IR 05000400/1988039
| ML18005A739 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/20/1988 |
| From: | Bradford W, Fredrickson P, Shannon M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18005A738 | List: |
| References | |
| 50-400-88-39, NUDOCS 8812300201 | |
| Download: ML18005A739 (14) | |
Text
gp,R PEG'~
'o
- 'lQyp'.
Jere* <<<<
UNITED'STATES NUCL'EAR'REGULATORY'COMMISSION
REGION II
101 MARIETTAST., N.W.
ATLANTA,GEORGIA 30323 Report No.:
50-400/88-39 Licensee:
Carolina Power and Light Company P.
0.
Box 1551 Raleigh, NC 27602 Docket No.:
50-400 Facility Name:
Harris 1 Inspection Conducted:
October 21 - November 20, 1988 Inspectors:
ra or License No.:
NPF-63 a
e 1gne Z ga Approved by:
annon roc son, ec son le Reactor Projects Section lA Division of Reactor Projects a
e gne iz zy IE e
gne Scope:
Results:
SUMMARY This routine safety inspection was conducted in the areas of operational safety verification, surveillance observations, maintenance observations, radiological protection program, physical security program, licensee event reports, followup of events at operating power reactors, and followup of items of noncompliance and deviations.
Review of the October 30, 1988 post trip data identified that operator errors combined with secondary plant instrument calibration problems resulted in an unnecessary manual reactor trip.
In addition, there was also an apparent weakness in operator knowledge secondary plant equipment, paragraph 8.
Mithin the areas inspected, no violations or deviations were identified.
881230020i 88i22i PDR ADQCK 05000400
REPORT DETAILS Persons Contacted Licensee Employees R.
A. Watson, Vice President, Harris Nuclear Project
~C.
G. Hinnant, Plant General Manager C.
R. Gibson, Director, Programs and Procedures
"DE L. Tibbits, Director, Regulatory Compliance C.
ST Bohanan, Director, Special Programs
~R.
B.
Van Metre, Manager, Technical Support
"T.
C. Morton, Manager, Maintenance
"J.
M. Collins, Manager, Operations J.
R. Sipp, Manager, Environmental and Radiation Monitoring D.
A. Braund, Supervisor, Security T.
F. Lent, Systems Engineering W.
R. Wilson, Reactor/Performance Engineering L. J.
Woods, Testing and Maintenance Support W.
H. Batts, Supervisor, Mechanical Maintenance J.
H. Smith, Supervisor, Operations Support C.
S. Olenik, Supervisor, Shift Operations G.
L. Forehand, Director, gA/gC
.
F.
E. Millet, Manager, Outages and Modifications Other licensee employees contacted during this inspection included technicians, operators, mechanics, security force members, engineering personnel and office personnel.
~Attended exit interview Acronyms and initialisms used throughout this report are listed in paragraph 11.
Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the report period.
Also, the operability of selected emergency systems was verified, tagout records were reviewed, and proper return to service of affected components was verified.
Tours of the plant were conducted to observe plant equipment conditions, including fluid leaks and excessive vibration, and general housekeeping efforts.
The inspectors verified compliance with
- selected LCOs and results of selected surveillance tests.
The verifica-tions were accomplished by direct observation of monitoring instrumenta-tion, valve positions, switch positions, accessible pipe snubbers, and review of completed logs, records and chemistry results.
The licensee's compliance with LCO action statements was reviewed as events occurre The inspectors routinely attended meetings with certain licensee management and observed various shift turnovers.
These meetings and discussions provided a daily status of plant operations, maintenance, and testing activities in progress, as well as discussions of significant problems.
The inspectors reviewed the shift foreman's log, control room operator's
.log, clearance center tag out logs, system status logs, chemistry and health physics logs, and control status boards.
The inspector noted that the operators appeared to be alert and aw'are of changing plant conditions.
The following condition was reviewed:
lAV-EGA, Air Cleaning Unit RAB Emergency Exhaust:
During a routine visual inspection, the inspector noted that two of six unit prefilters were lying on the floor of the unit.
This item is still under review and will be further addressed in inspection report 50-400/88-40.
No violations or deviations'ere identified.
3.
Monthly Surveillance Observation (71709)
The inspectors witnessed the licensee conducting surveillance test activities on safety-related systems and components to verify that the licensee performed the activities in accordance with applicable
.
requirements.
These observations included witnessing selected portions of each surveillance, review of the surveillance procedure to ensure that administrative controls were in force, determining that approval was obtained prior to conducting the surveillance test, and verifying that the individuals conducting the test were qualified in accordance with plant approved procedures.
Other observations included ascertaining that test instrumentation used was calibrated, data collected was within the specified requirements of TSs, any identified discrepancies were properly noted, and the systems were correctly returned to service.
Portions of the following test activities were observed or reviewed by the inspectors:
a.
OST 1004 (Revision 3),
Power Range Heat Balance Daily Interval Mode
(Above 15K Power)
The purpose of this surveillance is to verify the accuracy of the nuclear instrumentation when compared to the secondary heat balance and to make the necessary adjustments to the nuclear instruments.
The inspectors reviewed the complete procedure and determined that the rated thermal power calculation was satisfactory and that the final nuclear instrument setpoints were conservative.
b.
OST 1026 (Revision 3),
Reactor Coolant System Leakage Evaluation Daily Intervals Modes 1-2-3-4 The purpose of this surveillance is to verify the primary coolant system l.eakage rate and to ensure the leak rate meets TS require-rapped~
~
r g
'~
'
~
- v<<"'i ~ ~,
'
~
<
+
~'cyw v
) ~,
~ i e, ~v q,~ v wprw. +tz+c
C.
ments.
On November 6, 1988, the identified leakage began increasing and was immediately attributed to a packing leak on the pressurizer loop
"B" spray valve.
The leakage increased to 4.5 gallons per minute and has remained constant.
No deficiencies were identified during the review of various leak rate tests.
PPP-218, Turbine Heat Balance Test Using Automatic Data Acquisition Method The purpose of this test is to verify overall plant efficiency.
At this time, the licensee has not completed the final review of data and to date no discrepancies have been noted.
The inspectors observed control room activities prior to and during the performance of the test.
No deficiencies were identified during the conduct of this test.
No violations or deviations were identified.
4.
Monthly Maintenance Observations (62703)
Station maintenance activities of safety-related systems and components were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and were in conformance with TSs.
Items considered during the review included verification that LCOs were met while components or systems were removed from s'ervice; approvals were obtained prior to initiating the work; approved procedures were used; completed work was inspected as applicable; functional testing and/or calibrations were performed'rior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials were properly certified; and radiological and fire prevention controls were implemented.
Work requests were also reviewed to determine the status of outstanding jobs to assure that priority was assigned to safety-related equipment maintenance which may affect system performance.
Portions of the following activities were observed or reviewed:
Feedwater Heater Level Control Valve LCY-1HP-23: This valve was found by the inspector to be cycling and not maintaining proper feedwater heater water level.
Twenty-one work requests were reviewed to obtain the previous failure history for this valve.
Main Turbine Valve IMS-GV-4:
During governor.
valve testing on November 1, 1988, the valve was observed to have dual indication following valve trip signals.
A review of ten previous work requests indicated that faulty limit switches are an ongoing problem.
Feed Mater Pressure Switch FW-PS-2200A:
A review of calibration data indicated that the wrong setpoint value was used by the I8C technician and the FM pump low suction pressure trip was set at 305 psig versus 250 psig as originally designe No violations or deviations were identified.
Radiological Protection Program (71707)
Selected activities of the licensee's Radiological Protection Program were reviewed by the inspectors to verify conformance with plant procedures and NRC regulatory requirements.
The areas reviewed included organization and management of the plant's health physics staff, ALARA procedures, personnel exposure records, observation of wor k and personnel in radiation areas to verify compliance to radiation protection procedures, and control of radioactive materials.
No violations or deviations were identified.
Physical Security Program (71881)
Licensee's compliance to the approved security plan was reviewed by the inspectors.
The inspectors verified by observation and interviews with security force members that measures taken to assure the physical protection of the facility met current requirements.
Areas inspected included organization of the security force; establishment and maintenance of gates, doors, and isolation zones; access control; and badging procedures.
No violations or deviations were identified.
Licensee Event Reports (92700)
The following LERs were reviewed for potential generic problems to determine trends, to determine whether information included in the report meets the NRC reporting requirements and-to consider whether the corrective action discussed in the report appears appropriate.
The licensee action was reviewed to verify that the event has been reviewed and evaluated by the licensee as required by the TSs; that corrective action was taken by the licensee; and that safety limits, limiting safety setting and LCOs were not exceeded.
The inspector examined the incident report, logs and records, and interviewed selected personnel.
The following reports are considered closed:
LER 88-01 Emergency Boration Procedure Deficiency for Switchover to Recirculation After a Loss of Coolant Accident This event report identified a procedure deficiency which had previously existed in the EOPs which could have resulted in the failure of the ECCS during the re'circulation phase of an accident, under the scenario of a single failure of one RHR pump.
The deficiency involved an improper valve lineup which could cause a
RHR pump to exceed its design flow limit during ECCS recirculation.
This could take place if the LOCA was
sufficiently large enough to completely depressurize the Reactor Coolant System, and if one of the two RHR pumps had failed leaving only one RHR pump operating.
The licensee conducted a safety evaluation which concluded the plant would have been in an unanalyzed condition.
Emergency Operating Procedure EPP-010,
"Transfer to Cold Leg Recircu-lation" was revised to require closure of one of the low pressure safety injection containment isolation valves to ensure that runout of the remaining RHR pump would not occur during recirculation if the other RHR pump failed.
This event occurred prior to issuance of the operating license.
Since the licensing of the plait, further changes were made which required
CFR 50.59 evaluations prior to design
'mplementation, and FSAR changes are made following changes to the plant or procedures.
LER 88-02 Maintenance Surveillance Test Not Performed as Required on Radiation Monitor RM IN/-3547-1 The vent stack wide range gas monitor was declared inoperable per TS requirements when the l.icense discovered that a
maintenance surveillance test MST-10380 had not been completed as scheduled.
The surveillance test was successfully completed on January 26, 1988, and the radiation monitor was declared operable.
Corrective action involved a revision to Plant Procedure PLP-103, Surveillance and Test Program.
LER 88-03 Reactor Overpower Due to Condenser Steam Dump Valves Opening Caused by Faulty Miring in Cabinet A loop calibration on turbine first stage pressure channel P-446 was in progress with the plant at 100'ower.
Two condenser steam dump valves opened and caused total steam flow to increase, which in turn increased reactor power to 103K.
Power was reduced to 98K and stabilized at lOOX.
A jumper was found around the output contacts providing turbine tr)p indication to the steam dump logic.
The jumper had been installed at the time of manufacture and allowed the steam dump logic to operate in the Tavg and steam pressure mode simultaneously when steam pressure mode was selected.
The licensee conducted a wire check of the solid state protection system and found additional wiring errors.
All wiring errors were corrected and each error was evaluated for safety significance.
'>>'.;, '<<>> t"
'>>, p>>>> qq ~ y~qi(g'>> nor,">>>>@per
<<>>,
. rm ga~<<>> p; gy e>>>>t>>g.,w,
>>
", I;. pg>
~ >>;">>>>>>w,i~a'o't, 'r~c.. t~,.P...ppppp~
LER 88-04 Low Level in Spray Additive Tank Due to Incorrect Level Indication The containment spray additive tank level was.. found to be less than that required by TS 3.6.2.2 after completing calibration of tank level indicators LT-ICT-7150SA and LT-ICT-7166SB.
The spray additive system was declared inoperable until the tank level was restored to normal.
The cause of the incorrect level indication was attributed to air in the sensing lines to the level transmitters.
The level transmitters were vented and calibrated.
LER 88-05 Turbine Building Vent Stack Monitor Sampling Room Air Due to Mispositioned Valves Caused by Inadequate Procedures Due to mispositioned valves, the monitor was sampling room air instead of vent stack air as required.
The procedure for restoration did not adequately address proper valve alignment.
The lineup was corrected, procedures were revised to include valve lineups, and independent verification is required to assure proper lineup.
I No violations or deviations were identified.
Onsite Followup of Events at Operating Power Reactors (93702)
On October 30, 1988, the plant experienced a loss of both heater drain pumps,
"B" condensate booster pump, and both main feedwater pumps.
After the loss of the last main feedwater pump, the operators initiated a manual reactor trip and a manual turbine trip.
The chronology of events is as follows:
Time (.m.
6:26:00 6:26:47 6:26:50 6'. 27: 14 6:27:16 6:28:00 6:28:22 Event
"Low delta pressure/low flow" for B
HDP received in control room.
Auxiliary operator dispatched.
B HDP tripped on low NPSH.
First turbine runback initiated automatically.
A HDP tripped on low FM Heater level.
Second turbine runback initiated automatically.
Bank low low rod insertion limit alarm received, rods placed in manual.
A and B
CBP speed controllers shifted to manual.
Runback complete, steam dumps closed, plant stable at 62K steam loa Time (.m.)
con Event 6:30:36 6:31:00 6:31:54 6:32:02
'6: 32: 32 6:32:44 6:32:45 6:33:36 B
CBP tripped on high discharge pressure (>60 seconds).
B MFP tripped manually.
Manual turbine runback initiated.
A and B AFM pumps manually started.
A MFP tripped on low suction pressure (>30 seconds).
Manual reactor trip.
Manual turbine trip.
A CBP tripped on high discharge pressure (>60 seconds).
The inspectors performed a detailed review of the completed OMM-004, Rev.
1, "Post Trip/Safeguards Review".
The review indicated that various plant deficiencies and questionable operator actions contributed to the event.
The following paragraphs detail the various trips, operator actions, and subsequent licensee followup.
B Heater Drain Pump Early in the morning on October 30, 1988, the licensee had noted
.
fluctuations in the NPSH meter for B HDP.
Because the potential of a low NPSH trip existed, power ascension was terminated at 98K power until the problem could be reviewed by technical support.
After the 6 p.m. shift turnover, on October 30, 1988, alarms were noted for low NPSH on the B
HDP.
An operator was dispatched to the HDP, but the pump tripped at 6:26 p.m.,
before any, corrective action could be initiated.
A Heater Drain Pump Following the initial loss of the B HDP, the main turbine ran power back to '86K turbine load as indicated by 'first stage pressure.
The runback caused steam flow perturbations to the feedwater heaters and the HDP discharge valve and recirculation valve controls were not rapid enough to prevent loss of level in the 4A and 4B feedwater heaters.
Since the plant does not have a heater drain tank, the HDPs must rely on the volume in the 4A and 4B feedwater heaters.
Consequently, the A
HDP lost its suction supply and tripped at 6:27 p.m., following the initial turbine runback.
Turbine Runback As indicated above, following the initial trip of HDP B,
a main turbine runback was automatically initiated and reduced turbine load to
=-86K turbi'ne power.
Following the loss of HDP A, another turbine runback initiated and reduced turbine load to =62K.
Although there
was no apparent reason for the second runback at the time, the operators took no action to limit the runback.
Subsequent review by the licensee found that the turbine controls responded as designed and that if the operators had placed the turbine controls in manual, a
runback to 0% power would have resulted.
Accordingly, plant operating procedures are being changed to address operator actions to defeat the turbine controller design which caused the second runback.
Main Feedwater Regulating Valves Following the turbine runbacks the steam generator water level's dropped from 65K to 60K and the control room operator placed the FRVs into manual in order to restore steam generator levels.
In manual, the operator is to maintain steam generator.levels at =65K; however, over a
four minute period, the operator overfilled two steam generators by lOX and then lowered the level in C steam generator to just above the 38.5X trip setpoint.
A graph of the C steam generator level versus time of event indicates that the C steam generator water level would probably have reached the low level reactor trip setpoint within 20 seconds even if the last running main feed pump had not tripped.
It appeared that steam generator levels were difficult to control adequately with FRVs in manual during the transient and therefore should be maintained in automatic whenever possible.
Operator actions appeared to be inadequate in that the C steam generator steam flow and feed flow were never matched and resulted in a rapid decrease in level on steam generator C.
8 Condensate Booster Pump During the initial runback, both of the CBPs automatic pressure controllers switched to manual at 90K CBP speed.
Normally, the controllers would adjust CBP speed in order to maintain a constant main feed pump suction pressure at 430 psig.
During the event, an operator was assigned to manually maintain the MFP suction pressure.
Following the turbine runbacks, and after reducing FRV flow due to two high steam generator levels, the CBP pressure increased and manual reduction in speed was required.
CBP discharge pressure reached 607 psig and remained above the trip reset for 60 seconds, resulting in a trip of the 8 CBP. It was noted that no attempt was made to return the CBP controls to automatic during this three minute evolution.
A subsequent review by the licensee indicated that by design, the high discharge pressure trip should have actuated at 625 psig,
.
indicating that it had apparently drifted low.
Further discussions with technical support engineers noted identified problems with the reset function of the high discharge pressure trip.
The reset function required a minimum reduction in pressure of 120 psig, which could reduce pressure below the normal system operating pressure.
The operations personnel were not aware of this deficienc B Main Feed Pump Following the loss of CBP B, the operator increased the speed of the A
CBP in order to maintain MFP, suction pressure.
NFP 8 suction pressure increased from 327 psig to 350 psig.
At 6:30 p.m.,
the operator unnecessarily manually tripped the B MFP because he believed that a
NFP was supposed to trip when its associated CBP trips.
Following a review by the inspector and licensee, this was shown not to be the case.
By design, the NFPs have a low suction pressure trip when less than 250 psig for 30 seconds; therefore, the NFP was not in jeopardy.
Accordingly, the associated training given to the operator was found to be inaccurate.
It was also noted that the operator delayed tripping the B
NFP for 24 seconds even though training indicated the pump should have tripped automatically.
A Nain Feed Pump Following the B
NFP manual trip, the operators noted that reactor power was indicating 72X and that a turbine runback to 60K was expected due to the loss of the MFP.
However, The turbine operator did not initiate any appreciable turbine unloading for over
seconds.
A review of post trip data. showed that the main turbine was already at
-=62K due to the loss of the second HDP earlier in the event.
During this same period, the A
CBP operator stated that he was manually decreasing speed of the A CBP in order to maintain the A MFP suction pressure at 400 psig.
It appeared that the operator was monitoring the CBP discharge pressure gauge in error and consequently reduced the CBP discharge pressure to 385 psig.
This resulted in a drop of MFP suction pressure to 314 psig.
The MFP suction pressure remained
,low for 30 seconds and resulted in MFP A tripping.
A subsequent review of I8C calibration records confirmed that the A NFP low suction pressure trip setpoint was improperly calibrated at 330 psig versus the original design value of 250 psig.
It was also noted that prior to the A NFP trip, the operator started the A
and B auxiliary feedwater pumps and initiated auxiliary feedwater flow to the steam generators.
The AFM pumps are low capacity and have a
minimum effect in restoring steam generator levels when at high plant power levels.
Discussions with Technical Support engineers revealed that when the plant is operated at 50-60K, that there is little to no main feedwater flow into the upper section of the main steam generators.
As a result, cold (=-70 F) condensate storage tank water is injected directly into the upper section of the steam generators without mixing with normal feedwater.
This could cause undesirable thermal stress to the steam generators and feedwater pipin A Condensate Booster Pump Following the A
MFP trip at 6:32 p.m.,
the CBP discharge pressure increased to 656 psig.
The assigned CBP operator did not reduce the CBP speed and the A CBP tripped
=-60 seconds later.
It was also noted that the CBP discharge pressure of 656 psig exceeded
'the CBP discharge piping design pressure of 650 psig.
A subsequent evaluation by the licensee indicated that the transient would not affect continued plant operations.
Reactor Trip and Subsequent Restart A
reactor trip and turbine trip were manually initiated following the loss of the A MFP and prior to the steam generator low water level trip.
All operator actions and equipment operation appeared to be adequate following the reactor trip.
Although the root cause of the initiating event (ie. the B
HDP trip)
could not be determined, the unit was returned to operation on November 1, 1988, with the intent to evaluate
"at power" secondary system data.
The technical support group subsequently analyzed the
"at power data" and determined that feedwater heater 3A was experiencing a high feedwater differential pressure.
The feedwater heater was isolated and a large piece of plywood was found at the
.
tubesheet.
The plywood had been left in the feedwater heater following the earlier refueling outage inspection.
The inspectors discussed with the licensee their concerns over the operator errors, secondary plant instrument calibration problems, and apparent weakness in operator knowledge of secondary plant equipment that were displayed during this event.
In response to the inspector's concerns, the licensee stated that a task force had been formed to review the event and recommend appropriate corrective actions.
Pending further NRC review, this item is considered an Inspector Followup Item:
Review Findings and Resultant Corrective Actions from Licensee's Task Force on October 30, 1988 Reactor Trip (400/88-39-01).
No violations or deviations were identified.
9.
Followup of Items of Noncompliance and Deviations (92700)
a 0 (Closed) Violation 400/87-21-01; Failure to Maintain an Operable Air Lock Door Closed.
This violation involved unacceptable operation of the containment building personnel air lock doors.
The inspector reviewed the corrective action described in the licensee's letter of July 29, 1987, and verified that revisions have been made to Operations Management Manual (OMM)-001, Operations - Conduct of Operations and Health Physics Procedure (HPP)-046, Containment Power and Initial Post Shutdown Entries.
OMM-001 requires a containment
access logbook for accountability.
During the past refueling outage, the inspectors were logged in and out of the containment building.
HPP-046 requires that the main control room be notified of entry and egress to the containment building.
This item is closed.
(Closed) Violation 400/87-31-01; Incorrect Position of Compressed Air Valve During Clearance Restoration.
This violation involved failure to follow established procedures in that the correct operational valve position was not specified in the restoration section of the maintenance work clearance.
The valve was not positioned in the open position.
This caused a loss of instrument air and a resultant trip of the reactor on low steam generator water levels.
The inspectors reviewed the corrective action described in the 1'icensee s letter of September 23, 1987, and verified that revisions had been made to Administrative Procedure 020, Clearance Procedure.
The procedure has been changed to require an independent review and signoff of the lineup preparation and restoration lineup.,
This item
>s closed.
(Closed)
Violation 400/87-34-01; Failure to Comply With Reactor Containment Building Technical Specifications.
This violation involved a violation of TS 3.6.3 in that steam generator lA blowdown valve (lBD-ll) became i.noperable.
This valve is a containment isolation valve.
This violation was the subject of an Enforcement Conference on October 21, 1987, in the NRC Region II offices in Atlanta, Georgia.
The inspector reviewed the licensee's letter of October 29, 1987, which details the corrective action.
Closed system valves which are isolation valves have been identified and written instructions have been provided to the operators.
In addition, procedural changes have been made to provide TS interpretation.
This item is closed.
(Closed) Violation 400/87-37-01; Operators Manipulating Valves Known to Operate in Unsafe Manner.
This violation involved operation of certain valves, even though it was known that opening these valves could cause uncontrolled opening of other valves.
This was authorized by the Shift Foreman and allowed a vent path for reactor coolant to containment atmosphere and into the pressurizer relief tank.
The inspectors reviewed the licensee's letter of February 26, 1988, which detailed the corrective action on this matter.
This corrective action included a review of valve design, sensitivity to plant operations which may cause safety concerns, and a review of the testing requirements of these valves in TS 4.4. 11. 1.
On February 12, 1988, Amendment No.
4 to the Harris Operating License NPF-63 was issued which deletes Surveillance Requirement 4.4. 11. 1 and also modifies Surveillance Requirement 4.4. 11.2b which changes the
testing requirements of block valves in question to once every
months.
The licensee has changed the ISI program to require testing only during shutdown.
This item is closed.
Exit Interview The inspection scope and findings were summarized during management interviews throughout the report period and on November 23, 1988, with the Plant Manager and selected members of his staff.
The inspection findings were discussed in detail, high lighting the concerns associated with the Inspector Followup Item listed below.
The licensee acknowledged the inspection findings and did not identify as proprietary any material reviewed by the inspectors during this inspection.
Item Number 88-39-01 List of Initialisms Status Open Descri tion/Reference Para ra h
IFI - Review Findings and Resultant Corrective Actions from Licensee's Task Force on October 30, 1988 Reactor Trip.
Paragraph 8.
AFM ALARA ASME BD CBP CFR CT ECCS EOP Eg FRV FSAR FM GPM GV HD HDP HPP IFI ISI LCO LER LOCA LT MFP MS MST Auxiliary Feedwater As Low As Reasonably Achievable American Society of Mechanical Engineers Blowdown Condensate Booster Pump Code of Federal Regulations Containment Spray System Emergency Core Cooling System Emergency Operatinq Procedure Environmental gualsfication Feedwater Regulatinq Valve Final Safety Analysis Report Feedwater Gallons Per Minute Governor Valve Heater Drain System Heater Drain Pump Health Physics Procedure Inspector Followup Item Inservice Inspection Limiting Condition for Operation Licensee Event Report Loss of Coolant Accident Level Transmitter Main Feed Pump Main Steam Maintenance Surveillance Test
No.
NPSH NRC OMM OST PLP PPP PS PSIG RAB RCS REM/RM RHR RWP TS WL Number Net Positive Suction Head Nuclear Regulatory Commission Operations Management Manual Operational Surveillance Test Plant Procedure Plant Performance Procedure Pressure Switch Pounds per Square Inch - Gauge Reactor Auxiliary Building Reactor.
Coolant System Radiation Monitor Residual Heat Removal Radiation Work Permit Technical Specification Liquid Waste