ML20042G630
| ML20042G630 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/09/1990 |
| From: | Hartzell C, Owen T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-90-025, NUDOCS 9005150182 | |
| Download: ML20042G630 (17) | |
Text
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PO llox 2S6 ' k Cloist, S C 29710 l l DUKE POWER May 3,.1990 t Document: Control. Desk '
-U. S. Nuclear Rogulatory Commission ! -Washington,.D. C. 20555 L.
Subject:
. Catawba-Nuclear Station
. Docket No. 50-413
.LER 413/90-25 Gentlemen:
' Attached is.Licensco. Event Report 413/90-25 concerning P-12 (LO-LO '
- TAVE) ENGINEERED SAFETY FEATURES SYSTEM ACTUATION DUE TO A DEFECTIVE PROCEDURE. l This event was contidored to be. of no significance wit) '
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to the health and safety of the public. j Very tur.ly yours,.
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Tony B. n-Station' Manager .; o ,^ .
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xc: Mr..S. D. Ebneter American Nuclear Insurers Regional Administrator, Region II c/o Dottie Sherman, ANI Library. U. ' S. Nuclear Regulator Comission The Exchango, Suite 245 -i 4 101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT 06032 h M & M Nuclear Consultants Mr. K. Jabbour h 1221 Avenues of the-Americas U. S. Nuclear Regulatory Commission ; New York, NY 10020 Office of Nuclear Reactor Regulation 'l Washington, D. C. 20555 i INPO Recoran Center
=Buito 1500 Mr. W. T. Orders 1100 Circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station h4 PDC
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'"P-12 (Lo Lo Tave) Engineered Saf ety Features System Actuation Duc To A Defective Procedure and Deficient Communication IVtNT DATE (Si LlR NUMetR Igl REPORT DATE 17) OTHE R F ACILITit$ INVOLVED (86 MONTH DAY y(AA Y(AR " MONTH DAY e ACittiv hauts DOCKEY NUM8tRtsi n f YEAR N/A 0l5l0l0l0l l l
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LSSTR ACT fomse ro i400 speses d e , eepree<=seet fureen sme's space typewntrea smesi (161 > On April 9, 1990, at approximately 0847 hours, with Unit 1 in Mode 3, Hot Standby, an unexpected Engineered Safety Features (ESP) Actuation occurred ;
-during the performanco of the Incore Thermocouple and RTD Cross Calibration Testing per IP/1/A/3231/01. The ESF Actuation, P-12 (Lo Lo TAVE) closed the steam dump valves due to an inadvertent low test signal inserted into the 7300 Process Control Cabinet (PCC). When the Reactor Coolant (NC) System RTD signals were reinstated, the Steam Dump valves opened, resulting in a brief cooldown which caused a shrink in Pressurizer level and Chemical and Volume Control'(NV)
System letdown isolation. Subsequent actions were taken to restore NV letdown L - and stabilize the plant at 557 degrees F while clearing the P-12 interlock permissive. This incident is attributed to a deficient procedure in that no precautions concerning the P-12 interlock and no requirement for verification of proper RTD simulator signals existed; and to deficient communication between Maintenance Engineering Services and Operations on the scope and effect of the test. Inappropriate action by Control Room Operators is considered a contributing cause. Corrective action will include revision of IP/1/A/3231/01 and review with Operations personnel the lessons learned from this event. l. 1 I l l-l l \ l
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UCENSEE EVENT REP 3RT (LER) TEXT CONTINUATION Arcovio oMe No. am-oio. I 8 EXPIRES: $/31/W PACILITY NAME (14 oOCKET NVhWER m HRNMIRW Pm @
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0l2 Ol8 l nxi w mme sRm a w me ***ewwc wm mw Im j i BACKGROUND ; Tho.Incore Instrumentation [EIIS:IG) (ENA) System provides information on the noutron flux distribution and fuel assembly outlet temperatures at selected coro ! locations. Using the information thus~obtained, it is possible to confirm the
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Reactor core design paramotors and calculated hot channel factors. The system providos means for acquiring data only, and performs no operational plant control. ! Tho_incore instrumentation consists of thormocouples positioned to measure fuel l assembly coolant outlet temperature, and incore flux thimbles to permit- l insertion of movable detectors [EIIS:XT] for measurement of the neutron flux I distribution within the Reactor core. Movable miniature neutron flux detectors g are available to scan the activo length of selected-fuel assemblies to provide. , remoto reading of the relative three dimensional flux distribution. j The purposes of the Incore Thermocouplo and RTD Cross Calibration are to: 1) ; measure the resistance of each Rosistance Temperature Detector (RTD) in the q Reactor Coolant [EIIS:AB] (NC) System at isothermal conditions; 2) record the j temperature of each thermoccuplo in the Incore Instrumentation System at isothormal conditions; 3) determine the installation error of each RTD; 4) , determino the computer [EIIS:IMoD] correction factors for each incore thermocouple; 5) provide data for the corrected loop calibration (cross calibration) of each RTD. 1
-The. test prerequisites require that the Unit is in Modes 3, Hot Standby, 4, Hot Shutdown, or 5_ Cold shutdown,-with the NC System constant within i 1.0 degree F, j Reactor. Trip Breakers [EIIS:BRK] are racked in-open, and all four NC pumps !
[EIIS:P] inservice unless RTD data is obtained before Unit heatup begins. The l calibrations are performed at three different temperatures of approximately 340 dogroos F,-450 degrees F, and 557 degrees F. With the NC System at isothermal conditions, data is taken from the NC System RTDs from the 7300 Process Control Cabinet using the Cross-Calibration Test set, . and the ENA System thermocouples from the Train A and B Plasma Display Monitors,- ' and the digital indicator [EIIS:XI] located on the Incore Instrumentation Panel. The Channel Test Relay [EIIS:RLY] switches [EIIS:XIS] on the Test Set remove the RTD signals from the 7300 Process Control System to allow the RTD signal to be monitored through the test set. Simulated RTD signals are inserted in the 7300 l- -Process Cabinets and adjusted to resistance values equivalent to the NC System l i temperature. The RTD switches are used to select a RTD, one at a time, to be road out on a precision digital multimeter. Four resistanco readings for each < RTD are recorded. q d
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MTf,W C;tiwba Nuclear Station, Unit 1 q0 0l3 TEXT (dr more apsee de regumt, use selsbeoner NNC Form JARCs107) o l5 lo lo lo l @ 1l3 9l 0 -- 0l2l 5 -- 0F 0l8 An average RTD resistance is converted to the equivalent temperature value using , the appropriate temperature resistance for the given RTD. The average NC System
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temperature is calculated from the narrow range RTD equivalent temperatures.
-This average temperature is used to calculate RTD installation errors and thermocouple computer correction factors.
The individual hot and cold loop temperature signals required for input to the Reactor trip circuits and interlocks are obtained using Resistance Temperature Detectors (RTDs) installed in each Reactor coolant loop. The hot leg temperature measurement on each NC System loop is accomplished with three fast response narrow range RTDs spatially located 120 degrees around the hot leg. A wide range RTD is installed in each hot leg also. One fast response narrow range RTD is located in each loop cold leg at the discharge of the Reactor Coolant system pump. In addition, one wide range RTD is installed in each cold leg. i The cold leg. temperature measurement, along with the average T-hot obtained from the three hot leg temperatures, is used to calculate the Reactor coolant loop Delta T and Tave. The purposes of the Steam Dump [EIIS:JI] (IDE) System are to: 1) enable the Reactor to follow Main Turbine [EIIS:TRB] load recuctions of >-10% and 30% stop : changes; 2) allow Unit: load reduction from 100% to plant auxiliary loads without 'l a Reactor trip; 3) allow a Turbine trip and Reactor trip from 100% without lifting the Main Steam [EIIS:SB) (SM) System Safety Valves [EIIS:V). The system accomplishes its purpose by the use of five banks of dump valves divided into condenser dumps and atmospheric dumps. Condenser dump valves are divided into-three banks with three valves per bank. Atmospheric dump valves are divided
-into two banks-with four.and five valves per bank, respectively. The total
' capacity of these dump valves is 71.5% of total Unit capacity.
The condenser and atmospheric dump valves are controlled by ona of three .
. controllers [EIIS:XC) (steam pressure, load rejection, plant trip). The selected controller actuates to control-Tave at or near a set reference signal.
The reference signal to the dump valves is filtered through a pneumatic circuit which contains block valves and arming valves. This " block" circuit prevents cooldown below 553 degrees F to ensure Tave remains above the minimum temperature for criticality. - The P-12, Lo Lo Tave Interlock, is part of the Engineered Safety Features Actuation System. The purpose of the interlock is to block steam dump valvo actuation to prevent excessive cooldown below the minimum temperature for criticality. The setpoint for P-12 is 553 degrees F on any two of four NC System loops. Catawba Technical Specifications state: the Engineered Safety Features Actuation System (ESFAS)' instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABIE per the requirements below, ozu mu m.s. am one we w
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0l2l 5 - 0l0 0l4 OF 0 l8 TEXT tar move apsee e soeuseer, ues saweener Nac poem JesAw (173 APPLICABILITY: As shown in Table 3.3-3 (A portion of table is shown) MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT- OF CHANNELS 'TO TRIP OPERABLE MODES ACTION Low-Low Tavg, 4 2 3 1,2,3 20 P-12' ACTION 20 - With less than-the Minimum Channels OPERABLE, within 1 hour- l determine by observation of the associated permissive status , light (s) that the interlock is in its required state for the l existing plant condition, or apply Specification 3.0.3. l l EVENT DESCRIPTION On April'2 a'nd'6, 1990, with Unit 1 in Mode 4, Hot Shutdown, Maintenance Engineering Services (MES) Engineers performed the Incore Thermocouple and RTD
' Cross Calibration tests for the 340 degrees F and 450 degrees F temperature plateaus, respectively.- Prior to performing each of these tests, a " tailgate" meeting was held between MES and Operations (OPS) personnel to discuss the impact of the test on plant status and operation. .The 340 degrees F temperature plateau-meeting was the most detailed and thorough in that all required equipment, plant status requirement, and applicable Technical Specifications mentioned in the procedure, IP/1/A/3231/01, were discussed and reviewed. The 340. degrees F and 45b degrees F temperature plateau tests were completed without :
F incident. On April 8,l1990, with Unit 1 in Mode 3, MES Engineers prepared and setup.the RTD Cross 1 Calibration Test Set per IP/1/A/3231/02 steps 10.2.1. through 10.2.9.
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'" -In addition,'the resistance.for each RTD temperature was measured and recorded in the simulator resistance' column on Enclosure.11.1.2 of the procedure. This data was then used to set the RTD simulators values per step 10.2.12. All work
.was documented in the appropriate enclosures of the procedure and independently verified.
- On' April 9, 1990, with Unit 1 in Mode 3, MES Engineers were conducting Incore
; Instrumentation and RTO Cross-Calibration Testing (557 degrees F Temperature
-Plateau);per IP/1/A/3231/01. The plant was maintaining Reactor Coolant System
'(NC) Tave at approximately 557 degrees F.using the Steam Dump valves with the Reactor Trip breakers racked in-open and four NC pumps operating to provide heat ,
to the Steam Generators [EIIS:HX]-(S/Gs). The discussion between OPS and MES personnel prior to beginning the 557 degrees F temperature plateau test was brief and only conveyed that the test consisted of measuring RTD temperature at p _M "* '"' " " ' " " * " ' " "
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATIDN' enoveoousNo.siwein EXPlRES. 8/31/5 FACILITY NAME (1) _ DOCKET NUMBER (2) LIR NUMSER 16) PAGE (3) v8'a
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.the 7300 Process Control Cabinet (PCC). At approximately 0830 hours, MES Engineers proceeded, per procedure, to isolate the NC System RTD signals from j the Control Room (C/R) 7300 PCCs while inserting the RTD Simulator test signals l by placing the Channel Test Relay Switches No. I through 24, located on the back ;
of the Test Set, in the closed (up) position. The RTD Simulator test signals i
- which had been previously set for 557 degrees F on April 8, 1990, were actually j low by approximately 10 degrees F. The low RTD simulator test signals '
simultaneously generated the Low and Emergency Low Tave annunciators and the P-12 permissive light on the Main Control board and the P-12 (Lo Lo Tave) interlock circuitry which caused the Steam Dump valves to close. With the Steam Dump valves isolated, S/G temperatures and pressures slowly increased during the , 15 minute test duration. In addition, the Steam Dump valve controller indicated a 30% demand for the valves. At approximately 0847 hours, at the end of the i test, the RTD Simulator test signals were removed from the C/R 7300 PCC and the actual NC System RTD signal was reinstated. i The Steam Dump valves, responding to the actual NC System loop temperature 4 (approximately 560 degrees F) sensed by the RTDs and steam header pressure (1125 psig), reopened. With the Steam Dump valves open, NC System temperature reduced to approximately 547 degrees F; causing a P-12 actuation as the setpoint of 553 degrees F was reached. Simultaneously with the P-12 actuation; the Low and EmergenJy Low Tave annunciators and the P-12 status light were lit. The cooldown also caused a shrink in Pressurizer level and NV System letdown j isolation. Subsequent actions taken by Control Room Operators (CROs) restored i NV System letdown and stabilized the plant at 557 degrees F. In addition, the Steam Dump Valves auto-opened as the P-12 interlock reset at 554 degrees F and main steam pressure increased above 1092 psig. , CONCLUSION l This incident is attributed to a deficient procedure. No precauti.on concerning l the P-12'(Lo Lo Tave) interlock was given in IP/1/A/3231/01, Incore Thermocouple and RTD Cross Calibration. In section 10.3 of the procedure, adequate cautions l are given which explain that performing the test has a potential to generate a Reactor. Trip and/or Feedwater Isolation Signal. However, the procedure lacks any precaution which explains that a potential exists to generate a P-12 signal. L In addition, the procedure did not indicate the necessary infcunation to be communicated to CROs on expected plant response. The procedure did not require verification of the RTD simulator signals which are inserted into the 7300 PCC and have a substantial effect on several interlocks and permissives. The sequence for removing the NC System RTD signals from service is considered inappropriate. Performing steps 10.3.1 through 10.3.3 removes and replaces all NC System RTDs with RTD simulator signals.
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.While these steps are being performed, all NC System RTD signals are
( simultaneously removed from service, leaving the Control Room Operators with j inadequate indication of NC System loop temperature. Furthermore, l simultaneously removing all: NC System RTDs from service while in Modes 1 through i 3, unnecessarily challenges Technical Specification 3.4.3.2, Engineered Safety l U Actuation System Instrumentation, and renders the P-12 Interlock inoperable. l i Another root cause for the incident is deficient communication between MES and 1 OPS Control Room personnel in that the discussion held prior to the 557 degrees ; F test was-less than adequate. MES Engineers performing the test did not convey j ' to OPS personnel the impact the test would have on Control Room indications, alarms, annunciators, interlocks, Technical Specifications, and plant systems (Steam Dump Valve System). In addition, OPS Control Room personnel did not , request enough information concerning the test to fully understand the impact { the test would have on the plant. This led to a less than adequate response by ! the Control Room Operators during the incident, f 5 I IP/1,2/A/3231/01 will be revised to include precautions concerning the effect on I the P-12 interlock and a requirement to verify the RTD simulator signals that are input into the 7300 PCCs. In addition, the procedure will indicate the - necessary information to be communicated to Operations. ; I l This incident is also assigned a contributing cause of inappropriate action in that the response by the CROs during the incident was not in accordance with their training and procedures. Because a less than adequate discussion was held ! prior to the test, Control Room Operators did not fully understand the test and j failed to recognize the need to respond to the low and Emergency Low Tave . annunciators and P-12 status light on the Main Control board. They. assumed that the output signals from the 7300 PCCs would be blocked and alarms and ] annunciators would be generated but no system actuations would occur. In ;
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addition, Control' Room personnel failed to enter T/S 3.4.3.2 in the Technical Specification Action Item Logbook (TSAIL) for the inoperable P-12 Interlock due . to all channels of RTDs being out of service simultaneously. However, all ! channels of RTDs were restored to operability in the one hour time frame .l specified in T/S Action Statement #20. 1 A thorough understanding of the test, including the impact on plant systems and interlocks, would have led to a more appropriate response by Operators which may have prevented the excessive cooldown and second P-12 actuation. Operations will emphasize to CRos the need to respond to all alarms and annunciators without assumptions as to their affects on existing plant conditions. A review of the OEP database for the past 24 months identified one previous incident where a deficient procedure led to an ESF actuation. LER 413/89-013 . involved an ESF actuation (Feedwater Isolation) due to Hi Hi Steam Generator level during performance of valve time testing. The root cause of the incident gow m. . .u.s. cr q as y es m y
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Cet:wba Nuclear Station. Unit 1 o l5 jo lo j o l @ 1l3 9l 0 - 0l2l 5 - 0l 0 0l7 oF 0l 8 ftXT y more esos 4,oguseer, ese esistooner MC Form JEM'st 07) was attributed'to inadequate procedure precautions concerning Unit status at the
-time of the test.- Corrective actions included a revision to the periodic test which was reflected in the outage schedule. Thus, this incident'is considered to be a recurring event in accordance with the Duke Power Company Nuclear Safety Assurance guidelines.
CORREC*IVE ACTION SUBSEQUENT
- 1) Control Room Operators restored NV System letdown, Pressurizer level, and stabilized the plant at 557 degrees F.
- 2) All testing (Rod Drop, RTD Cross Calibration, and RTD Response Time) was terminated or delayed until.the incident'was fully evaluated and understood.
- 3) IAE reviewed Rod Drop, RTD Cross Calibration, and RTD Response Time tests to determine the effect on plant operation.
- 4) The incident was explained and discussed with Operations Shift .,
Supervisors in the Shift Supervisor meeting held on April 27, 1990.
- 5) Operations emphasized to their-shift personnel, via an update on this incident, the need to evaluate the effect on existing plant conditions ,
of all alarms received during testing. ! PLANNED
- 1) IP/1,2/A/3231/01 will be revised to explain the effect on P-12, Technical Specifications, and plant conditions. In addition, the revisic> will include a method to remove RTDs from service one channel at a time.
- 2) NES will review and evaluate the'necessary qualifications and training required for conducting the-Incore Thermocouple and RTD Cross Calibration testing.
3)- Compliance will review if P-12 should be considered an ESF actuation.
- SAFETY-ANALYSIS The P-12 (Lo Lo Tave) interlock is an ESF actuation which blocks Steam Dump valves actuation to protect against excessive plant cooldown below the minimum temperature for criticality. Its setpoint is 553 degrees F on any two out of four NC System loops. If the logic is satisfied, then the Steam Dump valves are failed closed until NC System temperature is above 553 degrees F.
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0l2l 5 - 0l 0 olg or o[g nxi w,-. . w,w. mc r asuw nn i During this incident, the Steam Dump Valve System and the p-12 Interlock responded properly and as expected to the signals generated. MES Engineers, {1 performing the Incore Thermocouple and RTD Cross Calibration test at the 557 k! degrees F temperature plateau, inadvertently allowed-low (547 degrees F) RTD Simulator signals to be insected into the 7300 C/R Process Cabinets. Due to these low signals, the Steam Dump valves closed due to the P-12 interlock
- h. permissive being met. With the valves closed to prevent excessive plant cooldown, actual NC System temperature and pressure increased during the 15 minute test duration. When MES Engineers returned the NC System RTDs to service, they indicated the actual NC System temperature of 560 degrees F, which allowed the p-12 interlock to reset and along with a steam pressure demand sensed by the Steam Dump Valves controller, caused the Steam Dump valves to open. The Steam Dump valves rapidly brought NC System temperature down to approximately S47 degrees F which again caused a P-12 actuation as the setpoint of 553 degrees F was reached. In addition, a shrink in Pressurizer level and NV System letdown isolation occurred. Subsequent action taken by Control. Room.
Operators restored NV System letdown and Pressurizer level. In addition, the Steam Dumps reopened as NC System temperature increased above 553 degrees F. The transient experienced during this incident is bounded by FSAR analyses. The heatup portion of the transient is bounded by the Inadvertent Closure of a Main Steam Isolation Valve in FSAR Section 15.1.4. The total NC temperature increase was approximately 3 degrees F. Pressurizer PORVs and Safety Valves were not
. challenged.
The-cooldown portion of the transient is bounded by the Inadvertent' opening of a Steam Generator Relief or Safety Valve described in FSAR Section 15.1.4. The
' total'NC temperature decrease was approximately 13 degrees F. Pressurizer level decreased below 17%, resulting in NC letdown isolation as expected. Neither the steamline pressure nor the NC pressure transient resulted in Safety Injection Actuation signals.
The combined heatup/cooldown transient is adequately accounted for in the transient cycle analysis, g , The health and safety of the public were not affected by this incident. L l l l
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REFERENCES- '4 .. p 93
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- Di MP/1/A/3231/01,11ncore Thermocouple _ and RTD Cross Calibration ' N-q clp6 f . ..
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- CNS Alarm Summarys Reportifor .4/9/90 - ' '
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, SAFETY REVIEW GROUP CONSIDERATION-OF PART 210REPORTABILITY
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- Yes- No.
7 T.- Deficiency of a plant security system? . _x_ . p,- '1..'Could defeet' create.a. substantial: safety hazard? _xi .
- l;IfLyesi-provide thisLinformation/to Compliance SectionLimmediately for determination of Part.
, 121.reportability.
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.II. Is ! the, plant structure.-' system, component or- part .
.b thereof'necessary'to' ensure:
4 1. 'The integrity of the' reactor-coolant boundary?' _xi . ~ ' 4 -. .
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'2._ The. capability-to shutdown reactor and maintain', s it in al safe ' shutdown condition? _x_.
,[ [3,D.The capability -to prevent ~ or mitigate the-b'? lconsequencesinf accidents,which coulduresult 4
' :.in pot.cntial' of f-site' ' exposure ~ comparable to?
' ' . Lthose referred to' in 10CFR100,117 _x; III.- Is thel defect in a basic' component one that has been P
z accepted J for ownership? _x
-Installed for use or. operation? '
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,c a" > j1fla yes in'II--and J IIILabove, could defect create a & , substantial = safety. hazard? _x_
m If1yes, provide this information to Compliance'Section 3 -immediately for determination of Part 21 reportability. 3 J: LIV..jI's!efectin:abasiccomponent.: d
- A- condition that could contribute to exceeding'
'; 'of a safety limit'as: defined in the Technical Spec.ifications?. .
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"If ;yes in II and IV above, . provide this ' information b I to: Compliance section immediately for-determination ~of Li
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.l ; Prepared By: ,'/// Date: f 9 0 Revieweil By: Date: 5690 3 p J V- Comp 1innee l p, .g
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i ! e i F ENCLOSURE 8.7 Safety Review Group Signatures i Prepared By: Derrick D. Miller Dates- May 2. 1990 ,1 Reviewed By: Md _ Date: f. 93 __ _k ! Date: S %~80
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Date 6 70 Date: N- 2 -9 D-
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( .i [ M U. b Date: 5 9 o - I E Approved By: Date: (~2-90 Chairman. CSRG , Distribution: CSRG Files L.F. Firebaugh, ODA Support Engineer R.C. Cole, OEP Coordinator P.M. Abraham, Nuclear Safety Analysia
- 4. W.T. Orders, NRC Resident Inspector L.R. Davison, QA Tech Services F.G. Hudson,. Design Engineering Master File: 832.04
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- PERSONNEL REFERENCED' ,
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~ LAllen' Smith; 1NES/Srt Tech Specialist-D (. ;TommyLRobinson) MES/Sr.-Tech Specialist. -
l OLi L[iM M[o 3Andy Gooch iMES/ Associate Engineer 4 j k" 6' LBrack? East l 'MES/ Engineering Supervisor-- m 4 h T**Gail: Ice OPS /Shitt-Supervisor
$g#Sf; ;p ? i ?*Jeff Bradley- iOPS/NC0a
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x j F s!.n , : Stan' Marion : -OPS / Tech Specialist'III- M PGn <
;BoblFergusois , OPS /ShifL'Operattons: Manager m '
Jf [ SJoefMarlow! IAE/ Specialist-
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o o i 4 l ro6sas72(ms.ao) - 4^ COMPLETE FORM BY PRINTING WITH BLACK BALL POINT PEN OR TYPE DUKE POWER COMPANY [@a NUCLEAR STATION Problem investigation Report Serial No. ) ~ Cf8 -Oll9 i Station . NAW Licensee Event Report No. 4/. f /96-J5'
//A? C90-03JV-/
- l. Problem Occurred Time /Date: N ' T* 9' Discovered-Time /Date: 0$Y7! '/ 9 40 Unit (s): I I Unit Status At Time Problem Occurred /Disgovered: U Description and Cause of Problem: b- o arm mod detery/im n n.
Other Duke Stations Affected O Yes 0/No Determined By/Date: M hv "* Comments: M ,/ ~ Location of Problem: C#5 /Aad / Method Used to Discover Problem: Immediate Corrective Actions Taken/To Be Taken: Ad Work Stoppage Notification (Form OCK 2A) Serial No.: 4 Information Sources /fleferences (Work Requests, Documents Violated, etc.): to.heb : k JL % f7)'4 0 Ga. / Tce , C4=_- Cet .at) U/*M/**//* N
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3, Originated By: 40 M Date: @i D Dept./ Group /Section: A 'l
.(I'i. Compliance Eval (ation Itemp)y,s, tertApplicable
/JO erable IYes u O No O Not l Evaluated By/Date: VW M*~ d'f W Comments !
Reportable EfYes O Nof' Reportabl'e Per: D 50.73 Section _ 't W:id s O'50.72 Section - (4R*A'l O 73.71 Section O T.SLic Corg Section AM O Part2PII6ffier: M O Part 50.9 Evaluated By/Date: M Me W ~f# Comments: N hM/W J
//
111. Telecon/ ENS Report to NRC me/Date: __ 42 d 6 1 fo NRC Contactee(s): DPC Contactor(s): 6 lh *- Telegraph /Mailgram/ Facsimile Trafismission to NRC-Date: A/4 Date Notified: NRC Res. Inspector: A* Station Manager: 4/4 l . General Office: # . Comments: # L IV. Investigation Assigned To: CMG NRC Report Due Date: 5'- T *f o Date Due to Compliance ef ter Evaluation: r. t. -f a
~
PIR Review (Compliance): A% 0% >-- Date: 8-8oo PIR Station Manager Approval: Rhm I4T Date:M C50_ f V. Further Action / Evaluation Required O Yes O No(Explain Below): a 1 Page 2 Assigned To: l Comments: Compliance Review: Date: OA Review: Date: -
"'"tribution tm nenalnu enc vrn M-7 Initial Originat6r 3u c1nnn nu c1nunr cgan nc Mnvenn up smith 3c Torrn l :- JM Knuti JR Fereuse, RN Casler E LaCasse TE Crawford WR McCollun JS Forben TB Owen Final Onginator
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- flA l-CfD-Olli A AM i I JkseY IfI l Event Description i O IAE. Technicians were conducting RTD thermocouple cross i
calibration testing while the plant was maintaining 557' F i l on the Steam Dumps. The test inserted a dummy signal into l the Tave circuitry. The dummy signal was set for 557' F but. 1 was actually approximately 10* F' low. This was a test i equipment. problem now being investigated. This caused P-12 to actuate and isolate the steam dumps. ll Steam generator 1 pressure and temperature slowly increased during the approximately 15 minute test duration. At the end of the test the actual temperature signal was re-instated and the steam dumps re-opened seeing an actual temperature of 560' F. The dump valves rapidly brought temperature back down to (0 547' F.again causing P-12 actuation as the setpoint of 553* F was reached. .The total temperature swing of the NC System was between 560.5* and 547* F. l I. l . . I. 1 YYu l 1}}