Text

Forwards Order Imposing Civil Monetary Penalty for Violation Noted in IE Insp Rept 50-269/82-11 on 820323-0401. Re-exam of Independent Verification Program Compliance w/NUREG-0737,Item I.C.6 Requested
	ML20027D730
Person / Time
Site:	Oconee
Issue date:	10/12/1982
From:	Deyoung R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:	Tucker H; DUKE POWER CO.
Shared Package
ML15223A833	List: IR 05000269/1982011; ML20027D730; ML20027E605;
References
	RTR-NUREG-0737, TASK-1.C.6, TASK-TM EA-82-065, NUDOCS 8211080311
	Download: ML20027D730 (2)
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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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Docket No. 50-269

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License No. DPR-38

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EA 82-65

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Duke Power Company

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ATTN: Mr. H. B. Tucker, Vice President

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Nuclear Production Department

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422 South Church Street

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Charlotte, NC 28242

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Gentlemen:

This refers to your letters of July 23, 1982 and September 15, 1982 in response

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to the Notice of Violation and Proposed Imposition of Civil Penalty sent to you

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with our letter of June 25, 1982. Our June 25 letter concerned a violation found

by our Resident Inspector on March 23, 1982, during a special inspection con-

ducted on March 23 - April 1,1982 of activities at the Oconee Nuclear Station,

Unit No. 1.

The circumstances are contained in Region II Inspection Report

No. 50-269/82-11.

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After careful consideration of your response, we have concluded 'that the civil

penalty as proposed is appropriate for the reasons given in the , enclosed Order.

However, with regard to the condition identified as Item 2 in the Notice, we

agree with your argument, based on tests you performed on July 1,1982, that the

initiating channel for the reactor building spray system was operable within

Technical Specification limits even with the cap missing from the instrument test

tee. Therefore, Item 2 in the Notice of Violation and Proposed Imposition of

Civil Penalty is hereby withdrawn.

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Our letter which transmitted the Notice of Violation and Proposed Imposition of

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Civil Penalty indicated that the penalty was being proposed to emphasize the need

to ensure that procedures affecting safe operation were meticulously followed.

We note that although your response addressed the conditions created by the

failure to follow an approved procedure and also discussed procedural revisions,

it did not describe corrective actions planned or taken to ensure that procedures

are meticulously followed in ~accordance with your Technical Specifications.

Therefore, please provide, within thirty days from the date of this letter, an

additional response to the Notice of Violation which includes the information

required by 10 CFR 2.201 relative to the failure to follow procedures.

Your attention is also invited to Paragraph D.2 of the Appendix to the Order

(Evaluations and Conclusions) which describes the implementation of the NRC

Confirmatory Order, dated July 10, 1981, directing Duke Power Company to take

certain actions described in NUREG-0737.

You are requested to reexamine your

program for independent verification of correct performance of operating activi-

ties to ensure that the required verifications are performed in accordance with

I.C.6 of NUREG-0737.

CERTIFIED MAIL

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RETURN RESE4PT DEQH M TED

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OCT 1 2 att

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Duke Power Company

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With regard to your concern about the issuance of a public announcement by the

NRC at the time civil penalties are proposed, that issue will be addressed in

separate correspondence.

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In accordance with Section 2.790 of the NRC's " Rules of Practice". Part 2,

Title 10, Code of Federal Regulations, a copy of this letter and the enclosures

will be placed in the NRC's Public Document Room.

The responses directed by this letter and the enclosures are not subject to the

clearance procedures of the Office of Management and Budget under the Paperwork.

Reduction Act of 1980 PL 96-511.

Sincerely.

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Ric ard C. DeYoung, Director

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Office of Inspection and Enforcement

Enclosure:

Order Imposing Civil Monetary

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Penalty

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J. E. Smith, Station Manager

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Distribution

Hon. Daniel R. McLeod

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PDR

NSIC

Attorney General

Robert C. Dennis Office Building

LPDR

ACRS

P. O. Box 11549

SECY

Columbia, SC 29211

CA

RCDeYoung, IE

Rudolph Mitchell, Chairman

JHSniezek, IE

Public Service Connission

JAxelrad, IE

111. Doctor's Circle

VStello, DED/ROGR

Columbia, SC 29203

JLieberman, ELD

JP0'Reilly, RII

HDenton, NRR

FIngram, PA

Director ES, RI, RII, RIII, RIV. RV

EA File

ES File

GBarber, IE

Resident Inspector, Oconee

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UNITED STATES

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NUCLEAR REGULATORY COMISSION

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In the Matter of

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Duke Power Company

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Docket No. 50-269

Oconee Nuclear Station

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License No. DPR-38

(Unit 1)

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EA 82-07

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ORDER IMPOSING CIVIL igDNETARY PENALTY

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Duke Power Company, 422 South. Church Street, Charlotte, North Carolina, 28242

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(the " licensee") is the holder of License No. DPR-38 (the " license") issued by

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the Nuclear Regulatory Commission (the " Commission").

The license authorizes

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operation of the Oconee Nuclear Station Unit 1 facility in Oconee Count'y, South

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Carolina under certain specified conditions and is due to expire on November 6,

2007.

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II

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An inspection of the licensee's activities under the license was conducted on

March 23 - April 1, 1982 at the OCONEE NUCLEAR STATION UNIT 1 facility in

Oconee County, South Carolina.

As a result of this inspection, it appears that

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the licensee has not conducted its activities in full compliance with the

conditions of its license. A written Notice of Violation and Proposed Imposition

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of Civil Penalty was served upon the licensee by letter dated June 23, 1982.

The Notice stated the nature of the violation, the provision of the license

condition which the licensee had violated, and the amount of civil penalty

imposed for the violation.

Answers dated July 23, 1982 and September 15, 1982

to the Notice of Violation and Proposed Imposition of Civil Penalty were received

frce the licensee.

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III

Upon consideration of the answers received and the statements of fact, explana-

tion, and arguments for remission or mitigation of the proposed civil penalty

contained therein, as set forth in the Appendix to this Order, the Director of

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the Office of nspection and Enforcement has determined that the penalty

proposed for the violation in the Notice of Violation and Proposed Imposition

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of Civil Penalty should be imposed.

The Director agrees with the licensee's

denial of the condition described as Item 2 in the violation in the Notice of

Violation and Proposed Imposition of Civil Penalty and withdraws that portion

of the violation dealing with the inoperabi,11ty of one of these channels of the

reactor building spray initiation system.

IV

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In view of the foregoing and pursuant to Section 234 of the Atomic Energy Act

of 1954, as amended (42 U.S.C. 2282, PL 96-295), and 10 CFR 2.205, IT IS HEREBY

ORDERED THAT:

The licensee pay a civil penalty in the amount of Forty-Four Thousand

Dollars within thirty days of the date of this Order, by check, draft, or

money order, payable to the Treasurer of the United States and mailed to

the Director of the Office of Inspection and Enforcement, U.S.NRC, Washington,

DC 20555.

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The licensee may within thirty days of the date of this Order request 'a hearing.

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A request for a hearing shall be addressed to the Director, Office of Inspection

and Enforcement.

If a hearing is requested, the Commission will issue an Order

designating the time and placi of heering.

Upon failure of the licensee to

request a hearing within thirty days of the date of this Order, the provisions

of this Order shall be effective without further proceedings; if pa,mnt has not

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been made t'y that time, the matter may be referred to the Attorney General for

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collection.

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VI

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In the event the licensee requests a hearing as provided above, the issues to

be considered at such hearing shall be:

(a) whether the licensee violated NRC license conditions as set forth in

the Notice of Violation and Proposed Imposition of Civil Penalty as

amended by Section III of this Order; and,

(b) whether, on the basis of such violation, this Order should bd

sustained.

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FOR THE NUCLEAR EGULATORY COMMISSION

rnwhA

chard C. DeYoung, Director

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ffice of Inspection and Enforcement

Dated at)Bethesda, Maryland

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this g{T day of October 1982

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. APPENDIX

EVALUATIONS AND CONCLUSIONS

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For the violation and associated civil penalty identified in the Notice of

Violation and Proposed Imposition of Civil Penalty for Duke Power Company's

Oconee station (Unit 1) dated June 23, 1982 the originial violation is restated

and the NRC's evaluation and conclusion regarding the licensee's responses (dated

July 23, 1982 and September 15,1982)'ispresented.

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ORIGINAL STATEMENT OF NONCOMPLIANCE

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Technical Specification 3.6.1 Tequins that containment integrity be maintained

whenever reactor coolant system (RCS) pressure is greater than 300 psig and

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temperature is greater than 200*F.

Technical Specification 3.5.1 requires that all three channels of both trains

of reactor building spray initiation be operable.when the reactor is critical.

. Technical Specification 6.4.1 requires that the plant be maintained in accor-

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dance with approved procedures.

Procedure IP/0/A/310/5D was established and

approved to implement 6.4.1.

Step 10.2.3 of the procedure requires replacement

of the cap on the 1/4-inch calibration line connected to the 1/2-inch sensing

line for reactor building pressure switch IPS-22.

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Contrary to the above, on July 9,1981, the licensee failed to follow step

10.2.3 of procedure IP/0/A/310/5D. As a result of the failure the following

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conditions existed between July 9,1%1 and March 23, 1982:

1.

Containment integrity of the Unit I reactor building was not maintained

for fifty-one days while RCS pressure was greater than 300 psig.and

temperature was greater than 200*F.

2.

For thirty-two days, one of three channels of Train A of reactor building

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spray initiation for Unit I was inoperable while the reactor was critical.

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EVALUATION AND CONCLUSIONS

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A.

Violation

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The licensee admitted its employees failed to follow required procedures

when calibrating reactor building pressure switch IPS-22 which is the

underlying violation for.which the civil penalty was proposed. The

licensee further"adnitted that containment integrity was not maintained.

However, the licensee denied that the reactor building spray initiation

channel was rendered inoperable by the missing cap.

Following receipt of the Notice of Violation and Proposed Imposition of

Civil Penalty, the licensee conducted a special test and determined that

pressure switch IPS-22 would actuate at approximately 22 psig which, while

greater than the nominal 10 psig setting, is within the Technical Specifi-

cation required value of 30 psig. Since the channel would operate within

che requirements of the Technical Specification, the NRC agrees that Item 2

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in the violation should be withdrawn.

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Appendix (Continued)

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8.

Assessment of Severity Level

The licensee argued that the violation should have been categorized at a

Severity Level IV because the potential increase in the offsite dose in

the event of an accident would have been negligible. 'While offsite dose

, consequences are a factor in determining the safety significance of a

violation, they are not the only factor.

In this case, the safety signi-

ficance lies primarily in the failure of the licensee's administrative and

management controls to ensure thai, procedures affecting safe operation

were meticulously followed for equipment important to safety which the

staff believes is cause for significant regulatory concern.

In the

present case the failure to follow procedures resulted in a degradation of

containment integrity, a violation of a limiting condition for operation

(LCO) and had the potential to preclude operation of a pressure switch in

the reactor building spray initiation system which would have violated yet

another LCO.

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The licensee argued that while Technical Specification 3.6.1 requiring

containment integrity was violated, the breach in containment would not

have resulted in a significant increase in the potential radiological

impact on the health and safety of the public at the site boundary in the

event o.f a design basis accident.

The NRC agrees.1 Nevertheless, in the

event of an accident, the breach in containment integrity could have resulted

in some additional release and this is of concern to the NRC because it

could have been avoided. Furthermore, the licensee did not address the

potential for increased exposure of plant personnel had entry into the

penetration room been required following an accident.

The NRC believes

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that such exposures could be significant.

In addition,lt was fortuitous

that both the pressure switch remained functional and the size of the contain-

ment broach restricted the potential radiological impact in the event of

an accioent.

Had failure to follow a procedure involved a larger contain-

ment penetration, the potential radiological consequences could have been

large and could have resulted in the violation being characterized as a

Severity Level II, in that the containment would not only have been degraded,

but would have been unable to perform its intended safety function.

Therefore, the staff has concluded that the violation was properly

categorized as a Severity Level III.

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1 While the staff agrees with the licensee's conclusion based on the calcula-

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tions performed, the licensee should have used the maximum hypothetical

accident as the basis for its analysis instead of the design basis loss of

coolant accident.

See Technical Information Document 14844, " Calculation of

Distance Factors fer Power and Test Reactor Sites."

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Appendix (Continued)

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C.

Assessment of the Civil Penalty

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Notwithstanding the withdrawal of the spray initiation channel operability

portion of the violation, the underlying procedural violation of Technical

Specification.6.4.1 remains significant.

Therefore, unless mitigation were

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,, appropriate, the staff would conclude that a civil penalty should be imposed.

D.

Mitigation Factors

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1.

Self-Identification

The licensee asserts that it identified the problem with its procedures

in January, 1982, that corrective action was taken at that time, and that

mitigation on that basis is required.

However, the need for independent

verfication had been previously identified by the NRC in NUREG-0585

and NUREG-0737, which were issued in November, 1979 and November, 1980,

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respectively, as a result of lessons learned from the Three Mile

Island accident.

Both recommended, among other things, that licensee's

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procedures "be reviewed and revised, as necessary, to assure an

effective system of verifying the correct performance of operating

activities is provided as a means of reducing human errors." Both

documents specifically referred to " human verification of operations

and maintenance independent of the people performing the activity"

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(Emphasis added).

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These provisions have been the subject of extensive correspondence

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over the past two years and of a Confirmatory Order issued on July 10,

1981. Thus, we do not believe any credit should be given to the

licensee for identifying the need for independent verification in

January, 1982.

2.

Corrective Action

The licensee claims that following its identificat.on of the potential

problem with failure of procedures to require ir,.ependent verification,

it took prompt and appropriate corrective actions to preclude repeti-

tion by changing its procedures.

Two points indicate otherwise.

First, the licensee provided, as a part of the response, a copy of a

memorandum from a site supervisor to his staff which required indepen-

dent verification by persons other than those doing the work. This

memorandum was limited in application to those supervised by the

author and thus did not precipitate or ensure generic corrective

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actions in other groups at the Oconee site.

Further, the instructions

were not provided in a controlled document within the meaning of

10 CFR 50, Appendix B, Criterion VI which would assure that future

employees would be informed of and understand the meaning of

" independent verification."

Second, it is noted that while Oconee procedures imply an independent

verification by the inclusion of two sign-off spaces on data sheets,

neither the body of the procedure nor any administrative control

explicitly establishes the meaning or significance of this entry.

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Appendix (Continued)

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Therefor., . do not believe that action taken was unusually prompt

or extensive and no mitigation based on corrective action is warranted.

3.

Enforcement History, Prior Notice and Multiple Examples

These factors were not used to increase the civil penalty above the

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base amount and the Policy does not provide for mitigation on the

basis of the absence of these factors.

Based on the above, the staff concludes that the civil penalty should not

be mitigated.

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DUKE POWER COMPANY

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July 23, 1982

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Mr. Richard C. DeYoung

Director, Office of Inspection and Enforcement

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U. S. Nuclear Regulatory Commission

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Washington, D. C.

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Re: Oconee Nuclear Station

Docket No. 50-269

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License No. DPR-38

Notice of Violation and Proposed Imposition of Civil Penalty

Dear Mr. DeYoung:

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Duke Power Company (Duke) hereby files its answer, in accordance with

10 CFR 2.205(b), to the " Notice of Violation and Proposed Imposition of Civil

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Penalty" issued in this docket by the NRC's Region II on June 25, 1982.

As

will be discussed in detail below, Duke does not believe that the incident

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which is the subject of the violation and the proposed civil penalty provides

the necessary basis for imposition of a civil penalty.

Therefore, Duke is

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protesting the , imposition of any civil penalty, and requests that, pursuant to

10 CFR 2.205, the NRC issue an order dismissing the proposed civil penalty.

The incident for which the civil penalty is proposed is discussed fully in

our response to the Notice of Violation, which is Attachment 1 to this letter.

That incident involved an apparent failure to replace an instrument cap from

the test tee for penetration WB-13.

While Duke's instrument calibration

procedures required replacement of that cap as a final step in returning the

instrument

to service,

the cap apparently was not

replaced foPowing

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instrument calibration.

On March 23, 1982, the NRC's Resident Inspector

discovered the missing instrument cap in a routine inspection of Oconee's Unit

1 Reactor Building.

It appears that the instrument cap may have been missing

from July 9,1981 until March 23, 1982.

The NRC concluded that containment

integrity was violated and the Reactor Building spray initiation system was

degraded during certain periods in the July

9,

1981 to March 23, 1982

interval. More specifically, two violations were alleged:

(1) " Containment integrity of the Unit I reactor building was not

maintained for fifty-one days while RCS pressure was greater than

300 psig and temperature was greater than 200 F."

(2) "For thirty-two days, one of three channels of Train A of reactor

building spray initiation for Unit 1 was inoperable while the reacgt4

was critical."

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As a result, the NRC Staff proposed a civil penalty in the amount of

Forty-four Thousand Dollars .($44,000) "[t]o emphasize the need for [ Duke] to

ensure that procedures affecting safe operation of the plant are meticulously

followed. "

Because of the apparent . duration of the event the base civil

penalty ($40,000) was increased by $4,000.

Duke believes that the proposed civil penalty should be withdrawn.

In

Duke's view there is no basis for imposing a penalty for this incident.

The

incident from which the alleged violations appear to have arisen occurred when

a step in the procedures for returning instruments to service after calibration

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was not followed.

Prior to the time that the missing instrument cap which is

the subject of this incident was discovered, Duke had identified the potential

. for a problem to occur when instrument calibration work was being conducted.

Therefore, Duke had amended its procedures to require that compliance with

those procedures be independently verified (by persons not from the crew

doing the work) when instruments are returned to service after being

calibrated.

Thus, to the extent that a deficiency in Duke's procedures led to

the incident, Duke had discovered that deficiency and had taken effective

corrective action to preclude an occurrence of the nature involved prior to

discovery of the missing instrument cap.

( Attachment 1, pp.1, 4.)

Moreover,

Duke

has

recently

completed

a

comprehensive

analysis

(Attachment 2 to this letter) of the potential significance of the incident.

That analysis shows that, with respect to the Reactor Building spray systems,

all three initiating channels were in fact operable within Technical Specification

limits even with the instrument cap missing from the instrument tee.

Thus

there was no violation of Technical Specification 3.5.1 relating to the Reactor

Building spray trains and, of course, there is no basis for a civil penalty in

this regard.

Moreover, even with the pressure cap missing, any doses to the

public which could have resulted from an accident would be less than the

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accident doses set out in the FSAR.

Therefore, the incident itself was not of

significance when measured 'against either the actual or potential impact on the

health and safety of the public.

Duke would like to emphasize, however, that even though the analyses

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show that there was no threat to the public health and safety as a result of

this incident, the Company both recognizes and appreciates the potential

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seriousness that might occur. from incidents of this nature. Duke believes that

its

record in identifying and correcting potential problem areas in its

operations is a good one, and will continue in the future to make every effort

to assure that it continues in that fashion.

The Notice states that the civil

penalty was proposed "to emphasize the need for [ Duke] to ensure that

procedures

affecting

safe

operation

of

the

plant

are

meticulously

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followed . . . "

Duke believes that the actions taken in regard to this incident

(both before and after discovery of the missing instrument cap) demonstrate

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the Company's concern with safe operation of the Oconee plant.

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Thus, Duke does not believe that a civil penalty is warranted.

In Duke's

view

the civil penalty

is

based

upon an incorrect application of the

Commission's Policy Statement to the incident and thus it should be withdrawn.

Moreover, other factors also warrant withdrawal of the proposed civil penalty.

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These factors include the fact that Duke itself identified as a potential problem

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in its procedures (prior to discovery of the missing cap) the lack of an

independent verification

for work of the nature done,

and had taken

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appropriate corrective action.

Though the change in procedures came too late

to prevent this incident, it will, of course, prevent the occurrence of similar

incidents in the future.

In addition, Duke's enforcement history at the Oconee

Station does not warrant imposition of a civil penalty.

Each of these points is discussed in greater detail below:

1.

The incident was of limited technical significance and posed no risk

to the public health and safety.

Moreover, D Ike's analysis shows that there

was no violation of Technical Specification 3.5.1.

First, as the analysis demonstrates conclusively, though there was in fact

a technical violation of Technical Specification 3.6.1,

there was no actual

impact to the public health and safety from the missing instrument cap.

So

far as any potential impact is concerned, the calculations performed show that,

under design basis accident conditions and the Reactor Building leakage that

existed at the time of the last Reactor Building leak rate test, even with the

instrument cap missing from the line, the doses resulting from the design basis

LOCA would be less than the doses calculated in the FSAR.

Thus, this

incident did not involve a significant potential impact to the public health and

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safety, as there would have been no discernible increase in the pctential

radiological consequences of postulated accidents.

Second, the analysis demonstrates that the incident did not lead to a

violation of Technical Specification 3.5.1,

in that it did not result in

inoperability of Channel 7 of the Reactor Building spray system.

To the

contrary, tests recently completed, described in Attachment 2, show that even

with the cap missing, the affected channel would have operated within the

Technical Specification limits if called upon.

2.

In light of this analysis, Duke believes that the proposed civil

penalty was based on an improper application of the Commission's Policy

Statement, in that the violatio

should be a Level IV violation rather than a .

Level III violation and thus a civil penalty is not warranted.

Application of

the guidance set forth in the Commission's Policy Statement to the specific

facts resulting from the analysis in Attachment 2 demonstrates this.

The fundamental criterion that the Commission intends be used in placing

a violation in a particular severity level is "the actual or potential impact on

the health and safety of the public." 47 F.R. 9968.

For Reactor Operations,

Severity Level III violations include one in which Technical Specification

Limiting Controls for Operation are exceeded, which results in loss of a safety

function, or one leading to a situation in which a system designed to prevent

or mitigate a serious safety event is not able to perform its intended

funcation .

10 CFR Part 2, Appendix C, Supplement I C.

Severity Level IV

violations

include

those

that

consist of a failure to meet

regulatory

requirements that have more than minor safety or environmental significance;

that is, violations of requirements which, if left uncorrected, could lead to a

more serious concern.

10 CFR Part 2, Appendix C, Section III; Supplement I

D.

Clearly , under application of these criteria,. this incident should not

result in a Level III violation.

As noted, one of the two violations cited by

the NRC resulting from this incident (relating to the Reactor Building spray

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system) turns out, upon analysis, not to be a violation.

With respect to the

violation relating to containment integrity, though there may have in fact been

a technical violation. of Technical Specification 3.6.1, in Duke's view this must

1

be weighed against the fact that, as shown in Attachment 2, such . violation

posed no threat to the public health and safety.

Therefore, the incident did

not involve exceeding a Technical Specification limit which resulted in loss of a

j

. safety function, nor did it -involve a situation in which a system designed to

t

mitigate consequences of or prevent a serious safety event was prevented from

performing its intended function.

.

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Indeed, Duke believes that the incident which led to the violations set

forth in, the Commission's June 25 Notice is a result of a problem in its

procedures which Duke had identified in January of 1982.

That is, Duke

i

determined that it was necessary to provide independent verification (by

persons not fro;; the crew doing the work) that an instrument is properly

l

returned to sert ice after being calibrated.

As Attachment 1 explains, upon

identifying that problem, Duke changed its procedures to provide for. such

independent verification.

This change in procedures .will preclude any such

occurrence in the future.

Nevertheless, when the missing instrument cap was

!

discovered in March, Duke reemphasized its change in procedures.

Thus, in

l

Duke's view, the problem with the procedures which led to the incident is no

L

more than a matter which has a "more than minor" safety significance.

That

is, if Duke had not found the potential problem with its procedures, it could

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have led to a more serious concern.

Consequently, Duke believes that the

incident should be classified as a Level IV violation, and no civil penalty

should be imposed.

i

Duke would like to add one thought with respect to the violation relating

'

to containment integrity.

It may be that the NRC has as a matter of policy

i

determined

that

any

breach of containment,

regardless of whether it

constitutes a potential impact to the public health and safety, is a Level III

,

l

violation and warrants a civil penalty.

If that is the case, the facts of this

matter clearly argue that the policy should be changed. Moreover, if the NRC

has made such a determination, it should announce that policy to the industry

I

and provide the technical basis for its judgment'.

3.

The proposed civil penalty should be withdrawn.

The Commission's

procedures provide that licensees, upon notification of a proposed civil

penalty, will have the opportunity to raise mitigating circumstances unique to

their particular cases.

These circumstances will be taken into account when

the decision is made whether or not to order the imposition of a civil penalty

for a particular incident. Licensees have been assured by the Commission that

"[m]itigation or remission of civil penalties based on such . . . responses is

.

not uncommon when compelling arguments are presented."

47 F.R. 9988.

In

Duke's view, complete remission of the proposed civil penalty is warranted in

this case.

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The primary reasons are, of course, the fact that the single violation

!

involving the missing cap constituted no risk to the public health and safety

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and that Duke itself identified the potential problem with its procedures which

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could lead to incidents such as the missing instrument cap and took corrective

[

action to preclude such incidents in 'the future.

However, in accordance with

the Notice, and the Commission's Policy Statement, Duke hereby addresses the

'

factors set out in Section IV B of 10 CFR Part 2, Appendix C.

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(i)

Prompt Identification and Reporting

As noted above, it is Duke's view that' the incident involving the missing

instrument cap which led to the violation resulted from the fact that the

i

instrument

calibration

procedures

did

not provide for an independent

verification

that the instrument had been properly returned to service

.l

following calibration.

As discussed more fully in Attachment 1, in January of

1982 Duke identified that potential problem and promptly took appropriate

corrective action.

This potential problem was not reportable to the NRC under

applicable requirements.

In light of the fact that the potential problem was

identified by Duke, the proposed civil penalty should be mitigated.

(ii)

Corrective Action to Prevent Recurrence

When Duke discovered the potential problem with its procedures, prompt

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corrective action was taken.

The corrective action is discussed fully in

Attachment 1, but, briefly, Duke changed its procedures in January to require

an independent verification (by persons not from the crew doing the work)

that equipment is returned to normal after being taken out of service.

If the

equipment is independently determined to be functional it must be documented

by the individual performing the verification.

Thus, prompt corrective action

!

was taken which will preclude occurrence of similar incidents in the future and

therefore the proposed civil penalty should be mitigated.

,

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(iii) Enforcement History

l

Duke believes that its enforcement history at Oconee is a factor which

should weigh in favor of mitigation of the proposed civil penalty.

In 1977,

Duke was assessed a civil penalty for an incident at Oconee.

A review of the

enforcement history since that time indicates that though violations have

occurred, none has been of a seriousness sufficient to warrant imposition of a

civil penalty.

Moreover, those violations are isolated in nature and not

recurring.

Because of this, it can be inferred that Duke's corrective actions

in response to these violations are effective, and that there is no serious

programmatic deficiency.

Thus, this factor should mitigate the proposed civil

penalty.

'

(iv)

Prior Notice of Similar Events

Duke, upon learning of potential inadequacies in its procedures, promptly

instituted corrective actions (see (ii) above) and thus this factor is not

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applicable.

(v)

Multiple Occurrences

This factor does not appear to be applicable .

There have been no

multiple examples of this incident.

As noted above, in Duke's view the subject incident at Oconee does not

warrant the imposition of any civil penalty. However, Duke wishes to comment

on what it believes is a fundamental unfairness in the way in which the

Commission's enforcement procedures, as set forth in its Policy Statement on

Enforcement, are carried out.

Duke makes these comments in a constructive

vein, recognizing that the Commission is monitoring activities under its Policy

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Statement on Enforcement, and has indicated its willingness to change the way

it, and its Staff, carries out activities under the statement if circumstances

warrant.

47 F.R. 9989.

To describe the process is to illustrate the problem.

Put briefly, when a

violation exists which might lead to an enforcement action, it is not uncommon

,

l

for an enforcement conference to be held.

At such a conference, the licensee

discusses the incident, including the sequence of events which led to it, the

actions tf ,, to mitigate the event, and the corrective actions which have been

and will ae taken to prevent recurrence' of similar events in the future.

Any

questions which the NRC Staff might have are discussed.

In short , the

,

enforcement conference serves as

a useful forum

for an exchange of

information and viewpoints between the licensee and the NRC Staff relating to

!

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a specific incident which might lead to a notice of violation and proposed

. imposition of a civil penalty.

'

Following the enforcen,ent conference, the NRC Staff then considers what

,

further action to take.

If in the Staff's view a violation exists, it will issue a

l

Notice of Violation and, if it believes it is warranted, a Notice of Proposed'

Civil Penalty.

Simultaneously with its issuance of the Notice of Proposed Civil

i

Penalty, the NRC Staff issues a press release announcing the violation and the

Notice of Proposed Civil P4nalty.

A press release is issued only if notice is

i

given of a proposed civil penalty.

The Commission views the publicity

(

attending a civil penalty as an important part of its enforcement procedures.

10 CFR Part 2, Appendix C, SIV. 47 F.R. 9990.

'

The Commission's procedures provide a mechanism for a licensee to

demonstrate why it believes a civil penalty is not warranted, or why one of a

lesser amount is justJied.

See, e.g. , 47 F.R. 9988; 10 CFR Part 2, Appendix

C, SIV B 1, 2.

However, only after the Notice of Proposed Civil Penalty and

the press release have been issued is a licensee permitted to make its case in

that regard.

Thus, to the extent factors exist which would eliminate or

mitigate

the proposed civil penalty,

licensees

are foreclosed

from

the

opportunity of being heard before the NRC takes action and the public

documents and press release which the NRC issues cannot reflect licensee's

,

!

position.

This problem is particularly acute when dealing with Level III

l

violations, where a significant amount of judgment is involved and a violation

does not always result in a notice of proposed civil penalty and attendant

press release.

In Duke's view, as a matter of basic fairness, licensees should be given

an opportunity to respond (as provided in the Policy Statement) to a proposed

civil penalty, putting forth the factors which it believes justify cancellation or

mitigation before public announcement of such a penalty is made.

In light of

the fact that the Commission views the press release mechanism as a punitive

measure, it should take particular pains to ensure that, at least in the case of

Level III violations, no public announcements are made of civil penalties until

it has considered carefully all factors involved and such a penalty is actually

assessed.

The instant matter perfectly illustrates the problems inherent in this

process .

An enforcement conference was held on May 21, 1982.

At that time,

the subject incident was discussed.

That discussion included the fact that

Duke had identified the potential problem with its procedures and had taken

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appropriate corrective action.

The incident of the missing instrument cap was

also discussed in detail.

Thus, following that conference, both the NRC Staff

and Duke had an understanding of the facts surrounding that incident.

Specifically, a full discussion was held on the potential problem with Duke's

procedures , the fact that Duke had identified that problem, the corrective

action taken, and the missing cap itself.

However, the analyses presented in

Attachment 2 had not been prepared and were not discussed.

The question of

a civil penalty was not discussed, beyond an indication by the NRC Staff that

the incident appeared to them to be a Level III violation which might warrant a

civil penalty.

On June 25, the NRC Staff issued its " Notice of Violation and Proposed

Imposition of Civil Penalty" and press release concerning the incident.

As a

result of that issuance, a substantial amount of publicity resulted, both local

and national.

However, because of the nature of the process, the NRC had

not--and could not have--considered any of the factors which Duke has raised

in this letter and its attachments which warrant withdrawal of the proposed

civil penalty.

And it naturally follows that, because of the timing of the

public announcement by the NRC, none of these factors could be reflected

therein, thus neither the notice nor the press release reflected in any way

Duke's position on the matter.

Of particular concern is the fact that, based

on the timing of the NRC's release, it could not clearly state that the incident

had no potential to affect adversely the public health and safety and therefore

could not put this incident in its proper perspective.

Therefore, in light of all these circumstances, Duke is compelled to

protest this procedure.

Duke believes that it has been made to suffer

unneeded and unjustified harm through adverse publicity caused by the

premature issuance of the Commission's press release, when the facts clearly

demonstrate no potential danger to the public health and safety existed and

that no civil penalty is warranted.

Duke would like to make a second comment.

The NRC, before imposing a

civil penalty for an incident at a facility, should take into account the

enforcement history of the licensee at that facility.

The Commission's Policy

Statement indicates that the enforcement history will be considered, but only

in the context of determining whether the base civil penalty to be assessed for

an incident should be increased.

10 CFR Part 2, Appendix C, Section IV B 3.

Duke believes that, when a licensee's enforcement history warrants, credit

should be given (in combination with other factors set out in Section IV B of

Appendix C) in considering whether a civil penalty should be either mitigated

or imposed at all.

In light of the above discussion, and the analyses and explanation set

forth in Attachments 1 and 2, Duke does not believe that imposition of any

civil penalty is warranted in this matter.

Therefore, Duke requests that the

NRC

issue

an

order

which

withdraws

the civil penalty

proposed

in

Mr. O'Reilley's letter of June 25, 1982, and further requests that the order be

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accompanied by a press release which announces that the NRC is withdrawing

its proposed civil penalty and explains the reasons for its withdrawal.

Very truly yours,

,

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<a . . rd ' . cj ~'-

William O. Parker, Jr.

g

I,

William

O.

Parker,

Jr., hereby affirm that I have read the foregoing

document and that it is true and correct to the. best of my knowledge and

behef.

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.

Mr. Richard C. DeYoung

July 23, 1982

Page 9

cc: Mr. James P. O'Reilly, Regional Administrator

U. S. Nuclear Regulatory Commission

Region II

101 Marietta Street, Suite 3100

Atlanta, Georgia 30303

Mr. W. T. Orders

NRC Resident Inspector

'

Oconee Nuclear Station

Mr. Philip C. Wagner

Office of Nuclear Reactor Regulation

U. S. Nuclear Regulatory Commission

Washington, D. C. 20555

.

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ATTACHMENT 1

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RESPONSE TO NOTICE OF VIOLATION

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on July 9, 1981 pressure switch IPS22, one of the three providing input to

Channel 7 of Engineered Safeguards which initiates the Reactor Building Spray

System was calibrated in accordance with Technical Specification 4.1.

Unit 1

was in the refueling mode and this test was scheduled for completion during the

1

outage. The instrument was required to be returned to service following calibra-

tion by the following Step 10.2.10 of Procedure IP/0/A/310/5D.

10.2.10

Decrease pressure to zero; disconnect pneumatic calibrator;

i

replace cap on tee; open isolation valve and return switch

'~

to service.

(See Attachment 4.)

'

The procedure also required redundant verification and usually involved two

l

persons from the same crew..

Apparently, the test cap on the k inch calibration line was left off and was

not discovered until March 23, 1982. The missing cap in the Penetration Room

could have allowed flow from the Reactor Building into the Penetration Room.

!

This constituted a violation of Technical Specification 3.6.1.

(See the response

to the Notice of Violation contained in this attachment.)

On January 1 and 2, 1982, two unit trips occurred as a result of certain secondary

1

side non-safety related instrumentation being inoperable because their isolation

,

valves were closed. The valves were not returned to the normal position after

'

testing, as a result of personnel error, which identified a need for a more

,

i

reliable method of verification of return to normal after work is completed.

The instrument procedure, at the time, required redundant verification of restora-

tion of the tested instrument to normal.

This involved two persons from the same

crew who performed the calibration and verified that the instruments'were properly

returned to normal.

On January 15, 1982, the Oconee I&E Engineer issued a letter

(Attachment 3) requiring independent verification, in an effort to ensure that for

all future tests, equipment would be returned to normal and proper documentation

would exist. This independent verification requires that an individual unconnected

with the crew performing the testing and calibration confirm that all the required

procedural steps are complete and that the system is properly returned to an

operable condition.

On March 23, 1982, the NRC Resident inspector discovered during an inspection tour

that a test tee cap to the instrument line for pressure switch 1PS22 was missing.

A cap was immediately installed and all units were checked for similar situations.

!

No other abnormalities were found. A verbal report was provided to the NRC that

l

day and Licensee Event Report number R0-269/82-08 was submitted on April 6, 1982

l

(revised July 23, 1982). On March 26, the Oconee Station Manager issued a letter

'

emphasizing the need for an improved return-to-normal system that would include

signing off each step of the test procedure required to return the in.strument to

normal.

Independent verification of these steps is also to be conducted in accor-

dance with the procedural changes accomplished in January. Refer to Attachment 5

for a revised version of IP/0/A/310/5D.

On May 21, 1982, Duke met with the NRC and discussed the incident and the improve-

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ments in the verification system initiated prior to and following the incident.

It was emphasized that procedural changes had been initiated prior to the incident

discovery in the area of independent verification.

In addition, Duke stated that

in its view, the health and safety of the public were not threatened. A detailed

evaluation of the potential impact on the health and safety of the public is con-

tained in Attachment 2 of this submittal.

The Notice of Violation states the reason for imposing this civil penalty on

Duke Power Company is "to emphasize the need for the licensee to ensure that

procedures affecting safe operation of the plant are meticulously followed..."

(

Duke considers that the actions taken prior to and following this incident demon-

strate that the company is vitally concerned with the safe operation of its facility;

thus, the proposed action by the NRC is not warranted to accomplish this stated

goal.

Following is the response to the Notice of Violation.

Violation

Technical Specification 3.6.1 requires that containment integrity be

maintainedwheneverreactorcoolantsystem(RCS)gressureisgreater

than 300 psig and temperature is greater than 200 F.

Technical Specification 3.5.1 requires that all three channels of

both trains of reactor building spray initiation be operable when

the reactor is critical.

Technical Specification 6.4.1 requires that the plant be maintained

in accordance with approved procedures. Procedure IP/0/A/310/5D was

!

established and approved to implement 6.4.1.

Step 10.2.10 of the

procedure requires replacement of the cap on the k inch calibration

i

line connected to the

inch sensing line for reactor building pressure

switch IPS-22.

Contrary to the above, on July 9, 1981, the licensee failed to follow

step 10.2.10 of procedure IP/0/A/310/5D. As a result of the failure

the following conditions existed between July 9, 1981 and March 23, 1982.

1.

Containment integrity of the Unit 1 reactor building was not

maintained for fifty-one days while RCS pressure was greater

than 300 psig and temperature was greater than 200 F.

2.

For thirty-two days, one of three channels of Train A of reactor

building spray initiation for Unit 1 was inoperable while the

reactor was critical.

This is a Severity Level III violation (Supplement I).

(Civil Penalty - $44,000)

Response

1.

Admission or denial of the alleged violation:

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Duke Power admits the violation outlined in Item 1; however, the company

does not agree with the severity level assigned to the violation nor the

imposition of a civil penalty. Technical bases justifying a lesser severity

level and a withdrawal of the proposed civil penalty are included in Attach-

ment 2.

Duke Power denies the violation contained in Item 2.

Technical

bases for denial are also included in Attachment 2.

2.-

Reasons for the violation:

I&E technicians performing the inctrument calibration apparently failed to

replace the calibration tee cap following completion of the calibration

procedure.

I

Prior to January 15, 1982, there was not a procedure which required inde-

pendent verification (by persons not from the crew doing the work) of return

to service of equipment by station I&E personnel.

3.

Corrective actions taken and results:

The immediate corrective action was to replace the missing cap for the test

i

tee and to check'all units for similar situations. No other abnormalities

were found.

All safety related Instrument Procedures have been revised to incorporate

a section for specifically identifying all equipment removed from service

as well as an appropriate independent verification of the equipment restora-

tion. The independent verification will require a second person (other than

those doing the work) to verify the specific items are restored to service

l

or verify operability by diverse means (Control Room indications, string

,

check, etc.).

l

4.

Corrective actions to be taken to avoid further violations:

All corrective actions have been taken as discussed in the preceding para-

graphs.

5.

Date when full compliance will be achieved:

Full compliance has been achieved.

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ATTACHMENT _2

TECHNICAL SIGNIFICANCE OF THE INCIDENT

The safety significance of an operation occurrence can be assessed in. terms

of_(1) actual impact on the health and safety of the public as a result of the

incident and (2) potential impact on the health and safety of the public via a

,'

significant degradation in the plant's design basis safety level.

Since this

incident did not involve any release in radioactive material, there was no

<

actual impact on the health and safety of the public.

.

EVALUATION OF POTENTIAL IMPACT ON THE HEALTH AND SAFETY OF THE PUBLIC

t

An operational occurrence might constitute an increased. risk to the health and

safety of the public if the incident resulted in a significant reduction in the

j

plant's design basis safety level. To determine whether a particular operational

occurrence constituted a significant reduction in the plant's design basis safety

i

level, the following questions must be answered:

(a) Did the event involve a significant adverse impact on the

l

radiological consequence of design basis accidents?

(b) Did the event result in a significant degradation of safety

,

systems designed to mitigate accidents?

(c) Did the event result in a significant increase in the like-

lihood of accidents?

The following evaluation addresses each of these questions with respect to the

specific incident under consideration.

Evaluation of Radiological Impact

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The reactor building is the major barrier designed to contain radioactivity

following accidents involving core damage and release into the containment. The

!

design basis criterion is to limit the RB leakage to within 0.25 weight percent /

day under the conditions of the maximum hypothetical accident as described in

I

Section 14 of the FSAR. During this incident a leakage path existed from the

'

RB to the penetration room through the " RB pressure sensing line and out

through the k" calibration tube with an effective flow path I.D. of 0.19".

i

Assuming the RB is at its design pressure of 59 psig continuously, the resulting

leakage into the penetration room would have been 0.36 weight percent / day. How-

ever, such an RB pressure behavior is rather impossible to be manifested during

the design basis accidents. Using a conservative envelope of the RB pressure

response for the limiting LOCA (59 psig for the first hour and 5 psig for the

,

subsequent 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />) as calculated in the FSAR (Figure 14-631), a leakage rate

!

of .093 weight percent / day is calculated to occur. This is considerably smaller

than the RB leakage limit of 0.25 weight percent / day. Furthermore, the most

i

recent (February 11, 1980) integrated leak rate test of the RB indicated a leakage

,

rate of 0.168 weight percent per day (normalized value at 59 psig from the test

I

pressure of 29.5 psig) with adequate margin to accomodate some unanticipated

leakage,

The most significant aspect of this incident is the potential impact of the leak

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path on the offsite accident doses. The penetration room is designed to filter

out the most significant species of radioisotopes (iodine and particulates) from

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the leakage out of the RB by passing through several filter assemblies prior

to discharge into the unit vent. In order to assess the impact of this incident

on the radiological consequences of potential accidents, it is necessary to

examine the contribution of the RB leakage through the penetration _ room to the

offsite doses following the design basis accidents. The FSAR offsite dose

analysis assumed 50 percent of the RB design basis leakage to pass through the

penetration room and the remaining 50 percent to bypass the penetration room.

Further, the RB leakage through the penetration room results in an elevated

I

release through the unit vent. Taking these features into account, the impact

of this incident on the radiological consequences of the design basis LOCA--the

limiting mechanistic accident in the RB--can be assessed as follows:

l -

(a) Ueing the FSAR assumption of 0.25 percent RB leakage and considering

the additional leakage of 0.36 percent during the first hour and

0.08 percent after the first hour, the 2-hour thyroid dose at the

exclusion distance in the event of a design basis LOCA would have

been 5 percent higher than the FSAR value.

Since the FSAR value is

only 1.5 percent of the 10 CFR 100 limit, tais increase represents

only a 0.077 percent of the 10 CFR 100 limit.

(b) Using the realistic value of the RB leakage as indicated by the RB

integrated leak rate test and considering the additional leakage of

0.36 during the first hour and 0.08 percent after the first hour,

the 2-hour thyroid dose at the exclusion distance in the event of a

design basis LOCA would have been less than the value calculated in

the FSAR.

(c)

Using the FSAR assumption of 0.25 percent leakage and considering

,

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the additional leakage of 0.36 percent during the first hour and

0.08 percent after the first hour, the 2-hour whole body dose at

the exclusion distance in the event of a design basis LOCA would

have been 40 percent higher than the FSAR value.

Since the FSAR

value for the 2-hcur LOCA whole body dose is only 0.04 percent of

the 10 CFR 100 limit, this increase represents only a 0.016 percent

of the 10 CFR 100 limit.

(d) Using the realistic value of the RB leakage as indicated by the RB

l

integrated leak rate test and considering the additional leakage of

0.36 during the first hour and 0.08 percent after the first hour, the

2-hour whole body dose at the exclusion distance in the event of a

design basis LOCA would have been less than the value calculated in

the FSAR.

Therefore, this incident did not involve a significant impact on the radiological

consequences of potential accidents, if they occurred.

l

Evaluation of the Impact on Safety Systems

The subject incident affected the performance of ~one of three independent channels

of one of the two redundant RB spray trains. It did not affect the performance or

operation of any other instrumentation or safety system. The RB spray system is a

safety system designed to limit the RB pressure to within the design limit during

accidents involving blowdown of mass and energy into the RB (LOCA and accidents

involving secondary system break within th.e RB).

It is also considered to be use-

ful in scrubbing the radioactivity from the RB atmosphere during the post-accident

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Phase of a LOCA. With respect to limiting the RB pressure, the RB cooling system,

which is independent of the spray system, is fully capable of limiting the RB

pressure to within the design limit, irrespective of the availability of the spray

system.

Each of the two redundant RB spray trains is automatically actuated upon tripping

of two of three redundant channels. The nominal trip setpoint for.each channel

is 10 psig with a required trip actuation (as required by Tech Spec and safety

analysis assumptions) of 30 psig. The subject incident caused Pressure Switch

IPS-22 not to trip at the nominal setpoint of 10 psig. However, it has now been

determined that the channel would have indeed tripped at a pressure of approximately

22 psig, well within the required setpoint of 30 psig. This determination is

based on recent tests conducted on the same channel of a similar unit simulating

i

the as-found condition and applying a gradually increasing pressure on the sensing

line. During this test the channel repeatedly tripped at approximately 22 psig;

therefore, this incident did not degrade _the RB spray system nor did it signifi-

cantly degrade the function of the channel.

It is also pointed out that even if

,

the channel were inoperable, the RB spray system can still accommodate a single

j

failure and achieve system design function.

Evaluation of Impact on the Likelihood of Accidents

i

This incident did not, in any manner, affect the initiation of or occurrence of

,

any accident.

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CONCLUSION

The foregoing evaluation has demonstrated that this incident did not have any

actual impact on the health and safety of the public. It also did not have any

significant potential impact on the health and safety of the public by virtue

of the fact that (1) it did not result in any increase in the likelihood of

accidents, (2) it did not result in any significant degradation of safety systems

!

designed to mitigate accidents, and (3) it did not involve a significant adverse

impact on the potential radiological consequences of design basis accidents.

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AttschmInt 3

Jcauary 15, 1982

.

INIRASTATION LETTER

OCONEE NUCLEAR STATION

To:

I&E Coordinators & Supervisors

SUBJECT:

Oconee Nuclear Station

Independent Verification

As you are aware, we experienced two (2) instances of instrumentation

being valved out during start-up after Unit 1 refueling outage.

We

experienced similar problems af ter Unic 3 refueling outage.

The pro-

blems on Unit 1 caused three (3) unit trips.

While these instances

represent only a minute fraction of the work performed they do indicate

a trend and we feel we should make adjustments to our practices in an

attempt to reduce this type of incident.

In the future we will require an independent (different from person (s)

doing work) verification that equipment is returned to normal if it can

not be confirmed by diverse means.

(Control room indication, string

checks, etc.)

This will require independent verification of all instru-

mentation connected to system piping when the system is not in service

and in some cases when the system is in service.

If equipment is deter-

mined to be functional by diverse means it shall be so documented by the

individual (s) performing the work.

Please insure that this requirement is presented to and understood by all

of your employees.

We realize that these changes will be costly in man-

power but believe them necessary to insure proper equipment status for

safe efficient operation of the plant.

We appreciate the effort and performance of everyone in the section and

hope that these changes will enable us to do an even better job in the

future.

l

M_M

/ R. C. Adams

I&E Engineer

RCA/pc

l

cc:

J. E. Smith

.

Gerald Vaughn

'

Joe Davis

J

.

-

-.

-

- -

_ .

Attzchm nt 4

'C

MASTER FILE

P _ SPD-10a-1

.

DUKE POWER COMPANY

(1)

ID No:IP/0/A/310/5D

PROCEDURE PREPARATION

Change (s)

3

to

PROCESS RECORD

1

Incorporated

(2)

STATION: Oconee

I

"

*E**

I* *

E

(3)

PROCEDURE TITLE:

R.B. Pressure Switch Calibration and Pressure Switch Contact Buffer Tests

(4)

PREPARED BY:

DATE:

[ "/h M/

-

- .i.f w

h k DATE: [- /[" [/

l

(5)

REVIEWED BY:

,-

---

Cross-Disciplinary Reviev'By:

N/R:

1

(6)

TEMPORARY APPROVAL (IF NECESSARY):

By:

(SRO)

Date:

By:

Date:

(7)

APPROVED BY:

% hwM

Date:

h/#/4/

(8)

MISCELLANEOUS:

Revi w d/4,. . . d By:#f(4ff ,

[ Date:_ f/jr g,/#/

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Reviewed / Approved By:

Date:

.

, , _ _ , . _ ,

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FORM SPD-1001-2

,

. . .

NUCLEAR SAFETY EVALUATION CHECK LIST

Oconee

(1) STATION:

UNIT: 1

2

3

OTHER:

(2) CHECK LIST APPLICABLE TO: IP/0/A/310/5D

O) SAFETY EVALUATION - PART A

The item to which this evaluation is applicable represents:

Yes

No _J__ A change to the station or procedures as described in the FSAI

or a test or experiment not described in the FSART

j

I

l

If the answer to the above is "Yes", attach a detailed description of the ites

being evaluated and an identification of the affected section(s) of the FSAR.

(4) SAFETY EVALUATION - PART 3

,

-

Yes

No

I

Will this item require a change to the station Technical

Specifications?

If the answer to the above is "Yes," identify the specification (s) affected

and/or attach the applicable pages(s) with the change (s) indicated.

i

(5) SAFETT EVALUATION - PART C

!

As a result of the item to which this evaluation is applicable:

Yes

No

I

Will the probability of an accident previously evaluated

in the FSAR be increased?

)

Yes

No

I

Will the consequences of an accident previously evaluated

in the FSAR be increased?

,

'

Yes

No

I

May the possibility of an accident which is different

than any already evaluated in the FSAR be created?

l

. Yes

No

I

Will the probability of a malfunction of equipment

!

1mportant to safety previously evaluated in the FSAR

'

be increased?

Yes

No

I

Will the consequences of a malfunction of equipment

important to safety previously evaluated in the FSAR

be increased?

Yes-

No, I

May the possibility of malfunction of equipment

important to safety different than any already evaluated

in the FSAR be created?

Yes

No

I

Will the margin of safety as defined in the bases to any

Technical Specification be reduced?

If the answer to any of the precading is "Yes", an unreviewed safety

question is involved.

Justify the conclusion that an unreviewed safety

question is or is not inv ved.

Attach additional pages as necessary.

I ~ l ) Of

C

(6). PREPARED BY:

-

DATE:

e'

DATE:

8"/9 N

(7) REVIEWED BY:

- ,

,

(8) Page 1 of

_

(

..

.

s

.

,

Form SPD-1002-1

DUKE POWER COMPANY

(1)

ID No: IP/0/A/310/5D

PROCEDURE PREPARATION

Change (s) 3

to

PROCESS RECORD

3

Incorporated

l

(2)

STATION:

Oconee

(3)

PROCEDURE TITLE: Ennineered Safestuards System Analog Channel C

R.B. Pressure Switch Calibration and Pressure Switch

-

. ..

-

.

(4)

PREPARED BY:

DATE:

3 ~~/3-8 /

(5)

REVIEWED BY:

i

x/

-

DATE: M/9 O

Cross-Disciplinary Review By:

N/R:

1

(6)

TEMPORARY APPROVAL (IF NECESSARY):

By:

(SRO)

Date:

1

By:

Date:

(7) APPROVED BY:

AM

MA

Date:

N/

v-

~-

~

<

(8) MISCELLANEOUS:

.

d/?r; : :d Byg///JNs

[ Date: S /24/8/

Ravi

~

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Reviewed / Approved By:

Date:

,

.

>

.

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--

- - -

-

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..

.

,

DUKE POWER COMPANY

(1) ID No.IP/0/A/310/5D

.

COMPLETED PROCEDURE

PROCESS RECORD

t

I

(2)

STATION:

oconee

Engineered Safeguards System Analog Channel *Cl R.B. Pressure

(3)

PROCEDURE TI M .

Switch Calibration and Pressure Switch Contact Buffer Tests

.

(4)

DATE(S) PERFORMED:

-

.

(5)

PROCZDURI COMPL.uGN VI327ICATION:

TIS

N/A

Check lists and/or blanks properly initialed, signed,

dated. or filled in N/A or N/R, as appropriace?

TES

N/A

Listed enclosures attached?

YES

N/A

Data sheets attached, completed, dated and signed?

TES

N/A

Charts, graphs, etc. attached and properly dated,

identified and marked?

,

TIS.

N/A

Accepeance criteria met?

Verified By:

DAM:

,

.

(6)

PROCEDURE COMPLETION APPROVED:

DATE:

l

(7)

RINAREM:

l

.

,

-

1

.

(8) Page 1 of

-

__

_

,

_ _ _

..

-

_. .

.

IP/0/A/310/5D

DUKE POWER COMPANY

OCONEE NUCLEAR STATION

ENGINEERED SAFEGUARDS SYSTEM

ANALOG CHANNEL C

l

R.B. PRESSURE SWITCH CALIBRATION

AND PRESSURE SWITCH CONTACT BUFFER TESTS

1.0 Purpose

'

1.1 To furnish a procedure for calibration of Reactor Building

pressure switch instrumentation.

1.2 To calibrate digital computer inputs.

1.3 To perform the functional and operational tests.

2.0 References (Use Current Copy)

l

2.1 Vol. 1 & 3 BMC. NI & RPS & ESS inst.

.

l

2.2 881 System Checkout Procedure

2.3 FSAR, Section 7.1.3.

2.4 Technical Specifications, Section 4.1.1.

3.0 Test Equipment Required

.

3.1 Pneumatic Calibrator, W/T series FA 145 Range 0-30 psi, or

equivalent.

4.0 Prerequisites Sign-off(s) on Enclosure 11.1

.4.1

Supervisor has reviewed and initialed all portions of this

procedure which are not applicable to the activity being

performed.

The supervisor's review is not required if the

procedure specifies sections to be omitted except as needed

for abnormal conditions.

4.2 Verify all changes on the control copy are incorporated on the

j

working copy.

4.3 Computer in operation.

.

-

--

-

-2-

.

4.4 Verify that other ES Channels are not in trip conditions.

5.0 Limits and Precautions

Use proper precautions while working with components that have HIGH

VOLTAGE, HIGH PRESSURE or HIGH TEMPERATURE present.

6.0 Unit Status Sign-off(s) on Enclosure 11.1

6.1 N/A

7.0 General Description

'

This instrumentation is used to sense high building pressure which

,

actuates the ES Building Spray System.

8.0 Major Components

Component

Description

Reference

BS4-PSS (PS-22)

Mercoid Snap Switch Control

Bailey Ref.

BS4-PS6 (PS-23)

Mercoid Snap Switch Control

Bailey Ref.

I

'

i

9.0 Equipment Specifications

9.1 Operating Range

Instrumentation Designation

Input

Output

BS4-PS5 (PS-22)

10 psig

Close

BS4-PS6 (PS-23)

10 psig

Close

l

9.2 Computer Inputs

.

From

BS4-PS5 (PS-22)

D1957

BS4-PS6 (PS-23)

D1958

10.0 Procedure

CAUTION:

If any component calibration is out of tolerance by 2%,

proceed to Maximum Tolerance Exceeded Sheet.

10.1 Pressure Switch Contact Buffer Tests

In this case there is a pressure switch (N/0 Contact) applied

to each relay circuit of the contact buffer.

Therefore, in

normal operation both lamps will be "0N".

.

.-..--....-,--.....,-r---,_..m-

,..,---,y,,,-ne--,w...,w--w,.r,w,mm,,www.,p-

m

.w

e s.

-

..

.

'

-3-

-

.-

10.1.1

Depress S1 (Switch at top).

DS1 should go "0FF"

(lamp at top).

10.1.2

While holding S1, check auxiliary relay for proper

indication (DS1 and DS2 on auxiliary relay will be

bright) .

The building spray trip lamp on the analog

indicating panel should be bright.

10.1.3

Release SI..

All of the lamps in steps 1 and 2

should return to their initial states.

Check computer

input D-1957 for proper indication.

Also check for

proper statalarm indication.

10.1.4

Depress S2 (Switch at bottom).

DS2 should go "0FF".

10.1.5

While holding S2, check auxiliary relay for proper

indication (DS3 and DS4 on auxiliary relay will be

bright).

The building spray trip lamp should once

again be bright.

10.1.6

Release S2.

All of the lamps mentioned in Steps 4

and 5 should return to their initial states.

Check

compute'r input D-1958 for proper indication.

Also

check.for proper statalarm indication.

10.2 Pressure Switch Setting

Complete Enclosures 11.2a and 11.2b using the following procedure

to calibrate:

10.2.1

Close isolation valve and connect Pneumatic Calibrator

to Tee.

l

10.2.2

Slowly increase pressure until respective contact

buffer light goes "out".

Record pressure on data

i

,

sheet under "As Found".

,

10.2.3

If switch makes within the pressure tolerance specified,

decrease pressure to zero, disconnect pneumatic

calibrator, put cap on Tee, open valve, and return

switch to service.

I

10.2.4

If setting is in error or needs to be changed,

continue with Step 5.

10.2.5

Remove cover from switch to expose the switch adjust-

ment.

10.2.6

Set input pressure to that value listed on data

I

sheet at which you want the switch to make.

10.2.7

Adjust switch setpoint until switch makes.

..

.

. - _ - _

..

- . .

. .

-

. . . .

.

-

1

-

l

-4-

-

,

.

,

i

.

l

10.2.8

If switch makes on pressure increase, reduce the pressure

until switch resets.

Slowly increase pressure until

switch just makes.

If it makes at desired setpoint,

record data on data sheet under "AS LEFT".

If still not.

.

within tolerance specified, repeat Steps 5, 6, and 7,

I

until it is within tolerance specified.

10.2.9

Replace cover on switch and recheck calibration to

insure switch is still within tolerance.

10.2.10

Decrease pressure to zero; disconnect pneumatic

calibrator; replace cap on tee; open isolation

valve and return switch to service.

10.2.11

Insure that Building Pressure Analog Channel C is reset

and not in the tripped state.

.

11.0 Enclosures

11.1 Sign-Off Sheet

11.2 Calibration Data Sheets

11.2.a

PS- 22 .( Bs4*-PS 5)

11.2.b

PS-23 (BS4-PS 6)

11.3 Tolerance Limit Exceeded Sheet

.

9

.

p

-p.

, , .

-r--

mm--+.----

-

- . _ _ _ _

e.

.-.y-

. . - ,

,

-

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. _ ..__ _ _.

,

.

,

.

,

.

,

.

i

<

.

ENCLOSURE 11.1

.

t

.,

Sign-off(s)

Unit Status

Data Began

l

Prerequisites

6.1

Data Completed

i

4.1

-

WRf

<

4

'

4.2

Unit _

'

i

4.3

i

r

!

4.4

-

,

Data Sheets

.

11.2.a

-

,

i

-

11.2.b

!

PERFORMED BY

REMARKS

.

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imaw w e-

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+-

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.

DUEE POWER C0tfPANY

OCONEE NUCLEAR STATION

l UNIT #

l

ENCLOSURE 11.2.a

l

CALIBRATION DATA SEET

IP/0/A/310/5D

l

ITEM Pressure Switch

TEST EQUIPMENT USED

MFG.

turcoid

ITEM

SN

TYPE

APW-7041-153

__

.

Satpoint

Tolerance

.5 % of Span

t .095 PSIG

_

SYSTEM ES Building Spray

'

INSTRUMENT NO.

PS-22 (BS4-?SS)

PRODUCT INSTRUCTION

SPAN

19 PSIG

Switch to close on pressure increase.

SWITCH LOC.

.

!

Input

Desired

As

As Left

Error

to

Actuation

Found

Setpoint

PSI

Switch, Point (PSIG)

PSI

i

10

All isolation valves left open - Verified By

,

MAXIMUM ERROR IN PSI

PERFORMED BY

DATE

>

_.

_

. _ _ _ . - _

_

__

_.

__

,

.-

..

.

DUKE POWER COMPANY

!

OC01E.E NUCLEAR STATION

UNIT #

4

.

ENCLOSURE 11.2.b

l'

CALIBRATION DATA SHEET

IP/0/A/310/SD

'

ITEM Pressure Switch

TEST EQUIPMENT USED

' MFG.

Mercoid

ITEM

SN

TYPE

APW-7041-153

l

Satpoint

Tolerance

t .5 % of Span

1 095 PSIG

SYSTEM ES Building Spray

INSTRUMENT NO.

PS-23 (BS4-PS6)

PRODUCT INSTRUCTION

i

SPAN

19 PSIG

Switch to close on pressure increase.

SWITCH LOC.

'

l

.

Input

Desired

As

As Left

Error

to

Actuation

Found

Setpoint

PSI

Switch

Point (PSIG)

PSI

PSIG

10

All icolation valves left open - Verified By

_,

MAXIMUM ERROR IN PSI

PERFORMED BY

DATE

'

i

- , - - - . - , - . . - . . . , - - . . . , . . . - . - .

- - , - -

--

.

- --

- - - -

. ,._. - -.. - -

-

- ._ _

.

..

.

ENCLOSURE 11.3

IP/0/A/310/5D

MAXIMUM TOLERANCE LIMIT EXCEEDED SHEET

Initici & Date

(A).

Notified Instrument Supervisor that a tolerance

Tech

of 2% was exceeded on the following components:

l

,

.

f

l

Initiel & Date

(B).

An evaluation was made on the above problem (s)

Inst. Sup.

and the following corrective action was taken:

.

f

-

I

,-

, _ , . . _

. . . - .

. - - . _ - - .

.-

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.

,

.

MAST 2.R m.r__.p,

Form SPD-1002-1

DUKE POWER COMPANY

(1)

ID No:_ IP/0/A/310/5D

PROCEDURE PREPARATION

PROCESS RECORD

Change (s) V

to

_ If

Incorporated

(2)

STATION :

n,.,,,,..

(3)

PROCEDURE TITLE: Engineered Safeguards System Analog Channel C R.B.

Pressure Switch Calibration and Pressure Switch Contact Buffer Tests

(4)

PREPARED BY:

,

DATE:

Y - B 'O L

(5)

REVIEWED BY:

DATE:

/2. [d

Cross-Disciplinary Review By:

N/R:

(6)

TEMPORARY APPROVAL (IF NECESSARY):

!

By:-

(SRO)

Date:

l

By:

Date:

(7) APPROVED BY:

I .e b

f

Date:

Y~24'8/

i

(8)

MISCELLANEOUS:

Revieve

pp;;.;d Sy:M MrM

Date:

.J/ M

~

Reviewed / Approved By:

Date:

i

'

!

l

,

l

t

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.

4

Lw,

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FORM SPD-1001-2

DUKE POWER COMPANY

NUCLEAR SAFETY EVALUATION CHECK LIST

,

(1) STATION:

Ocones

UNIT: 1

_2

3

OTHER:

(2) CHECK LIST APPLICABLE TO:

IP/0/A/310/Su

(3) SAFETY EVALUATION - PART A

The item to which this evaluation is applicable represents:

Yes

No

A change to the station or procedures as described in the FSAR;

!

or a test or experiment not described in the FSAR?

If the answer to the above is "Yes", attach a detailed descriptien of the item

being evaluated and an identification of the affected section(s) of the FSAR.

(4) SAFETY EVALUATION ,- PART B

Yes

No

Will this item require a change to the station Technical

Specifications?

If the answer to the abov'e is "Yes," identify the specitication(s) affected

and/or attach the applicable pages(s) with the change (s) indicated.

(5) SAFETY EVALUATION - PART C

.

As a result of the item to which this evaluation is applicable:

/

l

Yes

No

Will the probability of an accident previously evaluated

in the FSAR be increased?

Yes

No M ill the consequences of an accident previously evaluated

Jn the FSAR be increased?

Yes

No

/May the possibility of an accident which is dif ferent

than any already evaluated in the FSAR be created?

Yes

No g ill the probability of a malfunction of equipment

important to safety previously evaluated in the FSAR

be increased?

No M ll the consequences of a malfunction of equipment

Yes

important to safety previously evaluated in the FSAR

increased?

l

Yes

No

May the possibility of malfunction of equipment

important to safety different than any already evaluated

i the FSAR be created?

i

Yes

No

ill the margin of safety as defined in the bases to any

.

Technical Specification be reduced?

l

'

If the answer to any of the preceding is "Yes", an unreviewed safety

question is involved. Ju tify the conclusion that an-unreviewed saf ety

question is or is nor in'o ved.

Attach additional pages as necessary.

wC U

j

(6) PREPARED BY:

_

CU

DATE:

/2/ M

(7) REVIEWED BY:

. [--

DATE:

/7

7 t.

-N

/

(8) Page 1 of

'

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_

,

,,__

_

_ _...

i

_

-

_

'

.

IP/0/A/310/5D

DUKE POWER COMPANY

,

u.

OCONEE NUEEE.:IIATION

-

ENGINEERED SAFEGUARDS SYSTEM

ANALOG CHANNEL C

~

R.B. PRESSURE SWITCH CALIBRATION

AND PRFSSURE SWITCH CONTACT BUFFER TESTS

.

1.0 Purpose

1.1 To furnish a procedure for calibration of React.or Building

pressure switch instrumentation.

1.2 To calibrate digital computer inputs.

1.3 To perform the functional and operational tests.

2.0 References (Use Current Copy)

2.1 Vol. 1 & 3 BMC..NI & RPS & ESS inst.

2.2 881 System Checkout Procedure

2.3 FSAR, Section 7.1.3.

2.4-

Technical Specifications -Section 4.1.1.

3.0 Test Equipment Required

3.1

Pneumatic Calibrator, W/T series FA 145 Range 0-30 psi, or

equivalent.

4.0

Prerequisites Sign-off(s) on Enclosure 11.1

-4 .1

Supervisor has reviewed and initialed all portions of this

procedure which are not applicable to the activity being

performed.

The supervisor's review is not required if tne

procedure specifies . Sections to be omitted except as needed

for ^ '-J .rnad ' .i ::3.

4.2

Verify all change = ou the control copy are incorporated on the

working copy.

4.3

Computer in operation.

.

.---,w,

--

.,

--

-

.

'

.

2-

.

-

,

.

4.4

Verify that other ES Channels are not in trip conditions.

5.0 Limits and Precautions

Use proper precautions while working with components that have HIGH

FOLTAGE, HIE _EiiESSURE or HIGH TEMPEF.L'"URE present.

6.0 Unit Status Sign-off(s) on Enclosure 11.1

_

6.1 N/A

7.0 General Description

This instrumentation is used to sense high building pressure which

actuates the ES Building Spray System.

8.0 Major Components

Component

Description

Reference

BS4-PSS (PS-22)

Mercoid Snap Switch Control

Bailey Ref.

BS4-PS6 (PS-23)

Mercoid Snap Switch Control

Bailey Ref.

9.0 Equipment Specifications

9.1

Operating Range

Instrumentation Designation

Input

Output

BS4-PSS (PS-22)

10 psig

Close

BS4-PS6 (PS-23)

10 psig

Close

.

9.2 Computer Inputs

From

BS4-PSS (PS-22)

D1957

BS4-PS6 (PS-23)

D1958

10.0 Procedure

CAUTION:

If any component calibration is out of tolerance by 2%,

proceed to Maximum Tole =nsce 5xceeded Sheet.

.

10.1 ?wi.z :.~ .S.::'-" Cantact ?=~is: ~ 7* .

In this case there is a pressure switch (N/0 Contact) applied

to each relay circuit of the contact buffer.

Therefore, in

normal operation both lamps will be "ON".

,

-

, , - -

, - -

-w

- ,- .-

c.,

>-g

y,

,ge. ,, ar-o.e r-es spae w a mw e r me*,.w-- e e m eew_e ssron

e-a-.= sew-.-

r-

"

-

-4

- -- --

.

'

-3-

.

10.1.1

Depress S1 (Switch at top).

DS1 should go "0FF"

(lamp at top).

10.1.2

While holding S1, check auxiliary relay for proper

l

indication (DS1 and DS2 on auxiliary relay will be

% 6:Eght).

The bui?mE=s spray trip lamp on the analog

indicating panel should be bright.

,

10.1.3

Release S1.

All of the lamps in steps 1 and 2

should return to their initial states.

Check computer

input D-1957 for proper indication.

Also check for

l

proper statalarm indication.

10.1.4

Depress S2 (Switch at bottom).

DS2 should go "0FF".

.

10.1.5

While holding S2, check auxiliary relay for proper

!

indication (DS3 and DS4 on auxiliary relay will be

bright).

The building spray trip lamp should once

again be bright.

10.1.6

Release S2.

All of the lamps mentioned in Steps 4

i

and 5 should return to their initial states.

Check

!

computer input D-1958 for proper indication.

Also

check for proper statalarm indication.

l

10.2 Pressure Switch Setting

[

Complete Enclosures 11.2a and 11.2b using the following procedure

to calibrate:

l

10.2.1

Close isolation valve, remove test tee cap, and connect

Pneumatic Calibrator to Tee.

Sign off on data sheet.

10.2.2

Slowly increase pressure until respective contact

buffer light goes "out".

Record pressure on data

sheet under "As.Jound".

NOTE:

DS1 is for PS-22 and DS2 is for PS-23.

I

10.2.3

If switch makes within the pressure tolerance specified,

I

decrease pressure to zero, disconnect pneumatic

calibrator, put cap on Tee, open valve, and return

(

switch to service.

10.2.4

If setting is in error or needs to be changed,

.

continue with St.rces..

1

i

1_ J.P:.%:e:toww i_u_ s._=:n to expose the switch adjust-

ment.

10.2.6

Set input pressure to that value listed on data

sheet at which you want the switch to make.

10.2.7

Adjust switch setpoint until switch makes.

-

_- _ _ .

-_.

.

_ . . . _ . , . . _

_,. .. ...._,. _ _ .._ _ .

-_

-

.

4-

-

.

,

10.2.8

If switch makes on pressure increase, reduce the pressure

until switch resets.

Slowly increase pressure until

switch just makes.

If it makes at desired setpoint,

record data on data sheet under "AS LEFT".

If still not

within tolerance specified, repeat Steps 5, 6, and 7,

> until it is wi thin ."x.lerance specified.

.. .

'

10.2.9

Replace cover on switch and recheck calibration to

insure switch is still within tolerance.

10.2.10

Decrease pressure to zero; disconnect pneumatic

calibrator; replace cap on tee; open isolation

valve and return switch to service.

Sign off on datt

sheet.

.

10.2.11

Insure that Building Pressure Analog Channel C is reset

,

and not in the tripped state after both pressure switches

are complete.

11.0 Enclosures

11.1 Sign-Off Sheet

,

!

11.2 Calibration Data Sheets

11.2.a

PS-22 (BS4-PSS)

11.2.b

PS-23 (BS4-PS6)

11.3 Tolerance Limit Exceeded Sheet

!

.

O

e

%*

. .

-

!

. _ . . -

_ _ . _ . _ _ . . _

. _ _ . _ _ . _ . . _ . _ . . . . . . . . . . . - . . , . . . . . , _ . ,

_.

. __.

-. . . . . . . _

. . . - -

.

.. _

_

- _ _ . _ . _

.

ENCLOSURE 11.1

-

Sign-off(s)

Unit Status

Date Began

i

Prerequisites

6.1

Date Completed

-

,

4.t

WR#

,

I

'

4.2

Unit _ ,

4.3

4.4

(

Data Sheets

l

11.2.a

,

'

11.2.b

PERFORMED BY

b'

REMARKS

i

.

l

!

-

O

J

, . . - . - -

, - - -

, - . _

. - _ _ -

- , - . - - - - -

. , _ - _

_

. - -

-

_ _ _ .

- - - .

.

'

.

DUKE POWER COMPANY

OCONEE NUCLEAR STATION

I! NIT d

.

ENCLOSURE 11.2.a

CALIBRATION DATA SHEET

IP/0/A/310/5D

s

ITEM Pressure Switch

TEST EQUIPMENT USED

tGC, .

Mercoid

ITEM

SN

TYPE

APs '041-153

'

Satpoint

Tolerance

.5 % of Span

.095 PSIG

.

1

SYSTEM ES Building Spray

INSTRUMENT NO.

PS-22 (BS4-PSS)

PRODUCT INSTRUCTION

SPAN

19 PSIG

Switch to close on pressure increase.

SWITCH LOC.

Step 10.2.1 Removal from service

,

Input

Desired

As

As Left

Error

DS1

to

Actuation

Found

Setpoint

PSI

Light Goes

Switch

Point (PSIG)

PSI

Out

PSI

YES

NO

10

Test Tee Cap Replaced - Verified By

,

All isolation valves. left. wen - V " Sed 2y ,

,

l

MAXIMUM ERROR IN PSI

PERFORMED BY

DATE

,

h

-_ . _-

__

_. _

_

.

..

.

i

DUKE POWER COMPANY

OCONEE NUCLEAR STATION

g m ,,.

-

ENCLOSURE 11.2.b

CALIBRATION DATA SHEET

,

IP/0/A/310/5D

ITEM Pressure Switch

TEST EQUIPMENT USED

-

11FG.

Mercoid

ITEM

SN

TYPE

APW-7041-IS3

-

Satpoint

Tolerance

.5 % of Span

.095 PSIG

'

SYSTEM ES Building Spray

1

INSTRUMENT NO.

PS-23 (BS4-PS6)

PRODUCT INSTRUCTION

SPAN

19 PSIG

Switch to close on pressure increase.

SWITCH LOC.

i

Stcp 10.2.1 Removal from service

,

Input

Desired

As

As Left

Error

DS2

.

l

to

Actuation

Found

Setpoint

PSI

Light Goes

l

Switch

Point (PSIG)

PSI

Out

YES

NO

l

PSIG

l

10

l

Test Tee Cap Replaced - Verified By

.

,

. All isolation valveer'W opea - Vocided 103 ^'

,

M IMUM ERROR IN PSI

PERFORMED BY

DATE

l

-

j,

l

-

_ . - , _ , . _ .

, _ _ , _ _ _ _ _ , _ _ , , , _ _ _ _ , _ _ _ , , . , , , , , . , , ,

__

-

.

.:

.

ENCLOSURE 11.3

IP/0/A/310/5D

MAXIMUM TOLERANCE LIMIT EXCEEDED SHEET

-

.

Initial & Date

(A).

Notified Instrument Supervisor that a tolerance

Tech

of 2% was exceeded on the following components:

Initial & Date

.

l

4

l

(B).

An evaluation was made on the above problem (s)

Inst. Sup.

and the following corrective action was taken:

l

.

%..

-

:

.

-

._... ..-.-,

,

.

~

.[

RETURFT0 IETE3

m

.

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.

RIN 2 5 sey

t

Docket No. 50-269

License No. DPR-38

EA 82-65

Duke Power Company

ATTN: Mr. W. O. Parker, Jr.

Vice President Steam Production

P. O. Box 2178

Charlotte, NC 28242

Gentlemen:

A special inspection was conducted by inspectors from Region II, U.S. Nuclear

Regulatory Commission from March 23 to April 1,1982 at your Oconee Nuclear

Station Unit 1.

The purpose of this inspection was to evaluate the safety

significance of a breach in Unit 1 reactor building containment integrity

discovered by an NRC Region II inspector on March 23, 1982.

The findings from this special

inspection indicated that two Technical

Specification Limiting Conditions for Operations had been exceeded during the

interval from July 1981 to March 1982. These findings were discussed in detail

with plant management on April 2,1982. In addition,' an Enforcement Conference

was held in the Region II office on May 21, 1982 in which NRC's overall safety

concerns relating to this event were discussed. At this meeting it was stated

i

that the immediate cause of these violations was a failure to follow surveillance

test procedures.

After consultation with the Director of the Office of Inspection and Enforcement,

I have been authorized to issue the enclosed Notice of Violation and Proposed

Imposition of Civil Penalty in the amount of Forty-four Thousand Dollars. We

propose to impose this civil penalty in order to emohasize the need for Duke

,

l

Power Company to ensure that procedures affecting safe operation of the nuclear

plant are meticulously followed and their completions appropriately verified. The

'

base penalty of Forty Thousand Dollars has been increased by Four Thousand

Dollars to Forty-four Thousand Dollars to reflect the significance of the

violation with respect to its duration.

i

In preparing your required response, you should follow the instructions specified

in the Notice which is enclosed with this letter. We note in your Licensee Event

Report (RO-269/82-08) that you have already taken several corrective actions.

Your required response should include, as a minimum, a complete description of

I

CERTIFIED MAIL

RETURNED RECEIPT REQUESTED

l

-

.

l

.

18(p

Dune Power Company

2

actions taken to ensure that procecures affecting safety--elated systems 9 ave

accropriate signoffs anc verifications te orecluce future violations of tnis

nature. Your reply and the results of future inspections will ce consicerec in

determining wnetner further enforcement action is appropriate.

In accordance with Section 2.790 of the NRC's " Rules of Practice". Part 2. Title

10. Coce of Feceral Regulations, a cooy of this letter and the enclosure will ce

placed in the NRC Puolic Document Room.

The responses directed oy this letter anc the enclosure are not sucject to the

clearance procedures of the Office of Management and Bucget as

eouired oy tne

Paperwork Reduction Act of 1980. PL 96-511.

Sincerely.

/s/

James P. O'Reilly

Regional Acministrator

Enclosure:

Notice of Violation anc Proposed

Imposition of Civil Penalty

ec w/ encl:

J. E. Smith Station Manager

i

.

b

.

' -

- - * ~

-

p

tT- e esg< p - y a w--e g' w p ms ,,- . _ _

-"=w.,

-

.

JUN 2 51982

.

NOTICE OF VIOLATION

AND

PROPOSED IMPOSITION OF CIVIL PENALTY

Duke Power Company

Docket No. 50-269

Oconee Nuclear Station Unit 1

License No. DPR-38

EA S2-65

As a result of a special inspection conducted, by the NRC Region II staff, from

March 23 to April 1,1982 at the Oconee Nuclear Station Unit I near Seneca, South

Carolina, it appears that a violation of NRC requirements occurred.

The

inspection findings were discussed with the station management at the conclusion

of the inspection.

NRC concerns regarding the violation were the subject of an

Enforcement Conference held at the Region II office in Atlanta on May 21, 1982

with officials of the Duke Power Company.

On March 23, 1982, the NRC Resident Inspector found that an instrument test con-

nection cap had been left off a one quarter-inch instrument calibration line

connected to the instrument sensing line that provided a direct pathway between

the Unit I reactor building atmosphere and the penetration room.

The licensee

immediately replaced the cap and thereby restored the reactor building contain-

ment integrity.

Licensee investigation revealed that most probably the indi-

vidual who had calibrated the associated pressure switch on July 9,1981 had

failed to replace the calibration line cap.

As a result of this failure,

containment integrity was violated and the reactor building spray initiation

system was degraded during certain periods in the July 9, 1981 to March 23, 1982

interval.

l

To emphasize the need for the licensee to ensure that procedures affecting safe

l

operation of the plant are meticulously followed, the NRC proooses to impose a

civil penalty of Forty-four Thousand Dollars for this matter.

The base penalty

for a violation of the severity level of this event is 540,000, as determined

from Tables IA and IB of the NRC Enforcement Policy (10 CFR Part 2. Appendix C)

47 FR 9987 (March 9, 1982).

Because of the duration of this event the civil

penalty has been increased by Four Thousand Dollars. In accordance with the NRC

l

Enforcement Policy and Section 234 of the Atomic Energy Act of 1954, as amenced

l

("Act"), 42 U.S.C. 2282, PL 96-295, and 10 CFF 2.205, the particular violation

'

and associated civil penalty is set forth below:

Technical Specification 3.6.1 requires that containment integrity be

,

maintained whenever reactor coolant system (RCS) pressure is greater than

l

1

300 psig ana temperature is greater than 200*F.

l

Technical Specification 3.5.1 requires that all three cnannels of both

trains of reactor building spray initiation ce operable when the reactor is

critical.

-

1

!

l2 inh

Gf *~m ,1n~

o " Y I " Y T ' l

1

. . _ _

M 2 5 agg

t

.

Appendix (Continued)

2

Technical Specification 6.4.1 requires that the olant be maintained in

accordance with approved procecures.

Procedure IP/0/A/310/50 was estao-

lished and approved to implement 6.4.1.

Step 10.2.10 of the procedure

requires replacement of the cap on the 1/4-inch calibration line connected

to the 1/2-inch sensing line for reactor building pressure switen IPS-22.

Contrary to the above, on July 9, 1981, the licensee failed to follow step

10.2.10 of procedure IP/0/A/310/50.

As result of the failure the following

conditions existed between July 9,1981 and March 23, 1982.

1.

Containment integrity of the Unit I reactor building was not maintained for-

fifty-one days while RCS pressure was greater than 300 psig and temperature

was greater than 200'F.

2.

For thirty-two days, one of three channels of Train A of reactor building

spray initiation for Unit I was inoperable while the reactor was critical.

This is a Severity Level III violation (Supplement I).

(Civil Penalty - 544,000).

Pursuant to the provisions of 10 CFR 2.201, Duke Power Company is hereby required

to submit to the Director, Office of Inspection and Enforcement, USNRC,

Washington, DC 20555, and a copy to the Regional Administrator. USNRC Region II,

within thirty days of the date of this Notice, a written statement or exoianation

in reply, incluaing for the violation:

(1) admission or denial of the alleged

violation; (2) the reasons for the violation if admitted: (3) the corrective

steps which have been taken and the results achieved: (4) the corrective steps

which will be taken to avoid further violations: and (5) the date when full

compliance will be achieved.

Consideration may be given to extending the

response time for good cause shown.

Under the authority of Section IS2 of the

Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Within the same time as provided for the response required above under 10 CFR 2.201, Duke Power Company may pay the civil penalty of 544,000 or may protest

imposition of the civil penalty in whole or in part by a written answer. Should

.

Duke Power Company fail to answer within the time specified, the Director. Office

l

of Inspection and Enforcement will issue an order imposing the civil penalty

proposed above.

Should Duke Power Company elect to file an answer in accordanca

with 10 CFR 2.205 protesting the civil penalty, sucn answer may:

(1) deny the

violation presented in this Notice in whole or in part: (2) demonstrate

extenuating circumstances

(3) show crror in this Notice; or (4) show otne -

reasons why the penalty should not be imposed.

In addition to protesting the

civil penalty in whole or in part, such answer may reouest remission or

mitigation of the penalty.

In reauesting mitigation of tne proposea penalty, the

five factors contained in Section IV(B) of 10 CFR Part 2, Appendix C shoula be

addressed.

Any written answer in accordance with 10 CFR 2.205 should be set

i

forth separately from the statement or explanation in reply pursuant to 10 CFR

'

2.201, but may incorporate by specific reference (e.g. , giving page and paragraph

numbers) to avoid repetition.

Duke Power Company's attention is directed to the

otner provisions of 10 CFR 2.205, regarcing the procecure for imposing a civil

oenalty.

.

~

._

'**

8 * * * # p 'e me e .' we + e e,aw e' e e,

w'

_

-

.

AN 2 5 f882

.

Appendix (Continued)

3

Upon failure to pay any civil penalty due. which has been subsequently determined

in accordance with the applicable provisions of 10 CFR 2.205, this matter may be

,

referred to the Attorney General, and the penalty, unless compromised, remitted,

'

or mitigated, may be collected by civil action pursuant to Section 234c of the

Act 42 U.S.C. 2282.

FOR THE NUCLEAR REGULATORY COMMISSION

James P. O'Reilly

Regional Administrator

Dated at Atlanta, Georgia

this

day of June 1982

j

.

e

I

l

.

l

.

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4.-4-w=-=

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ML20027D730

Text

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