ML19262A222

Abnormal Occurrence 50-289/75-10:on 750410,reactor Criticality at 505 F.Caused by Procedural Inadequacies.Plant Operations Review Committee Recommended Update of Fuel Reactivity Versus Core Burnup & Procedural Revisions

ML19262A222
Person / Time
Site:	Three Mile Island
Issue date:	04/21/1975
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19262A223	List: ML19262A222 ML19262A228
References
GQL-0952, GQL-952, NUDOCS 7910260562
Download: ML19262A222 (4)

Text

_NRC D1"TRIBUTION FOR PART 50 DOCXE7 ' ATERI AL

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CONTROL NO:- B FILE: INCIEF'72 ?EPCC ILE FROM: Me tropolitan Edison co. DATE OF DOC

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Reading, Penna. DATE REC'D LTR TWX RPT OTHER 4=21=75 4-26-75 XX TO.

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1 50-289 DESCRIPTION:

• ENCLOSURES:

Ltr. reporting Abnora. Occurr. # 75-10, on 4-10-75, ' concern. Reactor Criticality at 595* F * '

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PLANT NAME: Three Mile Island #1

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ACST CFFICE BOX 542 R E AOING. PENNSYLV ANI A 19603 TELEFwCNE 215 - 9292G1 April 21, 1975 GQL C952 Director Directorate of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20535

Dear Sir:

h Operating License DPR-50 Docket #50-289 In accordance with Technical Specifications for the Three Mile Island Nuclear S tation Unit 1, we are herewith submitting abnormal occurrence report number A0 50-289/75-10. This submittal is being made one day after the required due date, as was dicussed by your of fice's Mr. D. Jaffe and our Mr. D. N. Grace.

(1) Report Number: A0 50-289/75-10 (7a) Report Date: April 21, 1975 (2b) Occurrence Date: April 10,1975 (3) Facility : Three Mile Island Nuclear Generating Station Unit 1 (4) Identification of Occurrence:

Title: Reactor Criticality at 505 F Type: An abnormal occurrence as defined by the Technical Specificaticns, paragraph 1.8b , in that Reactor Criticclity at 505 F constitutes exceeding a Limiting Condition for Operating as contained in Specifica: ion 3.1.3.1.

(5) Conditions Prior to Occurrence :

A plant heatup was in progress , in preparation for a routine reactor startup, with =ajor plant parameters as follows:

Power: Core: 0%

Elec.: 0 (Gross)

RC Flow: 105 x 106 lb/hr )k RC Pressure : 2155 psig 1402Q RC Temp.: 505 F

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Directorate of

• ice. ing April 21, 1975 PRZR Level: 100 in.

PRZR Temp . : 650 F

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(6) Description of Occurrence:

While in the process of heating-up, prior to commencing a routine reactor start-ep, the reactor went critical at 505 F, which is in violation of Technical Specification 3.1.3.1, wherein it is required that the reactor coolant temperature be greater then 525 F before going critical. Upon making the determination that the reactor was critical, the group 4 control rods were inserted from 100% withdrawn to 68% withdrawn, which served to shutdown the reactor. A reactivity balance was then conducted which showed the reactor to be shutdown by greater than 1% AK/K.

(7) Designation of Apparent Cause of Occurrence:

The apparent cause of the occurrence is procedure, for reasons as follows:

a. Initial calculations had predicted that the reactor would go critical when both the Reactor Coolant System (RCS) temperature reeched 532 F, ,nd the RCS boron concentration had been decreased to 960 ppm. Fticher, for a RCS temperature of 532 F, it had been calculated that 21,000 gallons of demineralized water needed to be added to the RCS to achieve a boroa c(ncentration reduction from 1260 ppm to 960 ppm. Due to procedural inadequacies , however, while in the process of heating up addition of the 21,000 gallons of demineralized was completed when the RCS temperature was only 505 F, vice 532 F. This then resulted in excess positive re-activity being present due to the negative moderator temperature coef ficient of reactivity,

b. An additional procedural inadequacy, (which was not discovered until the next reactor startup on April 14, 1975) was that the f uel reactivity versus EFFD curve, which is usad in the procedure for calculating at what rod index the reactor w.'ll go critical, had not been up-dated to reflect the most current boron con-centration versus fuel burn-up reactivity data (note: this data had been derived af ter about 10% of the total core burnup, as suggested by Technical Specification 4.10 " Bases") . Fu rther ,

the f ailure to up-date the curve was such that there resulted a seccnd component of excess positive reactivity in addition to that in a. above.

(8) Analysis of Occurrence:

The maximun power attained was 2 x 10-10 amps on the intermediate range neutron detectors. The power rise was terminated by operator action.

In the event that no operator action was taken, the reactor would have been protected and the reactor power rise would have been terminated by negative Doppler effect , high reactor coolant system pressure trip or high neutron flux level trip. TMI-l Safety Analysis , Chapter 14 of the FSAR, states that the thermal power would not have approached the design overpower condition and the peak pressure would not have exceeded the t,4v_

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Directorste of Licensing April 21, 1975 allowable code limits. Based on the above, this event is not considered to have threatened the health and safety of the public.

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(9) Corrective Action:

Prior to the next reactor startup, immediate corrective action was taken to prevent simultaneous heatup and deboration prior to 525 F ,

until proper procedural guidance is obtained.

The Plant Operations Review Committee (PORC) met shortly af ter the incident, approved the i= mediate corrective action, and in addition recommended to the Station Superintendent that the additional long term corrective actions be taken

a. update the Fuel Reactivity versus Core Burnup curve

b. all licensed operators review this occurrence prior to conducting a plant heatup

c. when decreasing the RC3 boron concentration by more than 50 ppm while shutdown, plot 1/M (a measure of the neutron level count rate) versus the amount of water added to the RCS

d. when shutdown, record Source Range counts per second on a continuous recorder when decreasing RCS boron concentration.

The Station Superintendent reviewed and approved PORC's findings and has taken steps to ensure tha*. item a. above is accomplished prior to the next reactor startup, and that items b , c, and d are also implemented.

(10) Failure Data:

Not Applicable.

Since rely ,

. C. Arnold Vice President RCA:DNG:tas cc: Directorate of Regulatory Operations , Region i U.S . Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 File: 20.1.1 / 7.7.3.5.1 .

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