ML20058C772
| ML20058C772 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/04/1990 |
| From: | Kovach T COMMONWEALTH EDISON CO. |
| To: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 9011050073 | |
| Download: ML20058C772 (18) | |
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} Commonwealth Edison
{h 1400 opus Place L/
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Down;rs orove, imnois 60515 September 4,1990 Mr. A. Bert Davis
. Regional Administrator U.S. Nuclear Regulatory Commission Region lll 799 Roosevelt Road Glen Ellyn,IL 60137
Subject:
LaSalle County Station Units 1 and 2 Response to inspectin Report Nos.
50 373/90014 and 50 374/90015 NR0_Doche.tNoLED-3Z3_and_50 314_
Reference:
(a) E.G. Greenman Letter to C. Reed dated June 28,1990.
Dear Mr. Davis:
This letter is in response to the inspection conducted by Messrs. T. Tongue 7
and R. Kopriva on June 3 through July 17,1990 of certain activities at LaSalle County Station. Reference (a) Indicated that although no violations of NRC requirements were identified, a written response was requested to address five questions concerning the unanticipated cooldown of Unit 2. This was classified as an unresolved item. The Commonwealth Edison Company's response to the unresolved item is provided in the following attachments, if you have any questions regarding this matter, please contact this office.
l Very truly yours,
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T.J.
vach Nuclear Licensing Manager Attachments cc: T. Tongue - Senior Resident inspector - LSCS R. Pulsifer Project Manager - NRR
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e ATTACHMENT A
.UntewlyedJtettL3Z3190014 Oland.2Z4/20015-02 Question #1:
What procedures were being used at the time of this event and were they adequate?
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Response
The procedure being utilized by Operations personnel as the base procedure for shutting down the unit was LaSalle General Oarating Procedure LGP 2 2,
- Unit Shutdown from Power Operation to Hot Standby." Numerous other procedures were being utilized for the operation of plant equipment and the performance of requiro# surveillances. Operation of equipment and performance of surveillances is 1
outilned in LGP 2 2. At the time the reactivity addition occurred, the operating crew entered LaSalle Op"erational Abnormal procedurs L OA NB 07,
- inadvertent Reactivity Addition. Operation of the IRMs dunn the shutdown and the reactivity addition was in accordance with LaSalle Operati Procedure LOP NR 02, *IRM Operation." LaSalle Administrative Procedure LA 100 35,' Reactivity Management Controls,"is in effect during all modes of operation, particularly during maior evolutions involving reactivity changes.
A review of the procedures involved during the reactivity addition showed that the operating crew acted in accordance with approved procedures. The pu' pose of LGP 2 2 is to provide a method of bringing the reactor to a HOT STANDBY condition.
Step F.60.c of LGP 2 2 states:
if decay heat load is reduced to a level such that Reactor Pressure cannot be maintained at the desired value. ADD heat by withdrawing control rods in the prescribed sequence to bring the Reactor critical (if initially subcritical) and INCREASE powe' to the heating range (Range 7-8 on MM's).
This implies that it may be necessary to increase power in order to maintain the reactor in the heating range and, therefore, maintain reactor pressure at the desired value. LOP-NR-02 Instructs the operator to maintain the IRM's onscale during a shutdown In order that reactor power can be adequately monitored. Applicable steps of LOA NB-07 were periormed satisfactorily as time permitted.
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2 A thorough review conducted by station (Attachment B) and corporate personnel (Attachment C) has determined that the root cause of the event was inadequate procedural guidance for handling the cooldown effects during the reactor shutdown.
Specifically, LGP 2 2 does not provide a step which requires the operating crew to obtain a calculation of the expected decay heat based on the reactor power history.
The procedure should provide guidance for matching the steam loads to the estimated decay heat and aroperly controlling the pressure during reactor shutdown. Had the operat ng crew been provided guidance for balancing the steam loads, the depressurization, and hence, the reactivity addition, may not have taken place.
Another procedural deficiency was identified with regards to identifying the effects of the cooldown. Although it may not have prevented the reactivity addition from occurring, a precaution in LGP 2 2, reminding the operating crew of the positive reactivity addition that may occur from depressurization of the reactor, may have enabled the crew to predict that the power increase was going to occur.
Other procedures involved in this event were reviewed and found to be adequate.
Guidance to resolve the deficiencies identified in LGP-2-2 will be incorporated into the procedure. Consistent guidance will also be incorporated Inio LGP 21, " Normal Unit Shutdown," as applicable.
Question #2:
Why did the operator upscale the IRMs instead of inserting a manual scram or driving rods in when he saw the positive reactivity addition?
Response
The operator immediately verified that no other personnel were doing anything that would cause an unexpected reactivity addition. He then upranged the IRM's in accordance with LOP NR 02 in order to provide adequate nuclear instrumentation to monitor reacPar power. The Shift Engineer immediately reviewed reactor depressurization and cooldown trends on tho appl %able chart recorders. Within seconds of the initiation of the event, the Shm Engineer concluded that the reactivity addition was due to the cooldown. The Shift Eng neer then Instructed the operator not to insert any control rods because he believed this would only make the reactor depressurization worse. He believed that by allowing power to increase to the point of adding heat, the reactor cooldown would be terminatG and the positive reactivity addition would be negated.
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Subsequent review of the event by LaSalle and Corporate Nuclear Engineers (Attachment C) determined that the worth of the control rods that would have been allowed to be inserted in accordance with the Control Rod Sequence arobably would not have overcome the positive reactivity addition. This is due to the < act that the next rods to be inserted were peripheral rod moves from notch position 04 to position 00. In addition, insertion of control rods would have provided another source of distraction to the event that was occurring.
A scram was not required at this point based on guidance provided in LAP 100 35 and training the operating crew has received on Reactivity Management, i
Operations personnel are required to question reactivity effects and to take conservative action (including a manual scram) if reactivity effects cannot be explained. The operating crew had immediately identified the source of the reactivity addition and bWeved that they wer0 taking conservative actions to stop the reactivity addition. Had the opereting crew beon wrong about the source of the prior to reaching any Limiting Safety System Setpoint (LSSS).y scram the r reactivity addition, contingency plans were in place to manuall The rate of the power increase being experienced was actually very controllable (100 second period). Similar reactor periods are experienced during a normal reactor startup and controlled without difficulty, During the event, three half scrams occu* red due to IRM trips. One of these was due to a noise spike in the IRM drawer m.iciated with the change in power supplies when Ing from range 6 to range 7.
The second IRM trip was caused by a cognitive error the NSO who turned the wrong range switch. The NSO ranged up IRM channe G Instead of IRM channel E.
The other trip was due to neutron level actually reaching the high trip setpoint because the NSO had became distracted while checking reactor pressure and temperature changes.
The NSO and his management were able to determine and correct the cause and reset the half scrams without difficulty as they occurred.
Question #3:
What information was provided to the crew regarding the uniqueness of this evolution (e.g., lack of decay heat) and was the information adequate?
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Response
As stated in the response to Question 1, inadequate procedural guidance relative to this unique evaluation was identified as the root cause of the event. The information provided to the operating crew was insufficient to alert them to the f act that there l
would be a small amount of decay heat. The procedures involved provided mechanisms which could be used to remove various amounts of decay heat, up to the maximum decay heat loads. No information was provided to assist in selecting appropriate mechanisms for less than design heat load conditions. The operating crew involved was aware tnat they would need to accommodate the decay heat loads. On a previous shutdown on LaSalle Unit 1, the Station Control Room Engineer (SCRE) on this particular operatina crew had been involved in preparing to handle the decay heat loads that were in ex stence during that shutdown. Based on his previous experience, the SCRE emphasized to the Shift Engineer the importance of being prepared for a possible pressurization if they did not have enough steam draw off from the reactor when the MSIV's were closed. The Shift Engineer had conducted a briefing with the personnel involved to ensure that everyone knew ahead of time what the plan of action would be to control reactor pressure.
However, the crew did not recognize the f act that Unit 2 would not have nearly as much decay heat due to the fact that the unit had just started up from a refueling outage and had been at reduced power (25%) for a day prior to starting the shutdown.
LaSalle Station general procedures for Unit Shutdown will be revised to include guidance for calculating core decay heat based on the reactor power history and balancing steam loads properly to control pressure during reactor shutdowns.
Precautions will also be added to warn the operator of the potential to add positive reactivity during reactor cooldowns. Completion of the procedure revisions is expected by September 15,1990.
o Question #4:
What is your basis for not reporting this event within one hour as required by 10 CFR 50.727
.iesponse:
The notification requirement given in 10 CFR 50.72(b)(ll) is stated as follows:
Any event or condition During Operation that results in the condition o' M nucioar aower plant, including its principal safety barriers, being in a conditioikot covered ay the Plant's Operating and Emergency procedures.
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5 A review of the event shows that at no time was the plant operating in a condition not currently covered by both operating and emergency procedures or outside the design basis of the plant or any of its systems. LGF 2 2 provided !nstruction to control reactor pressure between 600 - 900 psig prior to closing the MISV's. The intent of this instruction was to prevent any chhges to reactor pressure from affecting reactivity. What LGP 2 2 fitiled to provide was a warning to the operating personnel of the affects that a reduced decay heat load would have on pressure i
control. The reduced decay heat load resulted in an unexpected pressure decrease that led to the reactivity adcition. LGP 2 2 discusses actions to be taken if the reactor pressure cannot be maintained at the desired value. These actions include withdrawing control rods such that the reactor reaches the point of adding heat (IRM ranges 7 to 8). Although it does not saecifically discuss upranging the IRM's, LOP NR 02 does require that the IRN's be maintained on scale for the purposes of i
monitoring reactor power level. By upranging the IRM's as power increased, the j
operator was not bypasting a LSSS. Had that been the intention, the operator would have immediately upranged the IRM's to range 10. The operator was utilizing the IRM's in accordance with their design function (i.e., upranging to maintain s
l readings on scale). Had the reactor power increase been too rapid, the operator would not have been able to uprange the IRM's fast enough to 3revent a reactor j
scram. This is in accordance with the design function of the IRM's to ensure a scram will occur if the reactor aeriod becomes too short for the operator to effectively monitor power leve. The APRM 15% scram in the STARTUP mode of i
operation provides the design basis power level scram to prevent fuel failure. The operating crew had already decided that they would manually scram the reactor if e
the APRM downscales (3%) had cleared, which is well below the 15% LSSS.
In addition, in accordance with guidance p)rovided in the Federal Re Ister N i
to events that significantly compromise plant safety. This is intende for significant events where the condition of the plant is seriously degraded. The safety consequences of this power increase event were minimal because the reactivity l
Insertion wac small (i.e. only 0.06% Ak) and no fuel design or safety limits were approached.
Question #5:
1 What were the root causes identified and the corrective actions taken with re ards to this event (e.g., was a nuclear engineer sufficiently involved; is operator tralnfng l
adequate)?
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Response
The root cause of this event as discussed in the response to Question 1 was a lack of guidance provided to the operating crew with ressect to the amount of decay heat that would be present and how they should handle che cooldown effects. A Nuclear Engineer in training was present in the control room and acting as the second venfier for control rod moves at the time of the event. This incividual was not sufficiently qualified to provide guidance to the shift in rcactivity management and l
decay heat consideration in addition, information was not provided in the procedure discussing the effects that may be seen if the cooldown is too rapid, speciffcally the positive reactivity addition.
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The immediate corrective actions taken as a result of this event are discussed in detallin Section E of Attachment B. These actions included stoppage of all activities that ws.re causing the reactor depressurization and contacting a Qualified Nuclear Engineer for guidance as to how the shutdown should proceed to ensure that a similar event would not occur.
Long term corrective actions were also discussed in Section E of Attachment B.
These actions included a review of the event by corporate personnel the following day, a complete review of the reactivity aspects of the event by CECO's Nuclear Fuel Services personnel, review and revision 01 apalicable plant procedures, and Operating and Nuclear Engineering personne training on the event including a simulator drill similar to the event.
The procedure revisions are being reviewed at this time. In addition to the aforementioned procedure revisions LOP 21 and LGP 2 2 will require that a Qualifisxf Nuclear Engineer provides input to Operations personnel during the shutdown process to ensure that special situations are addressed. Completion of the necessary procedure revisions is being tracked by LaSalle Station's Commitment Tracking System. Completion of the revisions is expected by September 15,1990.
The Operator Requalification simulator training has been modified to include a drill similar to this event. Licensed personnel are in the process of going through this drill. Nuclear Engineers also participate in these simulator drills Complet on of this training is expected on September ' 9,1990.
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ATTACHMENT B LaSalle County Station Unit 2 Deviation Report 0102 90-052,
- Inadvertent Reactivity Addition" l
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DEV1A110N REPORT f
DVR NO.
90 - 052 01 - 02 STA UW11 YFAR W3.
Forn Rev 2.0 PART I l Il1LE OF DEVIATION OCCURRED jg Inadvertent Reactivity Adriltion DATE TIME
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$Y$1EM AFFECTED PLANT $1ATUS AT TIME OF EVENT TESTING
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NM Mu0E 2
POWER (%)
0.04 WORKREQUESTN50 l
l l X l
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DESCRIPil0N OF EVENT During the performance of LGP 2-2 at Step F.51, a reactivity addition resulted in IRM's increasing from approximately Range 4 to approximately Range 7.
Closing of the M51V's was secured, control rod insertlon was secured, and depressurization of the RPV decreased until power increase stabilized. LOAJB-07 was consulted and under the direction of a QNE, control rod insertions recomenced.
POTENilALLY $1GNIFICANT EVENT PER NSD DIRECilVE A-07 YES NO g,
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10CFR50.72 NRC RED PHONE l
l M. A. Peters 6/24/90 g
.NOTIFICA110N MADE l
l l-X,,,l 11ME RESPONSIBLE SUPERVISOR DATE PART2l.OPERAllNGENGINEER'SCOMMENi$
Increase in reactivity was promptly identified by the NSO assigned to the 603 panel for CR0/ Neutron Monitoring. IRM's were ranged up in a controlled manner. The reason for the reactivity increase was quickly identified - temperature decrease in Rx coolant due to decreasing pressure. Reactor power was allow to increase to the heatup range, at which time the negative reactivity from cy(moderator ter('erature) caused power to stabilite., Quallfled Nuclear Engineer was contacted for reconnendations on inserting control rods prior to continuing closing M51V's. 005 was notified cf the reactivity / power change. NRC Resident was notified of event. Station Nuclear Duty Officer was notified.
Rx shutdown was continued with no further incident.
g, g NON RLPORTABLE EVENT
- i NOTIFICATION REGION 111 DATE TIME g
g;30DAYREPORTABLE/10CFR g
g.5DAYREPORTPER10CFR21 NSD DATE TIME g
{ ANNUAL /SPECIALREPORTREQUIRED CECO CORPORATE NOTIFICATION MADE A
N0ilHCAll0N n M M21 A.I.R.' #374-200-90-0$20) to 05204 TELECOPY,
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CFC0 CORPORATE OFF1tER DATE TIME PRELIMINARY REPORT
. COMPLETED AND REVIEWED W. Kirchhotf 6/25/90 OPERAllWG ENGINEER DATE a
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INVE$ilGATION REPORT & RESOLUTION L ebb ll '.N s
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ACCEPTED BY $1A110N REVIEW 7-EA4e n_ e (r
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RE50LU110N APPROVED AND
.(CA MI AUTHORIZED FOR DISTRIBUTION veJk 8fo p hM NAGER DATE 96-5176 11-89' (fort 5-52-1)
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! DOCUMENT 10 0760p l
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DEvlA110N INVL511GA110N REPORT TEXT CONilNUATION Fom Rev 2.0 FACIL!TY NAME DlR NUMBFR PAGE SEQULNilAL REVISION STA UN11 YEAR NUMBER NUMB (R 0l 0 2
0F 011
.a$alle County Station Unit 2 Ol I 01 2 91 0 015l2 TEXT Energy Industry identification System (Ell 5) codes are identified in the text as (XX) 8.
DESCRIPil0N OF EVENT (Ce9 ' Md)
Listed below is a chrono. 3 s description of the event:
Initially Unit 2 was operating in the Run Mode at 24.3% reactor power. ControlRods(RD)(AA)were being inserted to lower the Main Generator (TC) (TB) output from 192 Mwe to 60 Mwe in preparation for taking the Mein Generator off-line in accordance with LGP 2-2, " Unit shutdown From Power Operation to Hot $tandby ' This was a forced unit outage to facilitate repairs for the Unit 2 Main Turbine (TG)
(TA), 2A furbine Driven Reactor Feed Pug (FW) ($J) and the 28 Reactor Recirculation Hydraulic Power Unit (RR)(AD). The LGP 2-2 was being used for the Unit 2 shutdown because it provided direction for closing the Main Steam isolation Valves (M51V) (SB) during shutdown evolutions.
At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> following a preshift briefing, the Shift Engineer ($E, licensed senior reactor operator),
Shift Control Rocan Engineer ($CRE, licensed senior reactor operator), Senior Manager $RO on Shift (licensed senior reactor operator), and Wuclear Station Operators (NSO, licensed reactor operator) had a briefing in the Maln Control Room to discuss decay heat and pressure control during the shutdown J
I Cv31utions, it was discussed that the Main $ team Line (MSL) ($9] drains would not handle the decay heat load alone, therefore it would be necessary to have the Reactor Core Isclation Cooling (RCIC) (BN) on line prior to closing the M51V's. Once the preshift briefing was completed, Unit 2 reactor shutdown was centinued by inserting control rods at step 65 of the rod sequence package.
At 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />, LaSalle Operating Survelllance LOS 4tS-SR2, "R$CS Rod Block Operability Test" was completed. The crew continued to insert control rods to 100 Mwe and then performed tasalle Operating i
surveillance LOS-TG-5R3, " Turbine leeekly Canbined Internediate valve Surveillance Less Than 50 Percent Power" to troubleshoot the Unit 2 Turbine Electro Hydraulle Control (EHC) (TG) system. Once this
'LCsting was completed control rod insertion continued.
At 1725 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.563625e-4 months <br />, the Main Generator output was reduced to 60 Mwe by inserting control rods per LGP 2-2.
- The Main Get,erator was then taken off-line and additional control rods were inserted to continue the reactor shutdown.
At 1754 hours0.0203 days <br />0.487 hours <br />0.0029 weeks <br />6.67397e-4 months <br />, the Unit 2 reactor mode switch was placed in the Startup/ Hot Standby position and LaSalle Operating Surveillance LOS-NR-Id2, "lRM Detector Not Full in Rod Block Functional Test" was completed.
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'At 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, the operating crew placed 2A Residual Heat Removal (RHR) (BO) in the Suppression Pool Cooling mode of operation in preparation for starting the RCic system for reactor pressure control.
' At 1906 hours0.0221 days <br />0.529 hours <br />0.00315 weeks <br />7.25233e-4 months <br />, control rod insertion was cogleted up to step 36 of the rod sequence package, j
li At'1915 hours0.0222 days <br />0.532 hours <br />0.00317 weeks <br />7.286575e-4 months <br />, the Off-Gas Mechanical vacuum pum was started and steam was isolated to the Off-Gas l'
SteamJetAirEjectorstoreducesteamdemandonthereactorandtoallowRCICtocontrolreactor l'
pressure when placed on line. The Main Steam Line drains were open, the Gland Seal Steam system was in
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operation and the Mala turbine Stop Valve above seat drains were open at this point.
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- c DEVIA110N INVL$11GA110N REPORT TLXi CONilNUAll0N Fom Rev 2.0 fAClllIY NAME DlR NUMBER PAGE SEQUENilAL REVISION
$TA UNIT YEAR NUMBER NUMBER 01 0 OF 016 0l5l2 LaSalle County $tation Unit 2 01 1 01 2 91 0 TEXT Energy Industry identification System (Ell $) codes are identified in the text as (XX)
C.
APPARENT CAUSE OF EVENT The change in reactor power occurred because that reactor was not in an equilibrium condition where reactivity feedback effetts could conpensate for the cooldown in progress. After the reactor power (flux) drops below the heating point, tLe neutron flux is driven by the balance between moderator twperature and control rod density. Prior to the observed IRM response, positive reactivity was being added by the reactor cooldown. No increase in IRM reading was seen at that time primarily because the rod insertions were adding approximately the same (or slightly less) negative reactivity.
When the NSO stopped the rod moves to watch for effects of the MSIV closures, the negative reactivity insertion (from control rods) stopped but cooldown (positive reactivity insertion) continued. This allowed neutron flux to increase when the reactor cooled down. The flux increased exponentially (as in a nomal startup) until the reactor power (flux) reached the point of heating the coolant. After this point was reached, themal equilibrium was achieved and the neutron flux was observed to stabilire as expected.
The primary cause of this event was due to inadequate procedural guidance to operating personnel for handling the cooldown effects during part of the reactor shutdown sequence. The current shutdown procedures do not require the calculation to be perfomed to detemine decay heat load based on previous -
power history. There was lower than usual decay heat due to the recent refuel outage, startup, and extended hold at 25% power during the shutdown conpared to previous unit shutdowns where more decay heat
[
was experlenced.- If the operating crew had been able to detemine the decay heat load contribution to steam production during shddown (when power has been reduced to less than the point of adding heat),
this may have allowed the crew to better detemine what methods could be utilized to better control reactor pressure. The shutdown procedure currently does not list the steam removal capacities of various systems / components which may be on-line during shutdown evolutions.
Another cause of this event is that this is an infrequently perfonned task. A unit shutdown behind the Main Steam Isolation Valves is only perfomed on rare occasions. The operating crew did not have adequate training to' allow them to detemine t. hat the reactor pressure response would have been for the condition that they were in. The operating crew was expecting reactor pressure to increase as the Msiv's w(re being closed.
A contributing factor was that the procedure did not provide adequate coordination between control rod insertion and reactor pressure control. The number of control rods inserted reduced reactor power to less than the point of adding heat without an adequate means available to stabilire reactor pressure to
. prevent further cooldown. This contributed to the reactor power increase while reactor pressure was decreasing.
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5AFETY ANALYSIS OF EVENT The safety consequences of the event were minimal. All neutron monitoring channels were functional and a trip would have occurred if needed from either IRM or APRM channels. Both the $CRE and the N50 had mentally established criteria for manually tripping the reactor prior to the automatic safety setpoint (15% APRM flux). This action would have teminated the event if the actions to let the transient settle out had not resulted in the expected response.
r
e DEVIA110N INVE$ilG4 TION REPORT TEXT CONilNUATION l
Fom Rev 2.0 F ACILITY NAME DlR NUMBER PAGE SEQUENTIAL REV1510N
$TA UNIT YEAR NUMBER NUMBER j
01 0 6
0F 016 tasalle County station Unit 2 01 I 01 2 91 0 0l512
, TEXT Energy Industry identification Systen (Ells) codes are identifled in the text as (XX) l E.
CORRECTIVE ACil0NS After the reactor was stabilized, the SE, SCRE and Unit NSO discussed the situation and determined that if they saw power increase again, they would insert a manual scram.
The SE contacted the Quallfled Nuclear Engineer who provided guidance for continued rod insertion to prevent power from turning again during shutdown evolutions.
The operating crew reviewed LaSalle Abnonnel procedure LOA NS 47, " Inadvertent Reactivity Addition" to j
determine if any additional action was needed. No additional actions were required.
A Henan Performance Enhancement System (HPES) analysis was completed on this event and the findings were incorporated in this report.
A corporate review of this event was performed, in addition to this review Nuclear Fuel Services will provide the station with a written report. The conpletion of this corrective action will be tracked by Action item Record (AIR) number 374-200 90-05201.
A review of industry events relating to this event was perfonned to review corrective actions taken by other utilities for implementation at LaSalle. A similar event was documented at Limerick as LER
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352/88-012 which occurred on April 9, 1988. This LER has been reviewed to detennine if acditional corrective actions can be inplemented at LaSalle. No additional corrective actions were identified.
A review of thic event will be perfonned by the technical staff to detennine what procedural guidance can be added to the existing LaSalle General Procedures for Unit shutdown. This guidance will be provided by a procedure change. The cunpletion of this corrective action will be tracked by AIR nunber 374-200-90-05202.
A tallgate of this event will be provided to all licensed operating personnel, emphasiglng reactivity management. The etspletlen of this corrective action will be tracked by AIR nunber 374-200-90-05203.
I This event will be reviewed by training, to determine what changes or additions to training are needed.
The conpletion of this corrective action will be tracked by AIR nunber 374-200-90-05204, f.
PREVIOUS EVENTS None.
G.
COMPONENT FAILURE DATA Not Appilcable.
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x-ATTACHMENT C Nuclear Fuel Services' Evaluation of LaSalle County Station Unit 2 Inadvertent Reactivity Addition Event During Unit Shutdown
August 3, 1990 NFS:BSS:90-096 Mr. G. J. Diederich
Subject:
LaSalle Unit 2 Power Increase during Unit Shutdown
References:
- 1..
Letter, G. J. Diederich to T..Maiman, D. Galle, et. al.,
'Potentially Significant Event Report on Unit 2 Power Increase during Unit Shutdown in accordance with LGP-2-2, Unit Shutdown f, rom Power Operation to Hot Standby"', June 26, 1990.
2.
LaSalle Unit 2 Deviation Report No. 052, ' Inadvertent Reactivity Addition', June 26. 1990.
During-the shutdown of LaSalle Unit 2 on June 25, 1990, an unexpected.
power increase was expertenced while preparations were being made to close the MSIVs. Details of the event may be found in References 1 and 2.
NFS was requested by LaSalle to evaluate the core power response.
This review has been completed with the following conclusions:
The rapid cooldown and depressurization of the reactor was a result of the mismatch between decay heat and steam loads.
l The power increase was caused by the positive reactivity associated with the decrease in, moderator temperature, The reactor response was normal'given the changes in reactivity.
o The operating shift properly diagnosed the-cause of the power increase, and acted in a conservative manner in ac':ordance with l
Edison's' reactivity management philosophy.
l-Because of the low power levels and the sm:ll amount of reactivity addition, the safety consequences were minimal, t
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A detailed description of NFS's evaluation of the event may be found in the attachment and includes recommendations to prevent recurrence, t
If there are any comments or questions concerning this letter please call Mike Salva at extension 8754.
William F. Naughton Nuclear Fuel Services Nanager WFN:FMS:bl ID:ZBXL:11:38 i
I Attachment cc:
D. P. Galle/N. J. Kalivianakis J. Miller /E A. McVey B. B. Palagi/K. B. Ramsden /R. C. Chin K. N. Kovar/J. E. Ballard J. M. Dotter /M. E. Hagner File: L2C4 NFS-CF l
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Attachment I
Discussion:
NFS has detemined that the rapid reactor cooldown was a result of the mismatch between the steam load being produced by decay heat and the steam load being removed from the reactor, primarily by RCIC. Because Unit 2 recently resumed power operation following refuelinf the decay heat production was significantly lower than normal.
Using tb LaSalle Unit 2 power history data supplied by the station, NFS detemined that the decay heat on 6/23/90 at 19:33 was approximately 6.12 MWth.
If it is assumed that 1.5 MWth of the 6.12 MWth of decay heat is dissipated thru radiation from the vessel and rectre piping to the drywell, then 4.62 MWth of decay heat is required to be removed by decay heat cooling systems. The steam load being produced by the 4.62 MWth of decay heat is approximately 13,172 lba/hr. LaSalle also indicated that when RCIC was placed on line the bypass valve closed from,37 to 35% open. Therefore, an approximate estimate of the steam load being removed by RCIC is 2% of the steam flow thru one bypass valve. This is calculated to be approximately 17,875 lbs/hr (LaSalle has 5 bypass valves with a total bypass capacity of 25% rated steam flow, i.e. each bypass valve has 55 bypass capacity).
the above calculation shows that the heat removal from the reactor was gre,li Therefore, with all of the bypass valves closed and only RCIC on ater than the heat production. This led to a rapid cooldown of the reactor and eventually led to a positive reactivity insertion when control rod insertius ceased for closure of the MSIVs.
After a thorough review of reactor data, NFS has detemined that the reactivity insertion that led to the power increase can be accounted for by the rapid temperature decrease and the negative Moderator Temperature Coefficient (MTC). Neutron instrumentation indicated the reactor was critical after the rod insertions ceased, prior to closing the MSIVs. A reactivity insertion of only 0.06% Ak/k would have placed the reactor on a period of approximately 100 seconds.
NFS' discussions with GE regarding the value of the MTC at a temperature of 500 degrees Fahrenheit yielded an approximate value i
of. -6.0 -- -7.0E-05 a k/ AT ( F).
Therefore, the 15 degree Fahrenheit moderator cooldown which occurred from when the operator completed control rod insertion step #27 until the power increase event occurred, led to a positive MTC. reactivity. insertion of between 0.09 - 0.105% 6k/k. NFS used the cooldown from control rod insertion step f27 instead of from step #26 because step #26, a peripheral group, was detemined to be of low incremental worth (approximately 0.0015% Ak/k).
Furthermore, NFS believes that the power increase was not caused by a void collapse from the closure of the two MSIVs or from the RCIC turbine runback because:
The core void content and steaming rates were extremely low, and, No pressure increase was seen on the plant pressure recorJer, and 1
~. _ _. _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
only 2 of the 4 MS!Vs were closed and the rear.ining two MSIVs could easily accommodate any steam loads off the equalizing header.
The reactor flux level was below the point of adding nuclear heat when the power increase started, therefore, the flux increased until heating was reestablished and enough negative reactivity was inserted to terminate the flux increase and stabilize the reactor at a higher power level. NFS has also detemined that control rods in the next sequence step were peripheral rods with insufficient worth to termintte the event.
NFS also has reviewed the L2C4 core design and believes that the core resptmded as designed. The presence of GEg8 fuel in the core did not cause or aggravate the power increase event. GE has indicated that for a given lattice type, de MTC for the GE98 fuel design is the same as the MTC for the GE8 fuel i
design. especially considering the high control rod densities which existed at the time.
NFS believes that the operating shift acted conservatively in accordance with the reactivity management guidelines:
- The shift promptly recognized the power increase and properly diagnosed it as due to the rapid cooldown, and i
- The operating shift was prepared and had established criteria to manually scram the reactor (i.e. if the 35 APRM down:cale was cleared) prior to the neutron flux level clearing the automatic safety setpoint (15% APRM flux).
The safety consequences of this power increase event were minimal i
because the reactivity insertion was small (i.e. only 0.065ak) and no fuel design or safety limits were approached.
l^
l Recommendations:
- NFS agrees with the conclusions reached in the Reference 2 Deviation Report. The root cause of the power increase was inadequate procedural guidance for handling cooldown effects l'
during reactor shutdown. Therefore, NFS recommends that the e
station general procedures for shutdown be revised:
o To include guidance for calculating core decay heat based on the reactor power history. This guidance should include a procedure for balancing the steam loads and properly controlling the pressure during reactor shutdown, and To provide precautions to warn the operator that reactor cooldown adds positive reactivity and if rod inbrtions are interrupted during the shutdown a power incressa from cooldown can occur.
- NFS recossends that the LaSalle operating staff review this event during operator requalification. A similar event occurred at Limerick in April of 1988 and is subject of an INP0 case study.
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- simulator training on low power and shutdown e included in the reactivity drills prformed during ope requalification.
e rator
- Operations behind the stops with the reactor criti alininated or at least minimized cal should be 0'
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