ML19331C919

Responds to IE Bulletin 80-14, Degradation of BWR Scram Discharge Vol Capability. Notes 741231 Failure of One Scram Discharge Vol Level Switch.No Other Failures Recorded Re Water Hammer,Sdv Vent & Drain Valve Operability

ML19331C919
Person / Time
Site:	Cooper
Issue date:	07/14/1980
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
IEB-80-14, NUDOCS 8008200116
Download: ML19331C919 (3)

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l July 14, 1980 Mr. Karl V. Seyfrit, Director U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region IV 611 Ryan Plaza Drive Suite #1000 Arlington, Texas 76011

Subject:

IE Bulletin No. 80-14 Degradation of BWR Scram Discharge Volume Capability

Dear Mr. Seyfrit:

The subfect bulletin ~ described two' events which have raised concern regarding operation related to the control rod drive (CRD) system scram discharge volume (SDV). The bulletin required certain actions be taken and a report submitted in regards to these actions. This letter is submitted as the report on our actions. The following items correspond to the action items listed under Section A of the subject bulletin.

1. In December of 1974, new type SDV level switches were installed per the recommendations of G.E. FDI NSFT. During the hydrostatic testing of the newly installed switches, one of the safety related function switches failed. When removed'from the system and ex-amined, a collapsed float assembly was observed. The failed safety related function SDV level switch was replaced with the " alarm function" SDV level switch. The original model SDV level switch was installed in the alarm function location temporarily until the l new type switch could be obtained from the manufacturer. The

! original model SDV level switch in the alarm function location was l replaced in October, 1975 with the new type SDV level switch.

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c-Mr. Karl V. Seyfric July 14, 1980 Page 2.

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' Ranufac'ture'r' taalysis' of th'e ' failed"SDV' level switch provid4d no "

positive indication of.the cause of failure of the float assembly and switch. Failure of the swi.tch occurred at approximately 1750 psig hydrostatic test pressure; however, manufacturer's records indicate the subject switch had satisfactorily passed two previous factory hydrostatic tests to 1875 psig. Manufacturer tested four identical switches for failure and observed failure at pressures in excess of 1875 psig. Manufacturer speculated that the switch was

' damaged during shipment to the plant or'during installation is the

• system. .No other.instancea of degradation of.SDV level switches. .

have been recorded.

2. No instdnces of degradation of SDV vent and drain valve op'erability have been recorded in plant records.

.The ac.ceptance criterii for the' closing times of the SDV ' vent and. .-

drain valves is 430~seccnds. This'wa's determine'd bp' General Eiec-

• tric, the CRD System designer, to be sufficient to limit the'quan-

'C * " '" tity' off reastbh/ater Mischarged' into the react 6r bu11 ding drain' *

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~ ~ ~ ;*' c system after a scram.'

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'- - *v Th'e,sDE yen't and' drain S valve *closQre was'not rostinely timed sub-'

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seq'uant to' init.alfstartup i doveve'ri she' ' installed SD'V 'ven't, '

'and drain valves were known to'have closed in less than 5 seconds ' -

following startup testing.- The valves'have been recently tested ,

and closure (times are still less than 5 seconds.

3. Station operating procedures and the Technical Specifications require th'e SDV; vent and drain valves.to be open and operable with fuel in the reactor. (except during testing) . Procedures have been

. developed and implemented.to require. periodic, testing of.the:SDV .-- . ..

vent and drain valves (including closure times).

If the SDV vent and drain valves are not operable or are closed for i

more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during operation, the reason I will be logged and the NRC will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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( 4. No events of water hammer or events causing water hammer damage to l SDV related piping have been observed.

5. Existing surveillance procedures are considered satisfactory to detect any degradation of any SDV level switch and any failure will be reported to the NRC as an LER as requirsd'by Technical Specif-i ications.

e Mr. Karl V. Seyfrit July 14, 1980 Pa g e. 3.

• 6. A functional test'and calibrstion check of- the' SDV level' svitches - - - '- - - --

was satisfactorily perforned in March, 1980. A functienal tes: of the SDV revel switches was satisfactorily =ade in May,.1980..- -

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' Fu:ictional tests are scheduled at 3-:ench intervals and a cali- -

bration check is scheduled yearly during the refueling cutage. .

An esti=ated 100 nanhours of engineering tine was spent researching and preparing this response including revising procedures for Ite= 3.

Sincerely . , , , ,

M -

J. M. F11 ant -

Director of Licensing- -

and Quality Assurance

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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551. LITTLE ROCK. ARKANSAS 72203 (501) 371-4422 WILLIAM CAVANAUGH lil June 12, 1980 Vice President Peneraton & ConstruC'Jon 1-060-14 2-060-19 fir. K. V. Seyfrit, Director Office of Inspection & Enforcement U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011

Subject:

Arkansas Nuclear One - Units 1 & 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NTF-6 IE Bulletin No. 80-12 Decay Heat Removal Systen (DHRS)

]perability (File: 1510.1, 2-1510.1)

Gentlemen:

In accordance with Item 7 of the subject bulletin, the following is provided under 10 CFR 50.54(f).

Item a.

Changes to procedures (e.g., emergency, operational, administrative, maintenance, refueling) made or initiated as a rescit of your reviews and analyses, including the scheduled or actual dates c# accomplish-ment. (Note: NRC suggests that you consider the foi u ig. (1) limiting maintenance activities to assure redundancy c- iversity and integrity of DHR capability, ard (2) bypassing or disabling, where applicable, automatic actuation of ECCS recirculation in addition to disabling High Pressure Injection and Containment Spray preparatory to the cold shutdown or refueling mode. )

Response

In accordance with Items 1 and 2 of this bulletin, reviews of the Davis-Besse event and of all DHR degradation events, for which documentation exists, at ANO-1 and 2 has been completed. In addition to these reviews, analyses of our procedures for adequacy of safeguarding against loss of DHR capability and for adequacy of responding to DHR loss events are now being performed. We expect to complete these analyses by August 15, 1980.

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. In response to Item 6, procedure revisions necessary in order to ensure that removal of decay heat from the RCS is accomplished as necessary will be implemented by July 7,1980. Consideration will be given to the fiRC suggestions given above in our reviews and procedure revisions.

Item b.

The safeguards at your facility (ies) against DHR degradation, including your assessment of their adequacy.

Response - ANO-1 In order to address Items 4, 5, and 6 of this bulletin, the hardware capability of ANO-1 to prevent DHR loss events has been and is continuing to be reviewed. Also, our letter dated August 3,1976, to D. Ziemann of your staff presented our analysis to demonstrate that the Emergency Core Cooling System and subsystems meet the single failure criterion pursuant to the requirements of 10 CFR 50.46. Because parts of the ECCS system are also used for DHR, we feel that this analysis, plus the reviews being performed to address Items 4, 5, and 6, adequately addresses the concerns expressed in this bulletin.

Response - ANO-2 The Alio-2 FSAR states in Section 9.3.6 that: "I'o single failure of an active component during residual heat remov'al will result in permanent loss of cooling capability." And: "A single failure of a passive com-ponent in the low-pressure portion of the Shutdown Cooling System during residual heat removal may result in the interruption of the cooldown but will not result in a loss of core cooling." The Safety Evaluction per-formed in support of these statements is included in Section 9.3.6. We feel tha t this evaluation, along with the reviews being done in response to Items 4, 5, and 6, is sufficient to address the concerns of this bulletin.

Very truly yours,.

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William Cavanhugh, III/

l WC:!4AS:ska cc: Mr. Victor Stello, Jr. , Director Office of Inspecti >n and Enforcement U.< Nuclear Regulatory Cemmission Wasnington, D. C. 20555

i STATE OF ARVANSAS )

) SS COUNTY OF PULASKI )

William Cavanaugh III, being duly sworn, states that he is Vice President, Generation & Construction, for Arkansas Pcwer & Light Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this Supplementary Infonnation; that he has reviewed or caused to have reviewed all of the statements contained in such information, and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, infonna-tion and belief. l

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William Cavar,augh III SUBSCRIBED AND SWORN T0 before me, a Notary Public in and for the County and State above named, this /d. day of U 4 /_/ 1980.

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My Comission Expires:

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