ML20085L658

AO 50-267/75/7:on 750121-23,reactivity Anamoly Occurred Due to Release of Hopper of Reserve Shutdown Mall Into Core. Cause Not Stated.Valve Seats Modified

ML20085L658
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/04/1975
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	Howard E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20085L663	List: ML20085L658
References
AO-50-267-75-7, RO-750304, NUDOCS 8311020225
Download: ML20085L658 (8)

Text

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'Q, pubne service company *e celerado P. O. Box 361, Platteville, Colorado 80651 q

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March 4, 1975 ' f'.S i W w.h n. \=.

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Mr. E. Morris Howard, Director '

Nuclear Regulatory Commission Region IV , _ _ ,,

Office of Inspection & Enforcement 611 Ryan Plaza Drive ,

Suite 1000 Arlington, Te.tas 76012 REF: Facility Operating License No. DPR-34 Duuius Ku. 30-16's

Dear Mr. Howard:

Enclosed please find a copy of Abnormal Occurrence Report No. 50-267/75/7, preliminary, submitted per the requirements of the Technical Specifications.

Very truly yours,

'y H. Larry Brey Superintendent-Operations Fort St. Vrain Nuclear Generating Station MLB:il cc: Mr. Angelo Giambusso Q' /

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' REPORT DATE: January 21-23, 1975 ABNORMAL OCCURRENCE OUCURRENCE DATE: March 4, 1975 FORT ST. VRAIN NUCLEAR CENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO ,

P. O. BOX 361 7, , ,

PLATTEVILLE, COLORADO 80651 '-<.

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REPORT NO. 50-267/75/7 , ,'

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'e' PRELIMINARY -

IDENTIFICATION OF 'f-OCCURRENCE: '.

• N Because of the inadvertent release of one hopper of reserve shutdown material into the core and because of moisture in excess of 10,000 ppa, a reactivity ananoly of 3 0.012 LR occurred. This is defined as an Abnormal Occurrence Ter Technical Specification SR 5.1.4.

• CONDITIONS PRIOR TO OCCURRENCE: Steady State Power Routine Shutdown Hot Shutdown Routine load Change

• Conditions at  ;

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Shutdown on 1/23/75. .lefual ing fhutdown Reactor critical at acproxi-RouHn a Startup mately 10-5 % of Rated Thernal Power for training purposes.

The major plant parameters at the time of theevent were as follows:

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Sacondary Coolant Pressure 1250 psig Tem:perature 182 ay Tiow N349,000 #/hr.

Primary Coolant ,

Pressure 234 psig Temperature 182 *F Core Inlet 193 'F Core Outlet l

~ Elow 2 x 105 #/hr.

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• - 1 DESCRIPTION OF OCCURRENCE On January 21, 1975, the Reactor was taken critical for training purposes and it was noted that the critical configuration was somewhat different than ex-pected. As seen by .the data in Table 1, the core appreared to be less reactive by between .004 and .007 Ap. Since the reactivity difference between the enl-culated and actual critical control rod configuration throughout the "A" Series tests never exceeded .001 Ap, it was somewhat surprising except for the fact that the core te=perature was clearly not defined. The core temperature was assu=ed to be the same as the average outlet temperature, but since there was no Helium flow and steam had been flowing through the reheaters (an external source of heat) as evidenced by the change in average outlet temperature from 180* to 220*F between beginning and end of control rod withdrawal, it was assu=-

ed that this reactivity difference was probably due to a higher core temperature.

On the fellowing day, the Reactor was taken critical again and it was noted that the same reactivity difference as observed previously persisted. For both critical runs, the core appeared to be less reactive by about .008 Ap(see Table 1). There was no Pri=ary Coolant flow and the same uncertainity in core te=perature existed. In addition, there was a question as to whether or not the reserve shutdown balls had been accidentally inserted into Region No. 1 of the core during pressure testing of the new seats in the actuating valves.

To clearly resolve the reactivity difference, the "A" and "B" Helium circulators were placed on pelten wheel drives at 0130 the morning of January 23, 1975.

Plans were to establish a kncun uniform core te=perature, take the Reactor criti-cal and observe if a reactivity difference still remained. At 0630, criticality was achieved and although the core was still observed to be less reactive, N.005Ap, it was noted to be less than that of the previous day. It was thought the core temperature was being reduced as evidence by the fact that the critical control rod position was still changing. By 0800, the reactivity difference had disappeared co=pletely and the expected critical rod withdrawal position for group.2B was the same as the actual, 72.5 inches withdrawn. This is the same control rod group that had been calibrated during SUT A-8, the temperature coefficient measure = cats.

However, instead of remnining constant at this control rod position, the reactivity of the core continued to change and to hold the Reactor critical, group 2B had to be continually inserted. There was some indication that the moisture level in the vessel was increasing but it was not outside the range that had previously been encountered. See Abnor=al Occurrence Report No. 50-267/75/3.

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By 0915, group 2B had been fully inserted and about .0095 Ap had been added. The Reactor was scra==ed at this time until more definitive information on the moisture level could be established.

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.9 Paga 3 APPARENT CAUSE OF OCCURRENCE: ___

Design Unusual Service Cond.

Including Environment Manufacture Component Failure Installation /Const. X Other (specify)

Operator (See below)

Procedure During the modification work done on the Reserve Shutdown Actuating Valves to install the new sof t seats, a pressure test was run to determine the new seat leakage rate. It is felt that the reserve shutdown material was introduced into the core at that time.

Introduction of moisture into the Primary Coolant system occurred through the Helium circulator bearing water system.

ANALYSIS OF OCCURRENCE:

It was noted that if the Icserve shutdown material had been inadvertently inserted into Region 1 of the core prior to the first criticality on January 21, 1075, thc ocgotisc a Livity worth of.thu noturial la that portial'y wdded region would be about the same as the anomoly observed during Reactor operation on January 21, & 22, 1975. On January 23, 1975 it was also determined that the ,

moisture level in the Pri=ary Coolant was in excess of 10,000ppsv. In section 14 of the FSAR, the reactivity coefficient of water in the core is given as 2.1 x 10-5 Ap/ pound of water, so that a .01 Ao change in equivalent to about 500 pounds of water in the core. A test was completed in San Diego in which a graphite sanple was suspended in a Eelium at=osphere with a moisture content essentially the same as that in the Primary Coolant. In that' test it was shown~that as much as 1000 pounds of water could have been present in the core during the reactivity change.

If this were the case, the positive reactivity addition could have been'as large as .02 ap. The shutdown marg,ia would still be. greater than .09 Ap , which is nore than adequate.

As seen in Table 1, the unanticipated reactivity change was about +.015 Ap. This agrees well with what would be expected if about 750 pounds of water had been added to the core. Since no measure of the total reactivity change is known, it is assumed that about 1000 pounds of water had been present in the core during the incident.

On February 18, 1975, it was shown that the reserve shutdown material had definitely been inserted into Region 1.

Maintenance Procedure 11-2 was issued to cover the replacement of the hard-type seats of Valves "J" and "K". (See PI 11-1, attached) with soft-type seats, to correct leakage probleus.

- The hard seats were Temoved from all "J" and "K" Valves and modified to a sof t seat by General Atomic, using a design given by VELAN, the Valve. Manufacturer.

Public Service. Company personne1' removed and reinstalled the seats. ..

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To test seat leakage following installation of the modified seats, the section of pipe from the Helium Cylinder to the "J" and "K" valves was pressurized using the He bottle as a pressure source. MP 11-2 called for pressurizing to 100 lbs. thrcugh a pressure reduced and gauge installed between the bottle and

) the "M" valve, with the "J" and "K" valves closed and the downstream "H" valve open. The "H" valve was open to provide a path for leakage, if any.

A pressure decay on the regulator gauge would indicate leakage.

This test was run, on Region 1, sanctice between January 1 and January 5,1975.

It was found to be icpossible to build up any pressure in the test section.

Disasse=bly and exaninatica of the "J" and "K" valves showed that the seat mod-ification had changed the seat assembly configuration so that the seat did not seal with the ste= full-down.

The disc c:nfiguration was modified on all "J" and "K" valves to ensure seat-sealing and the procedure, including testing, co=pleted on all 37 regions.

Work =en doing MP 11-2 who tested Region 1, and discovered the lack of sealing of the "J" and "K" valves, report that, on opening the bottle to pressurice the pipe section, "the gauge swung rapidly upscale, to an unknown high and immediately fell to near zero." The section could not be pressurized as required, nor could Encugh the Reserve Shutdown hopper pressure switch be activated af ter this time.

gas at suf ficient pressure ta rupture the reserve shutdown hopper disc, was appar-ently released at the tire of the test.

This vole =e of gas, at a pressure above the 100 pound test limit, was caused by the work =an opening the bottle valve too fast, and over-pressurizing the disc before the regulator could respond.

Regica 1 was the first one tested, and no further difficulties were encountered.

The reserve shutdcwn caterial and broken rupture disc was recovered from Region No. 1 of the core, using the Reserve Shutdown Recovery Vacuus Tool.

It should be noted that the rupture disc broke as designed. The disc was completely rc=oved frc= the holder providing a , clear path for the reserve shutdown material into the core.

The reserve shutdown material was found to be in excellent condition with no material da= age and will be reused.

In July of 1974, a substantial amount of water had also been inadvertently added to the PCRV. Following that incident, a study was made to determine what ef fect storage and noisture exposure would have on the condition of the fuel rods used in the Fort St. Vrain Core. The results show that no degradation of the fuel elernnt components is to be expected.

Consideration has also been given to the condition of the lumped burnable poison  !

rods. It has been determined that burnable poison rods contained in water saturated l Helium gas at 100 to 150*F for as much as ninety (90) days would not suffer any l significant oxidation of the B C.

4 All of the B 02 3 present in the lumped burnable poison rods would hydroly:e to boric acid in this environment, but since the specification limits the B 02 3 content to less than 1% and the fabrication records show it to be about 0.5%, the maximum loss of burn would be limited to less than 1%

of the total. This would result in a reactivity change of about .001 op at the cost.

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n U Pagt 5 CORRECTIVE ACTION:

Since all region reserve shutdown system activation valves have been modified, successfully tested, and the reserve shutdown hopper rupture disc integrity has been verified, no correction action is required.

With respect to the source and ecchanism for getting moisture into the Primary Coolant, we are presently making an extensive study of the Helium Circulator bearing water and buffer helium systems. We are also reviewing the complete Helium Circulator operating procedure.

In addition to the review of the in-plant systems, we arc setting the spare helium circulator up in aspecial test stand in San Diego and will determine the effect on the bearing water / buffer helium systems of various transients we have experienced at the plant.

Dry-out of the Primary Coolant system began i==ediately following the discovery of high coisture in the system. As socn as the Primary System is dry, it is planned to take the Reactor critical and demonstrate that no reactivity difference still persists.

FAILURE DATA /

SDfILIAR REPORTED OCCURRENCES:

!.b=:-- ' 0::1 r:::: ": pert 50-2S?!?5/2 L: :::::L:t:1.:Lth th: ::== t=:ident.

PROGRMf!ATIC D' PACT Not known at this time.

CODE IMPACT lione RECOMMDIDED: _ APPROVED:

• l_f -- um Frank M. Mathie 'r'ederic E. Swart """

Superintendent, Maintenance Superintendent, Nuclear Production Ft. St. Vrain Nuclear Generating 7t. St. Vrain Nucicar Cencrating Station . Station e

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o-TABLE 1 ,

' - Description of Critical Configurations Following SUT A-8, Part 2 .

(1) Calculated Actual Reactivity

(* F) Critical- . Critical Difference Date Test Yeore Position- (2) Position ao '

CR GROUP 2B Withdrasm 11/30/74 Einal critical 154 62" 60.8" +.0003 after Cooldown

- for SUT A-8 12/22/74 RT-320 99 37.5" 37" s0

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1/21/75 Training Exercise 180 Initial 69" 89" .0075

- 220 Final 78" -

89" .0044 1/22/75 Training Exercise 210 . 76" 97" .008-1/22/75 Training Exercise 210 76" 97" .008

.1/23/75-0630 Training Exercise 195 72.5" 87" .0054 0800 195 72.5" 72.5" 0.0 0915 195 72.5" 0" +.0095 1 T - the 2::: l' aver ecnditions,? '*** 'aa- *-Sa- * * '- Y -- ' #d--' '- !

Technical Specification. core out .

With no helium flow, this is not an accurate indicator. I l ,

1 The calculated position assumes the nor::ial withdrawal sequence and is based on:

the core temperature as ceasured from SUT A-8. .

3. Steam was being introduced into the reheaters just prior to .this run...The-core

.had been %180* for the previous 3 days but with the steam addition and no helium flow, the average outlet te=perature was increasing but the actual' core' temperature-was not known. -

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