ML18059A449

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LER 93-010-00:on 930917,primary Coolant Sys Cooldown Rate as Defined in TS Exceeded.Caused by Combination of Personnel Error & Inadequate Procedures.Proper Temp Monitoring Techniques verified.W/931018 Ltr
ML18059A449
Person / Time
Site: Palisades Entergy icon.png
Issue date: 10/18/1993
From: Hillman C, Slade G
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-010, LER-93-10, NUDOCS 9310250185
Download: ML18059A449 (9)
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  • consumers Power GB Slade General Manager POWERiNii MICHlliAN'S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway. Coven. Ml 49043 October 18; 1993 Nuclear Regulatory Commission*

Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - LICENSEE EVENT REPORT 93-010 - EXCEEDING THE PRIMARY COOLANT SYSTEM COOLDOWN RATE AS DEFINED IN THE TECHNICAL SPECIFICATIONS .

Licensee Event Report (LER)93-010 is attached. This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B) as. operation of the plant outside the plant's Technical Specifications .

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NRC Form 366 19*831

  • LICENSEE EVENT REPORT (LERI
  • U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3160-0104 EXPIRES: 8/31 /86 FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 I

I Palisades Plant 015101010121515 1 OF 018 TITLE 141 EXCEEDING THE PRIMARY COOLANT SYSTEM COOL DOWN RATE AS DEFINED IN THE

.~ TECHNICAL SPECIFICATIONS EVENT D.ATE 161 LER NUMBER 161 REPORT DATE 161 OTHER FACILITIES INVOLVED !Bl SEQUENTIAL REVISION FACILITY NAMES MONTH DAY *YEAR YEAR NUMBER NUMBER MONTH DAY YEAR

' N/A 0 1~1°1°1°1 I 0 19 ih 9 3 913 ( ol 1 lo o lo i I o 1 I fl q I ~ N/A 0161010101 I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR t: IChKk OM°' more of tM following/ 1111 N

OPERATING MOOE 191 I

20.4021bl 20.4061oll1 llil - 20.406tcl 60.36icll1 l

60. 7 31oll211ivl 60.731oll2llvl 73.71 lbl 73.71 tel POWER LEVEL 1101 I I 0 20 .4061*111 )(ii)

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60. 7 31oll211viil OTHER !Specify in Abtitract 20.406toll1 lliiil 60.73i*ll2llil 60. 7 31oll211viiillAI below and in Text, 20.4061oll1 llivl 60.73toll2lliil 60. 7 3toll211viiillBI NRC Form 366AI 20.4061oll1 llvl 60. 7 31oll211iiil 60.731oll211xl LICENSEE CONTACT FOR THIS LER 1121 NAME TELEPHONE NUMBER Cris T. Hillman, Staff Licensing Engineer 6ARtA1CI°~ I 716141 -18191113 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 '

MANUFAC* REPORTABLE MANUFAC* REPORTABLE CAUSE SYSTEM COMPONENT TUR ER TO NPROS CAUSE SYSTEM COMPONENT TUR ER TO NPROS I I I I I I I I I I I I I I I I I I I I I I I I I I I I n

SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR '

EXPECTED SUBMISSION YES Vf y ... comp/ere EXPECTED SUBMISSION DATE!

~NO DATE 1161 ABSTRACT fl..imit to 1400 spaco:i. i.e., approxirnaioty fihoen 6in(ll.-st>>e* typt1written line:i) 1161 I I I On September 17, .1993' at 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br />, with the plant in cold shutdown, a primary coolant system (PCS) cooldown was in progress to perform repairs on a leaking pressurizer relief valve nozzle. During the cooldown, the PCS cooldown rate of 20°F/hr in any one hour, as defined in the Technical Specifications, was exceeded.

The cause of this event was a combination of personnel error and inadequate procedures.

Corrective action for this event includes evaluations of the affect on reactor vessel material as a* result of this event, evaluations of the heat-up and cooldown processes used at Palisades, procedure evaluations, veri fi ca ti on of proper temperature monitoring techniques for heat-up and cooldown, verifying equipment operation and reliability, remedial training and simulator performance evaluations of the operating crew involved in this event, management discussions with the operating crews about the implications of this event, and incorporating the. lessons learned from this event into training lesson plans.

\J.S. N\JCLEAA R£GUL.llTORY COM ... tSSIO"19-831 AP9ROVED OMB NO. 3160-0*0*

EXPIRES: 8131.116 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME 11 I OOCl'lET NUMBER 121 LER NUMBEJI 131 PAGE ' "

SEQUENTIAi. REVISION YEM NUMBER NUMBER Palisades Plant Q 5 0 0 Q 2 .5 5 9 3 - Q 1 Q - 0 Q Q 2 OF Q 8 EVENT DESCRIPTION On September 17, 1993, at 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br />, with the plant in cold shutdown, a primary coolant system (PCS) cooldown was in p~ogress. During the cooldown, the PCS cooldown rate of 20°F in any one hour with the cold leg temperature less than 170°F, as defined in Technical Specification 3.l.2a, was exceeded. At the time of the event, the shutdown cooling system [BP] was in service to reduce PCS temperature. The primary coolant pumps had been removed from servi.ce to permit depressurization of the primary ~oolant system following the identification of, and subsequent increased leakage from a through-wall crack on the pressurizer [AB;PZR] power operated relief valve [AB;RV] nozzle. During this period, the leakrate from the PORV nozzle was initially estimated to be 0.1 gallons per minute (gpm), based on the level increase in the containment sump. The PCS cooldown rate was subsequently calculated to be 49°F in the hour following the removal of the primary coolant pumps from service based on the most limiting temperature indicators.

This event is reportable to the NRC in accordance with 10CFR50.73(a)(2)(i)(B) as operation of the plant outside the plant's technical specifications.

CAUSE OF THE EVENT The cause of this event was a combination of personnel error and inadequate procedures.

Because this event involved personnel error, the following informatio~ is provided in accordance with 10CFR50.73(b)(2){J):

(J)(l) Operator action which affected the course of this event was the removal from service of the primary coolant pumps. The removal from service of the primary coolant pumps was a planned evolution during the normal cooldown process of the reactor and was controlled by procedure. System Operating Procedure (SOP) 1, "Primary Coolant System," contributed to the event in that it provided only general guidance as to the frequency of recording PCS heat-up or cooldown information, specifically, " ... data shall be recorded hourly or more frequently ... " Furthermore, the procedure did not sufficientlj caution the operator on what to expect once the PCPs were removed from service.

(J)(2)(i) The error was cognitive in that the reactor operators monitoring the cooldown were aware of the limits for the cooldown rate. They had. recorded times and temperatures that should have alerted them to the excessive

  • cooldown rate, but failed to recognize and act on the information.

U.S. NUCLEAA ~EGUU.TORY COMMtSSlC'<

19-831 APP'IOVED OMB '<0. 3160-0'0'.

  • EXPIRES: 8/31 :86 LICENSEE EVENT REPORT !LERI TEXT CONTINUATION

<.i.CILITY NI.ME 111 DOCKET NUMBER 121 LER NUMBER 131 P.i.GE !'1 SEQUENTIAi. ' REV15'0.N YEM NUMBER NUMBER Palisades Plant 0 5 0 0 0 2 5 5 9 3 - 0 1 0 - 0 0 0 3 c* 0 8 (J)(2)(ii) The error (1) was contrary to an approved procedure in that the PCS cooldown rate defined in the technical specifications and in SOP 1 was exceeded; (2) was aggravated by the procedure in that SOP 1 did not address how the PCS cooldown rate would be affected once the PCPs were removed from servfce, and (3) did not directly result from an activity or task not monitored or covered by the procedure in that the operators could have controlled the cooldown ~sing the guidance in SOP 1. -

(J)(2)(iii) There were no unusual conditions in the work area, the control room, that directly contributed to the error. .

(J)(2)(iv) The individuals involved in this event were licensed reactor operators.

This event_ did not involve the failure of any equipment important to safety.

ANALYSIS OF THE EVENT Prior to this event, a primary cool-ant system (PCS) heat-up was in progress following completion of the 1993 refueling outage activities. During the heat-up, a primary c6olant system (PCS) leak check discovered an uni~olatable leak on top of the pressurizer. *A PCS cool down was commenced in accordance with system operating procedure (SOP) 1, "Primary Coolant System."

The last two operating primary coolant pumps (PCPs) were removed from service in accordance with SOP 1 between 1348 hours0.0156 days <br />0.374 hours <br />0.00223 weeks <br />5.12914e-4 months <br /> and 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br /> on September 17, 1993, when PCS temperature reached 170~F. Normally, the PCPs are not removed from service at this PCS temperature; however, due to the presence of the leak in the pressurizer, the PCPs were removed from service to permit depressurization of the PCS in an attempt to reduce the leakrate.

The initial evaluation of the event determined that between 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br /> and 1456 hours0.0169 days <br />0.404 hours <br />0.00241 weeks <br />5.54008e-4 months <br />, immediately following the removal of primary coolant pump P50C from service, the PCS cooldown rate of 20°F in any one hour when the cold leg temperature is less than 170°F, as defined in Technical Specification 3.l.2a, was exceeded. The cooldown rate was calculated in accordance with the existing procedural requirements using the temperature difference between the outlet temperature of the shutdown cooling heat exchanger at 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br /> and 1456 hours0.0169 days <br />0.404 hours <br />0.00241 weeks <br />5.54008e-4 months <br />. The calculation showed that the PCS temperature had dropped 26°F in that hour using the shutdown cooling heat exchanger outlet temperat~re at the time the PCPs were removed from service as the reference point. An engineering analysis was performed as part of the corrective action for this event and determined that T~olp' at the time the PCPs were removed from service, should be used as the reference point.

Using this previously unrecognized information, the maximum cooldown rate in any one hour was calculated to be 49°F. Based on this new information regarding the use of Tcold

NRC F0tm 3eeA U.S. NUC.lEAA 'IEGUL.>.TORY COMMISSION 19-831 ~OVEO O"'B NO. 316_0-<l'O*

EXP!RES: 813 1:86

. LICENSEE EVENT REPORT !LERI TEXT CONTINUATION FAClllTY NAME 111 DOC.KET NUMBER 121 lER NUMBEJI 131 PAGE 1*1 SEQUENTIAL REVlSION YEAA NUMBER NUMBER Palisades Plant Q 5 Q Q Q 2 5 5 9 3 :- Q 1 Q - Q Q Q 4. OF Q 8 at the time the PCPs were removed frbm service as the. reference point, other previous plant cooldowns were reviewed. We determined that the cooldown rate (using T;old as the reference point following removal of the PCPs from service) _has exceeded the technical

  • J specification limit of 20°F/hr on nearly all of the previous cooldowns since the 1.

incorporation of the 20°F/hr limit with Amendment 131 to the Technical Specifications in 1990. . .

This event has also b~en evaluated by the licensed personnel responsible for exceeding the PCS cooldown rate, and, in a9dition, a Human Performance Enhancement System (HPES) evaluation *has been condOcted to determine what human factors may have significantly impacted this event. The results of both of these evaluations with respect to human performance are discussed below.

The evaluations. concluded that while two licensed reactor operators (ROs) had been assigned the responsibility for .the cooldown, both failed to adequately monitor the rate of PCS cooldown. ~actors that contributed to this failure were:

1. The pressurizer leak event drew the attention of the operators away from adequately monitoring the PCS ~ooldown rate.

The ROs felt a sense of urgency to depres~urize the PCS to stop the leak. Upon completion of depres*surizing th.e PCS-, a reduction in attentiveness to plant conditions occurred. * *

2. Neither ..of the ROs were specifica77y assigne_d nor did either of the ROs assume the responsibility of monitoring PCS cooldown rate following removal from service of the PCPs.

The shift supervisor conducted a pre-job briefing to discuss (1) th~ PCS cooldown, (2) securing. the operating PCPs and charging flow, (3) depressurizing the PCS, and (4) preparing for the PCS drain-down to a level below the leak.

The r.os knew they were responsible for monitoring the PCS cooldown rate but had not communicated the coordination of the activity.

3. The shutdown cooling {Sf!C) return temperature indicated on temperature recorder (TR) 0351 increased when the first primary coolant pump was r.emoved from service.

The ROs expected the SOC return temperature to drop when the first PCP was removed from service. Wh~n the ROs noted the temperature increase they felt they could shut down the second PCP with little or no effect on PCS temperature.

The PCS cooldown rate was exceeded when the second PCP was removed from service.

~RC Form 388A U.S. NUCLEAR ~ECiUL)ITQRY COMl*USSIC~

li-131 ...-ovro Ol.lB.NO. 3'60-0'0' EXF'IRES: 8131 *116 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME 111 DOCIC.ET NUMBER 121 LER NUMBE!I 131 PAGE 141 SEQUENTIAL REVISION YEM NUMBER NUMBER Palisades Plant 0 5 0 0 0 2 5 5 9 3 - 0 1 0 - 0 0 0 5 Q< 0 8

4. Methods for monitoring PCS cooldown, after the primary coolant pumps are removed from service, differs from the method used when the PCPs are in-service.

With the primary coolant system pumps running, the critical functions monitor (CFMS) displays an actual cooldown *rate. *After the PCPs are removed from service, the CFMS generated cooldown rate becomes invalid and TR-0351 is used for monitoring the cooldown rate. TR-0351 digitally displays and records SOC heat exchanger inlet and return temperatures. The cooldown rate must be manually calculated from the TR-0351 readings.

5. Lack of supervisory overview.

The SRO failed to adequately monitor the RO actions during the cool down.*

Safety Significance An erigineering evaluation was performed to evaluate the effects of the PCS cooldown of September 17, 1993 with respect to 10CFR50, Appendix G pressure and temperature limits for the Palisades reactor vessel. On September 17, 1993, a maximum cool down rate of 49°F per hour was achieved with the PCS temperature less than or equal to 170°F. The maximum cooldown rate allowed by technical specifications under these conditions is 20°F per hour. The analysis was also used to determine the need to perform an analysis in accordance with the requirements of ASME B&PV Code Section XI, Appendix E.

During the cooldown on September 17, 1993, the following primary coolant system conditions were established:

  • PCS pressure was being maintained at approximately 235-270 psia (in the manual mode) when the PCPs were removed from service between 1348 hours0.0156 days <br />0.374 hours <br />0.00223 weeks <br />5.12914e-4 months <br /> and 1356 hours0.0157 days <br />0.377 hours <br />0.00224 weeks <br />5.15958e-4 months <br />.
  • At 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br /> all charging pumps were removed from service eliminating the possibility of mass addition to the PCS.
  • At 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> the PCS was depressurized.
    • The PCS was cooling down, therefore, there was no pressur~ increase due to swell.

Since all the mass and volume increase terms were either controlled or eliminated during the cooldown of September 17, 1993 the only remaining limit is the 10CFR50, Appendix G, limits.

~RC Form JeeA U.S. NUCLEAA REGUL-'TORY COM .... 155'0<

19*831 APl"IOVED O°"'B NO. 3'60--010*

EXPIRES: 8131.116 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME-11 I DOCKE"T NUMBER 121 . LER NUMBER Ill PAGE "'

SEOUENTIAi.. RE\/1SION YEAA NUMBER NUMBER Palisades Plant 0 5 0 0 0 2 5 5 9 3 - 0 1 0 - 0 0 0 6 o* O 8 Table 1 shows the operating limits for the Palisades reactor vessel for fait fluence up to 1.8 El9 N/cm 2 in the lO"F to 200°F temperature range. The data was calculated from equations for the circumferential weld with measurement uncertainty included found from a previous engineering analysis and.is presented graphically in Figure 1. The data preserited differs from the Technical Specification, Figure 3-2, in that Figure 1 accounts for pressure and temperature measurement uncertainties. Consequently, the PT.

limits in Figure 1 are used when evaluating the effe~ts of actual plant operating data.

TABLE 1 - Allowable PCS pressure (psig) for various PCS temoeratures and cooldown rates PCS1Temperature Cool down 10°F 50°F 100°F 150°F 200°F Rate 0 deg/hr 383.8 389.4 402 .8 , 430.4 487.6 20 .deg/hr 325.2 331.1 345.4 374.9 435.8 40 deg/hr 264.9 271. 2 286.5 317.99 383.1 60 deg/hr 199.4 206.1 222.5 256.4 326.3

  • 90 deg/hr 139 .1 146.4 163.9 200.1 274.8 100 70.8 78.6 97.5 136.4 216.9 deg/hr Flgln 1

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As tan be seen on Figure 1, for the short time (approximately 10 minutes) that the PCS was pressurized following trip of the first primary coolant pump (PCS temp = 170°F, pressure= 260.psia (245 psig) or less) a cooldown rate of 60°F/hr can be supported without violating 10CFR50, Appendix G, limits. Following depressurization of the PCS to

'<RC F"'m Jee.o. U.S. NUCLEAA REGUL.o.TORY C:;M ... 1SSlCN li-831 ...-ovro o"'e ><O. 3150.,j"*

EXPIRES: 813 L'86 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME 111 DOCIC:ET NUMBER 121 LER NUMBER Ill ?AGE 1'1

  • SEQUENTIAL. REVISION YEM NUMBER NUMBER Palisades Plant Q 5 Q Q Q 2 5 .5 9 3 - Q 1 Q - Q Q Q 7 OF 0 8 16 psia, cooldown rates of up to l00°F/hr can be supported without violating 10CFR50, Appendix G, limits.

There are two other margins that have not been taken into account in the low temperature over-pressure protection (LTOP}, 10CFR50, Appendix G, limits in this analysis. First, the 10CFR50, Appendix G, limits have a term accounting for primary coolant pump differential pressure .. This term reduces the allowable indicated pressure for a given*

temperature since the location of pressure measurement is downstream (through the '

reactor internals) from the location of interest (beltline welds). Once the primary coolant pumps are removed from service, the associated flow losses essentially go to zero and the allowed indicated pressure increases. Second, during the September 17, 1993 event, the shutdown cooling system was in service. In the development of the LTOP setpoints, no credit is taken for the relief capacity of the shutdown cooling system relief valve.

The ASME B&PV Code Section XI, Appendix E, provides a methodology for determining the effects of exceeding cooldown limits. Appendix E need only be utilized if heat-up and/-0r cooldown limits have been exceeded during the course of a transient. As indicated earlier in the analysis, no 10CFR50, Appendix G1 limits were exceeded during the cooldown on September 17, 1993, therefore, no Appendix E analysis is necessary.

Conclusion

/' '

Although,the allowed cooldown rat~ of 20°F/hr was exceeded during the September 17, 1993 PCS cooldown, the 10CFR50, Apperidix G, limits pertaining to reactor vessel protection from brittle fracture were not violated. The caltulated maximum cooldown from this event was approximately 49°F/hr; however, cooldown rates up to 60°F/hr can be supported under th~ conditions which existed on September 17, 1993. Following depressurization,

  • cooldown rates of up to 100°F/hr can be supported. Since no *heat-up or cooldown limits related to 10CFR50, Appendix G, were violated, an analysis, in accordance with ASME Section XI, Appendii E, is .not necessary.

CORRECTIVE ACTION There are numerous corrective actions associated with this event that are summarized as:

  • evaluations of the effect on reactor vessel material as a result of this event and other previous events;
  • evaluations and revision of the heat-up and cooldown processes used at Palisades;
  • evaluations of procedures and incorporation of identified changes;
  • operator training on procedure changes;

NRC F0<m JeeA i.J.S. NUC\.EAA qEGUU.TORY COMMISSION

..* {i-831 -OVEO OMB NO. J\60.¢10*

EXPIRES: 8/J 1 i86 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME 111 DOCKET NUMBER 121 lEll NUMBEll 131 PAGE "1 SEQUENTIAL REVISION YEM NUMBEll NUMBER Palisades Plant Q 5 Q Q Q 2 5 5 9 3 - Q 1 Q - Q Q Q 8 CF Q 8

  • verificatiori of proper temperature monitoring techniques for heat-up and cool down;
  • verification of equipment operation and reliability;
  • removal from licensed operator duties pending evaluation for the SRO and ROs d1rectly involved in this event;
  • remedial training and simulator perfo~manc~ evaluations of the operating crew

.involved in this event prior to their return to licensed.operator duties;

  • discussion with all the operating crews about the implications of this event and similar past events involving personnel error; and
  • incorporating the lessons learned from this event into training lesson plans.

ADDITIONAL INFORMATION None

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