05000416/LER-1990-001, :on 900112,determined That Tech Spec Reactor Coolant Cooldown Rate Limit of 100 Degrees F Exceeded in Reactor Bottom Head Drain Pipe.Caused by Programmatic Deficiencies.Incident Rept Form Improved

:on 900112,determined That Tech Spec Reactor Coolant Cooldown Rate Limit of 100 Degrees F Exceeded in Reactor Bottom Head Drain Pipe.Caused by Programmatic Deficiencies.Incident Rept Form Improved

ML20006E046
Person / Time
Site:	Grand Gulf
Issue date:	02/09/1990
From:	Byrd R, Cottle W SYSTEM ENERGY RESOURCES, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AECM-90-0032, AECM-90-32, LER-90-001, LER-90-1, NUDOCS 9002160163
Download: ML20006E046 (5)

LER-1990-001, on 900112,determined That Tech Spec Reactor Coolant Cooldown Rate Limit of 100 Degrees F Exceeded in Reactor Bottom Head Drain Pipe.Caused by Programmatic Deficiencies.Incident Rept Form Improved

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(1)
4161990001R00 - NRC Website

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Emmy n-i a t o" us 39'50 b! (iO) 437 (609 Wmiam T. Cottle m nmw Naorm Op4"abons February 9,1990 U.S. Nuclear Regulatory Commission Mail Station PI-13*/

Washington, D.C.

20555 Attention: Document Control Desk Gentlemen:

l SUBJECT: Grand Gulf Nuclear Station I

Unit 1 Docket No. 50-416 License No. NPF-29 Technical Specification Actions Delinquent Due To Programmtic Deficiencies and Personnel Error LER 90-001-00 AECM-90/0032 Attached is Licensee Event Report (LER) 90-001-00 which is an interim report.

Yours truly, c.o F~~~ Cw :-

WTC:cg Attachment cc:

Mr. D. C. Illntz (w/a)

Mr. T.11. Cloninger (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr.11. L. Thomas (w/o)

Mr.11. O. Christensen (w/a)

Mr. Stewart D. Ebneter (w/a)

Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta St.,

N.W., Suite 2900 Atlanta, Georgia 30323 Mr. L. L. Kintner, Project Manager (w/a)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l

Mail Stop 14B20 Washington, D.C.

20555

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w, rawan enro om ove.m=~ e~.* con an evaluator conducting an Off-shift post Trip On January 12, 1990, Analysis determined that the Technical Specification reactor coolant cooldown rate limit of 100 degrees F per hour was exceeded in the reactor bottom head drain pipe following a manual reactor scram on An engineering evaluation of the out-of-limit December 30, 1989..

condition was not completed within the Technical Specification 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limit after discovery.

The failure to recognize the out-of-limit condition on December 30, The On-Shift Post 1989 was primarily due to programmatic deficiencies.

Trip Analysis procedure and lleatup/Cooldown Data Shoots have beenThe fa changed to preclude recurrence. evaluation within the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Technical Specificatio to a verbal misunderstanding betwocn the Shif t Superintendent and the Shitt Superintendents Operations Staff in the administrative office.and Operations Staff w distinct communication and thorough research in resolution of concerns.

The engineering evaluation of the out-of-limit condition concluded that the impact of the cooldown rate was insignificant when compared to There are no structural integrity concerns the design usage factor.

associated with continued operations.

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A.

Reportable Occurrence r

On January 12, 1990, an evaluator conducting an Off-Shift Post Trip i

Analysis determined that the Technical Specification reactor coolant cooldown rate limit of 100 degree F per hour was exceeded

'i following a manual reactor scram that -occurred na December 30, 1989.

Due to a verbal misunderstanding, the engineering evaluation of the.out-of-limit condition was not performed as requ. red by Technical Specification 3.4.6.1 until January 15, 1990.

This situation is reported pursuant to 10CFR50.73(a)(2)(1)(B).

I B.

Initial Conditions' The plant was operating at 100 percent power at the time of discovery on January 12, 1990.

t C.

Description of Event

On December 30, 1989 at 1758, Control Room Operators manually scrammed the reactor following a loss of Plant Service Water (EIIS code: KI).

During the course of the event, reactor pressure increased sufficiently to cause an ATWS/ARI Recirculation Pump Trip

. ith no recirculation pumps operating, there was less (RPT).

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coolant mixing in the bottom head region of the reactor vessel.

Thus, cooler water from the Control Rod Drive system (EIIS code:AA) caused a faster cooldown rate in the bottom head region.

In the following hour, the temperature indicated by the temperature element on the bottom head drain pipe decreased approximately 125 degrees F.

This cooldown rate exceeds the 100 degree F per hour cooldown rate limit of Technical Specification 3.4.6.1.

The manual scram event was reported separately in LER 89-019-00.

The excessive cooldown rate was not identified until January 12, 1990 during the Off-shift Post Trip Analysis. Technical Specification 3.4.6.1 requires the temperature to be restored within 30 minutes; an engineering evaluation be performed to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; and a determination be made that the reactor coolant system remains acceptable for continued operations or be in at least Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

'Because the out-of-limit condition was not recognized until January 12, 1990, the plant restarted on December 31, 1989 without conducting an engineering evaluation of the condition to determine the acceptability of the reactor coolant system for continued operation.

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When the condition was identified on Friday, January 12, 1990, an Incident Report was hand carried to the Operations Shift Superintendent for review and processing. The Shift Superintendent believed that the required action had already been completed and did not proceed with the actions required by Technical -

Specification 3.4.6.1, On Monday, January 15, 1990, plant Staff recognized the error and notified the Design Engineering organization to conduct the evaluation. SERI had previously conducted such an evaluation for similar cooldown conditions following an ATWS/ARI RpT on August 14, 1989 (LER 89-012). That evaluation concluded that the impact of both excessive heatup and cooldown rates experienced on August 14, 1989 were insignificant when compared to the design usage factor and that there were no structural integrity concerns associated L

with continued operations. The Design Engineering Group confirmed on January 15, 1990, that the impact of the cooldown rate l

experienced on December 30, 1989 was likewise insignificant when l

compared to the design usage factor and that there were no structural integrity concerns associated with continued opera ^ cions.

i D.

Apparent Cause The failure to recogn'ize the out-of-limit cooldown rate on December l

l 30 was primarily due to programmatic deficiencies. The Scram l

Recovery procedure provided data sheets for logging temperatures for a heatup/cooldown record. The data sheets also required review signatures after completion by the Cc. trol Room Operator and the l

Shift Supervisor. However the data sheets did not includo l

acceptance criteria and did not clearly assign responsibility for determining if out-of-limit conditions existed. Additionally, the

.On-shift post Trip Analysis procedure did not require a check of L

heatup or cooldown rates to ensure acceptability prior to plant l-restart.

The cause of the failure to initiate an engineering evaluation on January 12, 1990 was personnel error due to verbal misunderstandings between Operations Staff.and the licensed Shift Superintendent. The Incident Report was written by Operations Staff in the administrative office and carried to the Shift Superintendent to make required notifications and initiate appropriate Technical Specification actions. From his conversation with the Operations Staff, the Shift Superintendent concluded that the Technical Specification actions had already been complied with, g**amana*

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E.

Supplenental Corrective Actions The Heatup/Cooldown Data Sheets have been changed to require a verifier to evaluate the temperature differential at every 60 minute segment to identify any out-of-limit heatup/cooldown rates.

The data sheets will also be changed to require a verifier to

'i compare Reactor pressure and head flange and shell flange temperatures every 30 minutes. The verifer will ensure that a plot of Reactor metal temperatures vs Reactor pressure is to the right of the applicable curve as required in Tehnical Specification 3.4.6.1.

This change will be incorporated by 03/01/90.

Additionally, completed data cheets require a final documented determination of whether or not the recorded conditions are acceptable when compared to administrative and Technical Specification acceptance criteria. This acceptability i

determination must be approved by the Shif t Supervisor or Shif t Superintendent.

l The On-shift post Trip Analysis procedure was also changed to

' require a check.of the acceptability of the cooldown/heatup rates.

i A memorandum was issued to the operations Staff and to the Shift Superintendents informing them of the incident and the expectations of distinct communication and thorough research in the resolution of concerns. The Incident Report form will be enhanced to prompt more complete reviews for Technical Specification compliance. The change will be implemented by April 30, 1990.

Corrective Actions to preclude excessive cooldown rates are addressed in LER 89-019-00.

F.

Safety Assessment

Cooldown rates at the reactor vessel head flange, shell flange, bottom head, and in the recirculation loops were within the 100 degrees F per hour limit. The heatup/cooldown record shows that the temperatures at the bottom head drains. decreased from 550 degrees F at 1815 to 425 degrees F at 1915. All subsequent cooldown rates were within limits.

The engineering evaluation confirmed that the impact of this cooldown rate was insignificant when compared to the design usage factor and that there was no structural integrity concerns associated with continued operations.

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