
required the I&C supervisor to directly supervise restoration of the
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system to its normai alignment.
A precursor card was initiated and
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plant management was notified.
The licensee initiated a HPES evaluation
to determine probable root causes.
On August 15. 1996 a MRP meeting was
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held to discuss the issue, determine adequacy of the evaluation, and
begin the development of a corrective action plan.
The HPES evaluation identifiea several causes for the inappro)riate
actions, including: (1) not having the su) plied 3rocedure in land while
4
performing the actions, (2) not using STAR, (3)
laving a non TPM
qualified technician independently performing procedure steps, and (4)
failure to maintain constant communications while performing the task.
A number of contributing factors were also diccussed, including: (1)
management did not provide adequate training to temporary personnel on
the use of procedures
(2) management has not constantly reenforced the
Event-Free operations program to plant personnel, (3) effective
monitoring of the use of procedures is inadequate, (4) management has
not provided adequate training on STAR to temporary personnel , (5)
management has not conveyed adequately to all maintenance personnel the
need to eliminate all inadvertent actuations of equipment in the plant,
and (6) job assignment did not take into account the need for two TPM
qualified technicians to perform the independent tasks of PT-366.
The inspectors attended the MRP meeting where the results of the HPES
were discussed.
The DNPO constantly challenged the assessors to defend
their conclusions and to provide him with corrective actions that would
address the contributing factors, as well as the identified primary
4
'
causes, One comment made by the HPES evaluator was that there is
evidence of a perception at CR3 that adherence to procedures and STAR is
pro)ortional to the consequences of not using them.
The DNP0 agreed
wit 1 the statement, and directed the responsible personnel to determine
the root cause of this perception and develop corrective actions to
address it.
The MRP conducted for this event was very detailed and
stressed identifying all of the contributors to this problem.
Technical Specifications (TS) 5.6.1.1 requires written procedures be
established, implemented, and maintained covering the applicable
3rocedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A,
r bruary 1978.
RG 1.33, Appendix A requires administrative procedures
e
regarding prncedure adherence. AI-400E, Performance and Transmittal of
Procedures, paragraph 1.1, Policy, states that verbatim compliance of
procedures is required, but procedures must not be blindly f6110wed.
The technician's failure to follow the requirements of PT-366 is a
violation, VIO 50-302/96-09-01, Failure to follow a procedure resulting
in the inadvertent initiation of the control room emergency ventilation
'
system.
M3
Maintenance Procedures and Documentation
M3.1 Surveillance Observations (61726)
The inspectors witnessed portions of the performance of PT-325, Turbine
Generator Checks.
During the surveillance, the C MSR combined
intercept /stop valve failed to stroke.
Following maintenance
troubleshooting of the controller, the valve was stroked successfully.
The cause of the valve failing to stroke was not determined. The
-
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11
following day, while performing the procedure for the remaining MSRs,
the A valve stroked, as required, but the B and D valves failed to
stroke.
The licensee again performed troubleshooting, but no
discernible problems were identified.
After reviewing industry experience and discussing the situation with
.
the turbine vendor, the licensee determined that the most 3robable cause
of the problem was excessive moisture in the EHC fluid.
T1e moisture
would interfere with the operation of the control solenoid. As interim
corrective actions. the licensee replaced the Fuller earth filters on
the oil and filled the EHC oil tank, to decrease available volume for
moisture containing air to occu)y. The inspectors verified that the
interim corrective actions had Jeen completed. The licensee is
continuing to work on a permanent solution to address this concern.
M8
Miscellaneous Maintenance Issues
M8.1 Emeraency Feedwater Pioina (92902)
LER 50-302/95-027-00, Leak in underground emergency feedwater piping
suspected to be caused by nearby evacuation using acwer tools, was
submitted as a voluntary LER on January 5. 1996.
Juring the last refuel
outage (10R) the licensee excavated the EFW piping and corrected the
problem. As documented in IR 50-302/96-03, paragraph M1.1, the NRC
reviewed the licensee's actions and concluded that the licensee's
actions were appropriate. This LER is closed.
J1L. Enaineerina
E8
Miscellaneous Engineering Issues
E8.1
(Closed) URI 50-302/96-03-09. Ooeratina Curves Not Promotly Uodated to
Reflect 1981 Power Vocate
a.
Insoection Scoce (92903)
This issue was described in paragraph E8.1 of NRC Inspection Report
50-302/96-03, and was left unresolved pending NRC review to determine if
it was an additional example of an apparent violation of 10 CFR 50.
Appendix B, Criterion III. discussed at a predecisional enforcement
conference on March 27, 1996.
The issue involved failure to update,
until July of 1995, Operating Procedure OP-103A. Startup Curves, to
reflect the 1981 power uprate.
b.
Observations and Findinas
Upon further review, the NRC determined this issue is substantially
different and not an additional example of the apparent violation.
i
i
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12
c.
Conclusions
This issue is identified as VIO 50-302/96-09-04. Failure to Update
Operating Curves to Reflect 1981 Power Uprate.
URI 50-302/96-03-09 is
closed.
E8.2 (Closed) URI 50-302/96-04-09. Failure to Incoroorate Desian Information
- nto Ooerations Procedures
a.
Insnection Scooe (92903)
.
This issue was described in paragraph E2.3 of NRC Inspection Report
50-302/96-04, and was left unresolved pending NRC review to determine if
it was an additional example of an apparent violation of 10 CFR 50,
,
Appendix B. Criterion III, discussed at a predecisional enforcement
1
conference on March 27, 1996. This issue involved failure to revise
Operations Procedures, or to provide training to operators, after
modifying Valve MUV-64 to change the Valve Operator from a disabled air
operated valve (locked in the open position) to manual operation with a
handwheel, in accordance with MAR 95-01-07-01.
The MAR was implemented to address, in part, the concerns of operations
personnel regarding the loss of level indication in the Makeup Tank
(MUT) during Loss of Coolant Accident (LOCA) conditions, and the
)otential for hydrogen gas being entrained in the suction to the Makeup
Jumps (MVPs).
The design input record of the MAR package discussed
adding the manual gear driven chain operator for better access in the
manual mode, since it might be necessary to quickly stroke the valve
closed in certain accident scenarios with the new higher MUT hydrogen
gas pressures (e.g. , certain Appendix R fires, keeping the MUT level
indication on scale post LOCA rapid boration requirements, etc.). The
valve was identified in the valve lineup section of operating procedure
OP-402. Makeup and Purification System. Revision 84.
The procedure only
showed the valve position as being sealed open, and there was no
discussion on operation of the valve during potential accident
conditions.
b.
Observations and Findinas
Upon further review, the NRC determined this issue is substantially
different and not an additional example of the apparent violation,
c.
Conclusions
This issue is identified as VIO 50-302/96-09-05. Failure to Incorporate
Design Information Into Operations Procedures.
URI 50-302/96-04-09 is
closed.
As noted in NRC Inspection Report 50-302/96-04. licensee management
stated that they did not agree with the NRC position on this issue
because they had reviewed operations and other plant department
procedures (in accordance with their design control process) and
.
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13
determined that no procedures needed to be revised.
During the current
inspection, the licensee continued to disagree with the NRC position
since, based on calculations in process in parallel with implementation
of the MAR, they did not plan to operate valve MUV-64 during accident
conditions. After further discussion. in order to add conservatism
during accident conditions, the licensee decided to revise Emergency
Operating Procedures (EOPs) 7 and 8 to: (1) reverse the High Pressure
Injection (HPI) " piggyback" transfer prior to swapping the Building
Spray and Decay Heat Jumps to the Reactor Building (RB) sump, resulting
in isolation of the MJP from the_MUT earlier in the transfer sequence,
and-increasing the available NPSH for the HPI pumps, and (2) closing
,
MUV-64 if accessible.
In addition. a revision to the OPS Study Book
'
Entry would be made to describe the E0P changes.
The E0P changes and
the OPS Study Book Entry change were completed before close of the
inspection.
E8.3 (Closed) URI 50-302/96-03-07. Three Examoles of Desian Control Errors
a.
Insoection Scooe (92903. 37551)
This item was described in paragraph El.1 of NRC Inspection Report
50-302/96-03,
It involved two examples of calculation errors and an
example of an erroneously located inservice inspection boundary. The
item was considered unresolved pending an NRC rev'ew to determine if it
should be considered an additional example of an apparent violation
discussed at the predecisional enforcement conference on March 27, 1996,
b.
Observations and Findinas
On further review, the NRC determined that this item was substantially
different than the example referred to in the apparent violation.
c.
Conclusions
This item will be cited separately and will be identified as VIO
50-302/96-09-06. Erroneous calculation inputs and inservice inspection
boundary.
URI 50-302/96-03-07 is closed.
E8.4 (Closed) URI 50-302/96-04-06. Untimely Corrective Actions for the EFIC
System Concerns and Problems
-
a.
Insoection Scooe (92903)
This item was described in paragraph E2.4 of NRC Inspection Report
50-302/96-04.
It involved untimely corrective actions to address
Emergency Feedwater Initiation and Control (EFIC) system concerns and
problems.
The item was considered unresolved, pending an NRC review to
determine if it should be considered an additional example of an
apparent violation discussed at the predecisional enforcement conference
on March 27, 1996.
-
.
.
I
!
.
14
b.
Observations and Findinas
On further review, the NRC determined that this item was substantially
different and from the example cited in the apparent violation.
c.
Conclusions
This item will be cited separately and will be identified as-VIO
50-302/96-09-07, Untimely Corrective Actions for the EFIC System
Concerns and Problems.
URI 50-302/96-04-06 is closed,
ly
Plant Supoort
R8
Miscellaneous RP&C Issues
R8.1
(Closed) URI 50-302/96-04-07. Failure to Perform 10 CFR 50.59 Safety
Evaluation for Procedures Involvino Dissolved Hydroaen Concentration
Chanaes as Described in the FSAR
a.
Insoection Scooe (92904)
l
This issue was described in paragraph R1.1 of NRC Inspection Report
50-302/96-04 and was left unresolved, pending NRC review to determine if
.
it was an additional example of an apparent violation of 10 CFR 50.59
discussed at a predecisional enforcement conference on March 27. 1996.
The issue involved failure to perform a 10 CFR 50.59 Safety Evaluation
when the Reactor Coolant System dissolved hydrogen concentration was
changed from 15-40 cc/kg to 25-50 cc/kg.
b.
Observations and Findinas
Upon further review, the NRC determined this issue is substantially
different and not an additional example of the apparent violation.
Therefore, this issue is identified as VIO 50-302/96-09-03. Failure to
Perform 10 CFR 50.59 Safety Evaluation for Changes to Procedures
Described in the FSAR for Controlling Dissolved Hydrogen Concentration.
Paragraph R1.1 of NRC Inspection Report 50-302/96-04 noted that the
licensee has identified other examples where implementing procedures do
not agree with the FSAR. These exam)1es were identified as part of the
licensee's ongoing FSAR Operational Review Project, which is a detailed
review comparing the FSAR with the Design Basis Documents and other
plant documents to ensure consistency and compliance with the FSAR.
Although a significant number of discrepancies have been identified,
none have been determined to represent loss of safety function or to be
outside the design basis of the plant. As additional discrepancies are
identified that need further evaluation, they are being added to Problem
Report (PR)96-119 for resolution.
The review is scheduled to be
completed in February of 1997.
The resolution of these FSAR
discrepancies will be reviewed during future inspections of the results
of the FSAR Operational Review.
l
l
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15
c.
Conclusions
I
A violation has been identified as VIO 50-302/96-09-03. Failure to
Perform 10 CFR 50.59 Safety Evaluation for Changes to Procedures
Described in the FSAR for Controlling Dissolved Hydrogen Concentration.
URI 50-302/96-04-07 is closed.
j
I
P8
Miscellaneous EP Issues
ff
P8.1 TSC Emeraency Ventilation System (92904)
j
LER 50-302/95-14-01. Technical support center air flow deviates from
cable flow resulting in o]eration outside the' design basis was
act.,d on December 1. 1995. T1e NRC issued Deviation 50-302/96-15-05.
3
issue
'
Deviation from the design commitment for the Technical Support Center
emergency ventilation system.
By letters dated November 9, 1995.
February 7, 1996,' March 29, 1996, and April 30, 1996, the licensee
'
provided details of the TSC ventilation system modifications to make the
system function and to maintain the TSC habitable for postulated
radiological emergencies.
The licensee's corrective actions were
1
.
l
reviewed and found acceptoble as documented in IR 50-302/96-03,
paragraph P2 and IR 50-302/96-04. paragraph P8.1.
The Deviation. 50-
302/96-15-05. wa!.
% sed in IR 50-302/96-04. This LER is closed.
S1
Conduct of Security and Safeguards Activities
j
S1.1 Escort Duties (71750)
At approximately 11:45 a.m. on August 14, 1996, while exiting the TSC,
the inspector observed two vendor personnel, one visitor and one escort,
'
with a vending machine, waiting to exit in the hallway between tne TSC
and the TSC diesel generator room. All of the doors from the TSC into
the hallway were closed, and a security guard was unlocking the
i
emergency doors to cne outside to allow the machine to 5e removed.
A) proximately thirty seconds after the inspector enterad the hallway, a
4
t11rd vendor entered the hallway, unescorted, from the TSC.
This vendor
,
vas also wearing a visitor's-badge. There was no separate escort for
'
f
this visitor.
,
The inspector notified the security guard of the occurrence and
1
discussed the event with the security shift supervisor.
The supervisor
3
immediately initiated an investigation into the causes of the event.
The CR3 Physical Security Plan, paragraph 5.5.1. requires that all
.
personnel not issued a yellow or green identification badge he handled
i
as visitors and be escorted at all times while within the protected or-
1
vital areas.
The failure of the escort personnel to maintain visitor
personnel under surveillance at all times is a violation of the Physical
i
Security Plan and will be identified as VIO 50-302/96-09-02. Unescorted
'
visitor personnel within the protected area.
.
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,_
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16
y_,.
Manacement Meetinas
X1
Exit Meeting Summary
X1.1 During an interim exit meeting on August 30, 1996, the Director. Nuclear
Plant Operations questioned Violation 50-302/96-09-05 (relative to the
failure to change procedures after the modification of valve MUV-64)
since the valve lineup procedure was changed to show valve MUV-64 sealed
o]en.
The question was whether a violation existed if this 3rocedure
clange was made prior to identification of this problem by tie NRC. The
inspectors noted that change to the valve lineup procedure, without
additional in w w lons on when and under what plant conditions to
operate the va & does not satisfy the need for additional instructions
'
for the operators.
XI.2 The inspectica scope and findings were summarized on September 9.1996.
1
The inspectors described the areas inspected and discussed in detail the
inspection results listed below.
Proprietary information is not
contained in this report.
Except as noted in paragraph X1.1, dissenting
comments were not received from the licensee.
,
X3
Management Meeting Summary
,
X3.1 On August 28, 1996 a management meeting was held on site at FPC to
review the licensee'
CAP (Corrective Action Plan) to improve
performance.
A mating summary was issued on September 10, 1996.
' '
X3.2 On August 13, 1996, the inspectors attended an exit meeting between
licensee personnel and representatives of the U.S. Department of Labor,
'
Occupational Safety and Health Administration.
Discussions were held on
-
the areas of hearing protection programs and personnel protection from
,
rotating assemblies on various
aumps.
Final results will be presented
'
in the forthcoming report by OSiA.
No follow-up is planned by the NRC.
-
,
X4
Management Changes
'
X4.1 Personnel Chanaes
,
Mr. R. Enfinger was re) laced by Mr. D. Wilder as Manager, Safety
4
Assessment Team.
Mr. R. Enfinger has left FPC.
Mr. D. Wilder's former
position as Manager. Radiation Protection has not been filled at this
time.
The positions of Manager. Nuclear Production and Manager. Nuclear Outage
.
have been combined and will be headed up by Mr. H. Koon.
Mr. B. Moore, former Manager, Nuclear Production will assume a position
as Nuclear Shift Manager.
Mr. G. Wilson formerly one of six Nuclear Shift Managers, will become a
member of the bafety Assessment Team.
.
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17
Mr. J. Baumstark, formerly Assistant to Sr. Vice President, Nuclear
Operations, will assume the position of Director of Quality Programs.
!
effective October 1, 1996.
PARTIAL LIST OF PERSONS CONTACTED
Licensees
K. Baker, Manager. Nuclear Configuration Management
P. Beard, Senior Vice President Nuclear Operations
G. Boldt, Vice President Nuclear Production
J. Campbell. Manager, Nuclear Security
J. Cam) bell. Assistant Plant Director, Maintenance and Radiation Protection
W. Conclin, Jr., Director., Nuclear Operations Materials and Controls
R. Davis. Assistant Plant Director. Operations and Chemistry
D. DeMontfort, Manager, Nuclear Operations
M. Donovan, Supervisor. Rapid Engineering Response Team
R. Fuller, Manager. Nuclear Chemistry
B. Gutherman, Manager, Nuclear Licensing
G. Halnon. Assistant Director. Nuclear Operations Site Support
V. Hernandez. Employee Concerns Representative
B. Hickle Director, Nuclear Plant Operations
,
L. Kelley, Director Nuclear Operations Site Support
'
H. Koon Manager, Nuclear Production
'
!
K. Lancaster, Manager, Nuclear Projects
J. Maseda, Manager Engineering Programs
P. McKee. Director Quality Programs
-
R. McLaughlin. Nuclear Regulatory Specialist
j
'
W. Rossfeld, Manager. Site Nuclear Services
.J. Stephenson, Manager. Radiological Emergency Planning
,
F. Sullivan Manager. Nuclear Engineering Design
3
'
i
J. Terry. Manager, Nuclear Plant Technical Support
R. Widell. Director, Nuclear Operations Training
'
D. Wilder. Manager, Safety Assessment Team
,
K. Wilson, Principal Engineer, Operations
.
E
W. Bearden, Reactor Inspector Region II (August 19-23, 1996)
S. Cahill. Resident Inspector. Watts Barr (August 28-30, 1996)
K. Clark. Public Relations Region II (August 28, 1996)
B Crowley, Reactor Inspector. Region II (August 26-30. 1996)
S. Ebneter. Regional Administrator, Region II (August 28, 1996)
R. Gibbs, Reactor Ins)ector. Region II (August 19-23, 1996)
i
4
.
A. Gibson, Director.
)iv. of Reactor Safety. Region II (August 28, 1996)
E.'Girard, Reactor Inspector. Region II (August 26-30, 1996)
i
F. Hebdon, Director. Directorate II-3. NRR (August 28, 1996)
J..Hufham. Incident Res)onse Coordinator. Region II (August 22. 1996)
J.~ Jacobson. IPAP Team _eader. NRR (August 28, 1996)
J. Johnson Acting Di_ rector. Div. of Reactor Projects. Region II (August 28,
,
199G)
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18
K. Landis, Branch Chief. Region II (August 28, 1996)
L. Raghavan. Project Manager. NRR (August 28, 1996)
S. Varga, Director, Div. of Reactor Projects I/III. NRR (August 28, 1996)
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 40001:
Resolution of Employee Concerns
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving and
,
Preventing Problems
IP 61726:
Surveillance Observations
IP 62707.:
Conduct of Maintenance
IP 71707:
Plant Operations
,
IP 71750:
Plant Support Activities
IP 92902:
Followup - Maintenance
'
IP 92903:
Followup - Engineering
3
IP 92904:
Followup - Plant Support
i
ITEMS OPENED, CLOSED, AND DISCUSSED
i
i
Ooened
Typ_e Item Number
Status
Descriotion and Reference
l
50-302/96-09-01
Open
Failure to follow a procedure
resulting in the inadvertent
initiation of the control room
emergency ventilation system.
-
(paragraph M1.1)
,
50-302/96-09-02
Open
Unescorted visitor personnel within
'
i
the protected area. (paragraph S1.1)
!
50-302/96-09-03
Open
Failure to perform 10 CFR 50.59
.
safety evaluation for changes to
.
procedures described in the FSAR for
.
.
controlling dissolved hydrogen
concentration. (paragraph R8.1)
50-302/96-09-04
Open
Failure to update operating curves
L
to reflect 1981 power uprate.
(paragraph E8.1)
4
50-302/96-09-05
Open
Failure to incorporate design
information into operations
-
j
procedures. (paragraph E8.2)
50-302/96-09-06
Open
Erroneous calculation inputs and
'
inservice inspection boundary.
(paragraph E8.3)
i-
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50-302/96-09-07
Open
Untimely corrective actions for the
EFIC system concerns and problems.
(paragraph E8.4)
Iygg Item Number
Status
Descriotion and Reference
LER 50-302/95-014-01
Closed
Technical support center air flow
deviates from acceptable flow
resulting in operation outside the
i
design basis. (paragraph P8.1)
J
LER
50-302/95-27-00
Closed
Leak in underground emergency
,
feedwater piping suspected to be
'
caused by nearby evacuation using
power tools. (paragraph M8.1)
50-302/96-03-09
Closed
Operating curves not promptly
updated to reflect 1981 power
uprate. (paragraph E8.1)
50-302/96-04-09
Closed
Failure to incorporate design
information into operations
procedures. (paragraph E8.2)
50-302/96-04-07
Closed
Failure to perform 10 CFR 50.59
safety evaluation for procedures
involving dissolved hydrogen
concentration changes as described
'
in the FSAR. (paragraph R8.1)
50-302/96-03-07
Closed
Three examples of design control
errors. (paragraph E8.3)
50-302/96-04-06
Closed
Untimely corrective actions for the
EFIC System concerns and problems.
(paragraph E8.4)
,
LIST OF ACRONYMS USED
ac
- Alternating Current
ADI
- Absolute Drift Indications
AHD
- Air Handling Vent and Cooling Damper
-AHV
- Air Handling Vent and Cooling Valve
AI
- Administrative Instruction
ALARA - As low as Reasonably Achievable
ANSI
- American National Standards Institute
ANSS
- Assistant Nuclear Shift Supervisor
APC
Alternate Plugging Criteria
-ASME
- American Society of Mechanical Engineers
ASV
- Auxiliary Steam Valve
- Boiler and Pressure Vessel
,
.
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20
Babcock & Wilcox
-
Building Spray
BS
-
BSP
- Building Spray Pump
BVT
- Below Voltage Threshold
BWST
- Borated Water Storage Tank
- Confirmatory Action Letter
- Corrective Action Plan
- Closed Circuit Television
Condensate Pump
-
Code of Federal Regulations
CFR
-
- Core Flood Tank
CFV
- Core Flood Valve
- Comoliance Procedure
Con'rol Room Emergency Ventilation System
CREVS -
t
CR3
- Crystal River Unit 3
- Condensate Storage Tank
- Circulating Water Pump
dc
Direct Current
-
- Decay Heat Closed Cycle Cooling
DCHE
- DC Heat Exchanger
DEV
- Deviation
- Diesel Fuel Pump
DH
- Decay Heat
DHHE
- Decay Heat Heat Exchanger
DHP
- Decay Heat Pump
- Decay Heat Pemoval
DHV
- Decay Heat Valve
- Departure from Nucleate Boiling Limits
DNP0
- Director. Nuclear Plant Operations
dp
- Differential Pressure
- Enforcement Action
- Emergency Core Cooling System (s)
EDBD
- Enhanced Design Basis Document
- Escalation Enforcement Item
Emergency Feedwater Initiation and Control
-
- Emergency Feedwater Pum)
- Emergency Feedwater Tanc
- Emergency Feedwater
EFV
- Emergency Feedwater Valve
EGDG
- Emergency Plan Implementing Procedure
E0P
- Emergency Operating Procedure
- Engineered Safeguards
- Engineered Safeguards Feature
- Engineered Safety Actuation System
-
EVS
- Emergency Ventilation System
F
- Fahrenheit
- Florida Power Corporation
- Final Safety Analysis Report
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21
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FWHE
- Feedwater Heat Exchanger
'FWP
- Feedwater Pump _
FWV
- Feedwater Valve
GL:
- Generic Letter
g am
- Gallons Per Minute
,
- H)V
- Heater Drain Valve
_High Energy Line Break
- Health Physics
,
HPES
- Human Performance Evaluation System
- High Pressure Injection
'
HVAC-
- Heating. Ventilation and Air Conditioning
in. Hg - Inches of Mercury
- Instrumentation and Control
- Inadequate Core Cooling
- Integrated Control-System
IEEE
- Institute of Electrical and Electronics Engineers
-IFI
- Inspection Followup Item
INP0
- Institute of Nuclear Power Operations
IR
- Inspection Report
.
,
'
- Instrument Society of America
l
- Inservice Inspection
. ISO
- Isometric Drawing
- Inservice Test
- Improved Technical Specification
1
JC0
- Justification for Continued Operation
Kv
- Kilovolt
Kw
- Kilowatt
LCO
- Limiting Condition for Operation
LER
- Licensee Event Report
LOCA-
- Loss of Coolant Accident
LOV
- Lube Oil Valve
LTE
- Lower Tube End
LTS
- Lower Tube Sheet
- Modification Approval Record
MCB
- Main Control Board
- Motor Control Center
- Main Feedwater
- Motor Operated Valve-
MOVATS - Motor Operated Valve Analysis and Test System
- Maintenance Procedure
MRP-
-Management Review Panel
-MSV-
--Main Steam Valve-
-. Magnetic Particle Testing
MU
- Make Up
- Make-up Pum)
MUT
- Make-up-Tant
MUV
-- Make-up Valve
MW-
Megawatt
r
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NEP
- Nuclear Engineering Procedure
NMI
- Nuclear Monitoring Instrumentation
N00
- Nuclear Operations Department
- Net Positive Suction Head
NOI
- Non-Ouantifiable Indication
NRC
- Nuclear Regulatory Commission
- Office of Nuclear Reactor Regulation
,
NSM
- Nuclear Shift Manager
- Nuclear Steam System Supplier
NUREG - NRC technical report designation
- Operability Concerns Resolution
OP
- Operating Procedure
OSB
- Operations Study Book
- Occupational Safety and Health Administration
- Once Through Steam Generator
- Preventive Maintenance
- Power Operated Relief Valve
apb
- Parts Per Billion
3R
- Problem Report
- Plant Review Committee
- Preservice Inspection
.
asig
-
aounds aer square inch gauge
3T
-
_iquid penetrant
- Pressure and Temperature Limits Report
OC
- Quality Control
OA
- Quality Assurance
4
OAP
- Quality Assurance Procedure
- Reactor Building
RC
.
- Radiation Control Area
,
- Reactor Coolant Pump
RCPPM - Reactor Coolant Pump Power Monitor
REA
- Request for Engineering Assistance
RF0
- Refueling Outage
- Regulatory Guide
R0
- Reactor 0)erator
RPC
- Rotating Jancake Coil
RP&C
- Radiological Protection and Chemistry
- Revolutions Per Minute
Ri
- Radiographic Inspection
RN
- Nuclear Services and Decay Heat Seawater
- Nuclear Services and Decay Heat Seawater Pump
RWV
- Nuclear Services and Decay Heat Seawater Valve
}
- Systematic Assessment of Licensee Performance
- Systems Approach to Training
- Secondary Closed Cycle Cooling Pump
<
SDT
- Station Drain Tank
- Safety Evaluation Report
SFPD
'- Safety Function Determination Program
,
.
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23
- Significant 0)erating Event Report
- Surveillance )rocedure
SR-
- Surveillance Requirement
'
SS00
- Shift Supervisor on Duty
.
- Secondary Cycle Sampling Valve
- Stop. Think. Act. Review
!
- Short Term Instruction
- Nuclear Services Closed Cycle Cooling System
SWHE
- SW Heat Exchanger
SWP
- SW System Pump
,
SWV
- SW System Valve
TBF
- Turbine Generator Fan
.
T
- Cold Leg Temperature
T1
- Temporary Instruction
.
.TMAR
- Temporary Modification Approval Record
'
- Three Mile Island
-
TPM
- Task Performance Manual
'
TS
- Technical Specification
+
'
TSCR-
- Technical Specification Change Request
TW
- Through Wall
UAf
- A measure of heat exchanger effectiveness
- Unresolved. Item
.
- United States of America Standards
o
UT'
- Ultrasonic Test
,
- Violation
'
V0TES - Valve Operation Test and Evaluation System
V)p
- Volts point-to-point
,
- Work Request
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