ML19318A812

Responds to NRC Ltr Re Violations Noted in IE Insp Repts 50-313/80-05 & 50-368/80-05.Corrective Actions:Revised Procedure 1304.23 & 2304.15 Requiring Specific sign-off for Each Leak Rate Test Following Containment Entries

ML19318A812
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/09/1980
From:	Trimble D ARKANSAS POWER & LIGHT CO.
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML19318A811	List: ML19318A812
References
1-050-07, 1-50-7, 2-050-08, NUDOCS 8006240186
Download: ML19318A812 (5)

Text

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s ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK. ARKANSAS 72203 (5011371-4000 May 9, 1980 1-050-07 2-050-08 Mr. 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 and 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6 Response to Inspection Reports 50-313/80-05 and 50-368/80-05 (File: 0232, 2-0232)

Gentlemen:

In response to the Items of Noncompliance included in the subject re-ports, the following is provided.

NOTICE OF DEVIATION Rased on the results of an NRC inspection conducted during the period of February 22 through March 21, 1980, it appears that one of your ac-tivities was not conducted in accordance with your commitments to the Commission as indicated below:

Item 4 of IE Bulletin 79-21 required the licensee to " Review and revise, as necessary, emergency procedures to include specific information obtained from the review and evaluation of Items 1, 2 and 3 to ensure that the operators are instructed on the potential for and magnitude of erroneous level signals. All tables, curves, or correction factors that would be applied to post-accident mon-itors should be readily available to the operator. If revisions to procedures are required, provide a completion date for the revi-sions and completion date for operator training on the revisions."

The licensee's response to this Bulletin, dated September 24, 1979, included the following commitments with respect to Item 4: .

8 0062 40 b MEMBER MiCCLE SOUTH UTIUTIES SYSTEM L e

Response ANO-1

" Correction factors as appropriate from Tables 1-ANO-1 and 2-ANO-1 will be made readily available to the operators for post accident monitoring. Instruction will be provided as to its use. This will be completed by November 15, 1979.

Response ANO-2

" Correction factors as appropriate from Tables 1-ANO-2 and 2-ANO-2 will be made readily available to the operators for post accident monitoring. Instruction will be provided as to its use. This will be completed by November 15, 1979."

Contrary to the above, the correction factors had not been made available_to the operators of either Unit as of March 20, 1980.

This is a deviation. (313/80-05-01; 368/80-05-01)

Response

The correction factors _were made available on May 2, 1980 with the approval and distribution of Standing Order 47 meeting all com-mitments of I.E.Bulletin 79-21. An action tracking system has been established to assign and track Nuclear Regulatory Commission commitments and should prevent further noncompliances of this nature.

NOTICE OF VIOLATION Based on the results of an NRC inspection conducted during the period of February 22 through March 21, 1980, it appears that certain of your activities were not conducted in full compliance with NRC regulations and

-the conditions of your-license (DPR-51), as indicated below:

1. 10 CFR 50.59(b) states, in part, "The licensee shall maintain records of changes in the facility and of changes in procedures made pursuant to this section, to the extent that such' changes constitute changes in the facility as described in the safety

analysis report or constitute changes in procedures as described in the safety analysis report. The licensee shall also maintain records of tests and experiments carried out pursuant to paragraph (a) of this section. These records shall include a written safety evaluation which provides the bases for the determination that the change, test or' experiment does not involve an unreviewed safety question."

Contrary ~ to the above, the licensee did not retain a written safety evaluation for two design changes:

DCR 660,. Unit 1, Delete the group 7 in-limit for the CRDCS -

sequencer.

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f DCR 633, Unit 1, Replace the RCP lube oil level switches.

These design changes constituted changes in the facility as de-scribed in the Final Safety Analysis Report.

This is a deficiency. (313/80-05-02)

Response

Current ANO procedures require that a written Safety Evaluation be performed on all design changes initiated at ANO. The design changes in question were originated in the Little Rock General Office (LRGO). A written safety evaluation will be performed for these two DCR's by July 14, 1980. To prevent further noncom-pliance, the LRGO procedures will be changed to provide for an adequate review of all DCR's. This change will be made by July 14, 1980, to allow us adequate time to make the necessary reviews. In the interim, all DCR's originated in the LRGO will have a written safety evaluation attached prior to approval.

NOTICE OF VIOLATION Technical Specification 4.4.1.2.5 requires that, "If a personnel hatch or emergency hatch door is opened when reactor building integrity is required, the affected door seal shall be tested."

Technical Specification 3.6.1 requires that, " Reactor building integrity shall be maintained whenever all tree (3) of the following conditions exist:

a.

Reactorcoolantpressureis300psfgorgreater.

b. Reactor coolant temperature is 200 F or greater.

c. Nuclear fuel is in the core."

Contrary to the above, on June 20, 21, 22, and October 20, 1979, per-sonnel entries were made into the Unit I reactor building when all three conditions were met requiring reactor building integrity and no tests were conducted on the affected personnel hatch door seal.

This is an infraction. (313/80-05-03)

Response

A revision has been made to procedure 1304.23 and procedure 2304.15 (units 1 & 2 local leak-rate testing procedures) which require specific sign-off for each leak-rate test requirement following containment entries. It is felt that these additional' controls over key-control and logging entries will minimize the possibility of recurrence of such violations.

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' -a NOTICE OF VIOLATION Technical-Specification 4.8.1.1.2.c.2 requires that each diesel gen-erator shall be verified operable at least once per 18 months by " Veri-fying'.that the automatic sequence time delay relays are OPERABLE with the interval between each load block within i 10% of its design in-terval."

The load block design intervals are obtained from the component start

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times given in the Table 8.3-1 of the Final Safety Analysis Report.

This table lists component start times as 5 seconds for Service Water (SW) Pumps, 10 seconds for High Pressure Safety Injection (HPSI) Pumps,

- and 15 seconds for Low Pressure Safety Injection (LPSI) Pumps. Thus, the design intervals are 5 seconds between the SW and HPSI pumps and 5 seconds between the HPSI and LPSI pumps.

Contrary to the above, as-lef t data from the Integrated Safeguards Test, 2105.03, performed on February 21, 1980, anC from the Testing of

- ESF Time Delay Relays, 2304.87, performed on March 13, 1980, indicate

. that the interval between 2P89C (HPSI) and 2P60B (LPSI) was 4.38 seconds and the interval between 2P4B (SW) and 2P89C (HPSI) was 4.23 seconds.

Response

The criteria established in procedure 2105.03 is based on each separate load block closing in on the bus after its design delay (i.e., 5, 10, 15, or 25 seconds, etc.) within a tolerance based on the design interval between load blocks. Hence, if a component l were to load by design 5 seconds following the previously applied i i

load, and 5 seconds prior to the next load, criteria for time delay relay setting would be the design delay' time (5, 10, 15, 20, or 25 l seconds, etc.) plus or minus 10% of the design interval of 5 sec-onds. In the case of 2P-89C, a time delay criteria of 10 0.5 l

, seconds was established, and a time de: lay criteria of 5 0.5 seconds was established for 2P-4B loading.

In all test cases, load blocks were sequenced on to the bus within the criteria so established. Also, diesel generator performance (voltage and frequency response and stability) were monitored and

were satisfactory for all load applications.

It is our opinion that establishing load-block timing criteria in the manner discussed meets the requirements of Technical Specifi-cation 4.8.1.1.2.c.2, and that such interpretation does not degrade

safety, but does allow us the flexibility of testing and adjusting 4

ESF time delay relays independently of any other relay schemes.

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However, since Technical Specification interpretation is the main issue, a request to change the Technical Specifications will be generated by July 2,1980 to further clarify the subject specifi-cation.

Very truly yours, David C. Trimble Manager, Licensing DCT: MAS:nak

- cc: Mr. Victor Stello, Jr., Director

' Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 l

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