ML20141K636

Discusses Review of Licensee 960627 Response to 960531 NOV Re Activities Conducted at Cns.Informs That Although Violation of 10CFR50.59 as Cited in Violation B Did Not Occur,Nrc Requirements Violated.Reissues NOV

ML20141K636
Person / Time
Site:	Catawba
Issue date:	05/16/1997
From:	Jaudin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Mccollum W DUKE POWER CO.
Shared Package
ML20141K638	List: ML20141K636 ML20141K661
References
50-413-96-05, 50-413-96-5, 50-414-96-05, 50-414-96-5, EA-97-179, NUDOCS 9705290257
Download: ML20141K636 (8)

Text

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May 16, 1997 l EA 97-179 I

Duke Power Company ATTN: Mr. W. R. McCollum Site Vice President  ;

Catawba Site 4800 Concord Road York, SC 29745-9635

SUBJECT:

NOTICE OF VIOLATION l (NRC INSPECTION REPORT N0. 50 413/96 05, AND 50 414/96 05)

Dear Mr. McCollum:

We have completed our review of your response of June 27, 1996, to our Notice of Violation issued on May 31, 1996, concerning activities conducted at your Catawba facility. In your response, you admitted Violation A and denied Violation B.

After careful consideration of the basis for your denial of Violation B, we )

have concluded, for the reasons presented in the Enclosure to this letter, l l that although a violation of 10 CFR 50.59 as cited in Violation B did not occur, a violation of NRC requirements occurred. Violation B as stated in the Notice of Violation issued on May 31, 1996, is withdrawn and replaced by the i enclosed Notice of Violation. Please note that you are required to res>ond to this letter and should follow the instructions in the enclosed Notice w1en preparing your response. The NRC will use your response, in part, to determine whether further action is necessary to ensure compliance with l regulatory requirements.  !

We are concerned regarding statements made in your response to Finding 2 of l l Violation B. Your response implies that you find it acceptable to implement i design changes prior to the certification and release of design calculations.

This practice is contrary to your design control measures which implement the requirements of 10 CFR 50, Appendix B. Criterion III. Your response should also address clarification of your understanding of your design control

measures and the necessary corrective actions to comply with requirements of

! 10 CFR 50, Appendix B.

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.In accordance with 10 CFR 2.790 of the NRC's Rules of Practice," a copy of this letter and its enclosures will t,e placed in the NRC Public Document Room (PDR).

Sincerely, Original signed by Johns P. Jaudon Johns P. Jaudon, Director l Division of Reactor Safety Docket Nos. 50 413, and 50 414 i License Nos. NPF 35, and NPF 52 l

Enclosures:

1. Evaluations and Conclusion

2. Notice of Violation

, cc w/encls: Max Batavia, Chief l H. S. Kitlan Bureau of Radiological Health <

l Regulatory Compliance Manager S. C. Department of Health l Duke Power Company and Environmental Control 4800 Concord Road 2600 Bull Street York, SC 29745 9635 Columbia, SC 29201 Paul R. Newton Richard P. Wilson, Esq.

Legal Department (PB05E) Assistant Attorney General Duke Power Company S. C. Attorney General's Office 422 South Church Street P. O. Box 11549 Charlotte, NC 28242 0001 Columbia, SC 29211 Robert P. Gruber Michael Hirsch Executive Director Federal Emergency Management Agency Public Staff - NCUC 500 C Street, Sw, Room 840 P. O. Box 29520 Washington, D. C. 20472 Raleigh, NC 27626 0520 North Carolina Electric J. Michael McGarry, III, Esq. Membership Corporation

, Winston and Strawn P. O. Box 27306 1400 L Street, NW Raleigh, NC 27611 Washington, D. C. 20005 l Karen E. Long North Carolina MPA 1 Assistant Attorney General

! Suite 600 N. C. Department of Justice i P. O. Box 29513 P. O. Box 629 l Raleigh, NC 27626 0513 Raleigh, NC 27602 f (cc cont'd - See page 3) l l

. . , _ . _ . _ _ . . . . _ _ . _ . _ . _ . ~ . . . _ . _ , _ _ _ _ . _ _ _ . _ . _ _ _ _ . . ~ _ . . _ _

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(cc cont'd) Distribution w/encis: I Saluda River Electric P.. Tam, NRR .

Cooperative, Inc. R. Carroll, RII P. O. Box 929 R. V. Crienjak. RII i

. Laurens SC 29360 N. Economos, RII- '

l- . R. Baldwin, RII  :

i Peter R. Harden IV PUBLIC I i- ' Account Sales Manager I Power Systems Field Sales NRC Resident Inspector l Westinghouse Electric Corporation U.S. Nuclear Regulatory Commission

, P. 0.-Box 7288 4830 Concord Road l Charlotte, NC 28241 York, SC 29745 i

L County Manager of York County l

York County Courthouse l- York, SC 29745 Piedmont Municipal Power Agency 121 Village Drive Greer, SC 29651 G. A. Copp Licensing EC050 Duke Power Company P. O. Box 1006 Charlotte, NC 28201 1006

.T. Richard Puryear l Owners Group North Carolina Electric Membership Corporation ,

4800 Concord Road York, SC 29745 [ ehg f

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DATE 05 / 11 / 97 05 / 1 3 / 97 05 / v1 / 97 48hf / 97 05 / /9'05 / / 9:

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I Evaluation and Conclusion '

On May 31, 1996, a Notice of Violation (Notice) was issued for violations identified during a routine NRC inspection. Duke Power Company responded to the Notice on June 27, 1996. Duke denied Violation B based on their

• contention that they com) lied with their procedures in the unreviewed safety question screening and t1e operability review 3erformed prior to. revising the auxiliary feedwater operating procedure. The NRC's evaluations and ,

conclusions regarding the licensee's denial are as follows: i l Restatement of Violation B .

l 10 CFR 50, Appendix B, Criterion V, requires that activities affecting '

quality be prescribed by documented instructions or procedures, and shall be accomplished in accordance with these instructions or procedures.

l 10 CFR 50.59 requires the performance of an evaluation to determine if changes to the facility (systems, structures, or components) or facility operating procedures described in the l Safety Analysis Report (SAR) involves an unreviewed safety question.

P Duke Power Nuclear Station Directive (NSD) 209, 10 CFR 50.59 Evaluation, Revision 3, effective October 1,1995, implements the requirements of 10 CFR 50.59. Section 209.10.2 of NSD 209 specifies the screening process required to be performed to determine if a facility or procedure change constitutes an unreviewed safety question which in part requires negative answers to the following questions:

Does the activity change the facility as described in the SAR?

Could the activity adversely affect any system, structure, or component that is necessary in accordance with the SAR?

NSD 209 defines the SAR as the set of documents used to support issuance of a plant operating license. These cacuments include, but are not limited to, the Facility Operating License, the NRC  !

Safety Evaluation Report, the FSAR, the Technicel Specifications, l and other licensing documents.  !

Section 101.4.3 of Engineering Directives Manual EDM 101, Engineering Calculations / Analyses, Revision 4, dated March 30. i 1995, requires certification of design calculatior s prior to release of calculation results.

l l Enclosure 1 l

i

. j l

2 Contrary to the above:

1. The 50.59 evaluation was inadequate in that the negative i responses to the NSD 209 questions were incorrect for addressing the February 21, 1996, change to Enclosure 4.12 l of procedure OP/1/A/6250/02, Auxiliary Feedwater System.

l Increasing the allowable auxiliary feedwater )iping j temperature to 250 F changed the design of t1e auxiliary I feedwater system, as described in the SAR. The reduction of

.the concrete expansion safety factor,.from four to two, to  !

permit operability of the auxiliary feedwater piping at a '

i temperature of 250* F decreased the margin of safety and had l a potentially adverse effect on the design of the auxiliary L feedwater piping. NRC IE Bulletin 79 02, a licensing ,

document, requires a minimum safety factor of four for &

concrete expansion anchors. ,

l 2. Engineering calculations were released prior to completion i i

of the design certification process, in that on February 21, l 1996, a change to Enclosure 4.12 of Procedure OP/1/A/6250/02 l was made with uncertified calculations. In changing

, Procedure OP/1/A/6250/02, for raising the acceptable i l Auxiliary Feedwater suction temperature, approved February ,

21, 1996, engineering calculations supporting this change

~

l were not approved until on, or after, March 5, 1996. These calculations formed the bases for approval of the procedure change ,

=

This is a Severity Level IV violation (Supplement I)  :

Summary of Licensee's Response  ;

e The licensee contends that they complied with their procedures and NRC requirements for performing the 50.59 evaluation and that their responses made t in the Unreviewed Safety Question (USQ) screening were correct. The licensee  ;

also contends that the calculations performed to determine the acceptability  :

of 50.59 changes are not required to be design verified prior to j implementation of the change.

NRC Evaluation l i f i NRC has carefully reviewed the licensee's response. We have concluded that although a violation of 10 CFR 50.59 did not occur, the licensee failed to 5

, follow their procedures when performing the 50.59 screening (NSD 209) and the I

operability review (NSD 203). While the licensee's reasons for revising procedure OP/1/A/6250/02 to eliminate an operator workaround, and i

l l l Enclosure 1  !

! l i

?

I i l._______._. _ _ _ _ _ _ . _ . . -

~

3 to preclude frequent operation of the auxiliary feedwater pump are justified, their 50.59 evaluation and screening performed prior to implementing the procedure changes was not done in accordance with procedure NSD 209. The change to the operating temperatures for the AFW system was a design change.

NSD 203, Operability, Revision 4, effective date January 1, 1996, is the procedure which specifies the steas to be used in performance of. operability reviews. Paragraph 302.7.4 of NS) 203 requires performance of a 10 CFR 50.59 Evaluation for any system, structure, or component (SSC) which is concluded in an operability evaluation to be degraded. NSD 203 defines operable but degraded, in part, as a situation when a SSC relies on temporary changes to a design limit or design basis in order to remain operable. Paragraph 203.9.2 of NSD 203 requires that a calculation be originated in accordance with EDM-101, Engineering Calculations / Analyses, for a safety related (QA Condition)

SSC when the operability evaluation will change the design basis, design criteria, or design limit.

Nuclear System Directive (NSD) 209,10 CFR 50.59 Evaluations, is the licensee's procedure for performing reviews to determine if a design change changes the facility as described in the safety analysis report. NSD 209 provides an initial screening process which consist of five questions to determine if an unreviewed safety question evaluation is required. The five questions are as follows:

1. Does the activity change the facility as described in the SAR?

2. Does the activity change procedures, methods of operation, or alter a test or experiment as described in the SAR?

3. Does the activity appear significant enough to require inclusion in the SAR?

4. Could the activity adversely affect any SSC that is necessary to operate the facility in accordance with the SAR?

5. Does the activity perform a test or experiment that is NOT described in thn SAR?

The safety analysis report (SAR) is defined in NSD 209 as the set of documents used to support issuance of a plant operating license. These documents include, but are not limited to, the Facility Operating License, the NRC Safety Evaluation Report, the FSAR, the Technical Specifications and other licensing documents such as selected licensee commitments and other communications between the licensee and NRC.

If the answer to all five questions is "No", an unreviewed safety question l (USQ) evaluation is not required. The licensee's 50.59 evaluation for the I change to Procedure OP/1/A/6250/02 resulted in "No" answers to all five questions, thereby resulting in determination that a USQ evaluation was not t

Enclosure 1 l

l

4 I

required. The NRC determined that the answers to questions one and four ,

should have been "Yes", and that a USQ evaluation should have been performed.

The licensee contends that their "No" answers to these screening questions were correct.

For Question 1, the licensee argues that the change to the operating  :

. temperature . limit did not affect the structural integrity of the auxiliary '

feedwater system, or reduce the margin of safety. The licensee stated in  !

l their response that reduction of the factor of safety for loading of concrete  ;

expansion anchors from four to two complies with IE Bulletin (IEB) 79 02, Pipe  :

Support Base Plate Designs Using Concrete Expansion Anchor Bolts. IEB 79 02 i was used to su) port issuance of the plant operating license, and therefore, in '

accordance witi NSD 209, is included in the SAR. IEB 79 02 requires concrete expansion anchors to have a minimum factor of safety of four. The licensee  !'

interprets IEB 79 02 as >ermitting interim operation with a factor of safety of two until the anchor )olts can be modified at the next refueling butage.  !

NRC disagrees with the licensee regarding their interpretation of IEB 79 02.

The factor of safety of two criteria s)ecified in IEB 79-02 permitted interim l operations when concrete expansion anc1 ors were identified during inspections being performed by licensees to comply with IEB 79 02. Similar criteria for i interim operations for resolving degraded conditions are specified in NRC  ;

Generic Letter 91 18. Neither IEB 79 02 nor GL 91-18 address implementation l of design changes which result in reducing the factor of safety for concrete expansion anchors below the minimum values specified in IEB 79-02.

Implementation of the procedure change, i. e. design change, permitted increasing the operating temperature of the auxiliary feedwater piping. The

temperature increases resulted in increasing the loads acting on pipe support i concrete expansion anchors. Thus, this activity did affect the facility,  !

s)ecifically the margin of safety for the concrete expansion anchors. i Tierefore, the answer to Question 1 should have been "Yes", based on the ~

l criteria of NSD 209.

IEB 79-02 was issued by NRC due to numerous problems identified at several facilities with installation of concrete expansion anchors. IEB 79 02  ;

required a minimum factor of safety of four for the drilled in shell or wedge type concrete expansion anchors used at the licensee's facility. The factor of safety is defined as the ultimate capacity of the anchor determined by  !

! testing divided by the design load. This requirement is implemented in i paragraph 3.10 of Duke Specification CNS 1206.00-04 0001, Design Specification i for Nuclear Safety Related (QA Condition 1) and QA Condition 4 Component i Supports. The reduction of the factor of safety as much as 50 percent, from 1 i specified minimum value of four, to factors of safety as low as two, '

essentially doubled the design load carried by the anchors. This activity had the potential to adversely affect the AFW system, which is necessary to operate the 31 ant in accordance with the SAR. Therefore, the answers to question num)er 4 should have been "Yes".

! Enclosure 1 l

5 The positive answers to Questions one and four would have required the licensee to perform additional review of the issue to determine if the temperature changes would result in an unreviewed safety question. These additional reviews are described in NSD 209.

The licensee also contends that, in finding 2 of Violation B, NRC misinterpreted Duke arocedure NSD 209, which specifies the requirements for performance of 10 CFR 50.59 evaluations. The licensee stated since NSD 209 specifies that an engineering review was an adequate level of review to perform the 10 CFR 50.59 screening, the requirements for design control reasures do not apply to 50.59 changes. The licensee's response implies that it is not necessary to perform calculations and have the calculations design verified and certified prior to implementing a 10 CFR 50.59 change. However, as discussed above, Duke procedure requires origination of design calculations in accordance with EDM-101 for operability reviews for conditions which will change the design basis, design criteria, or design limit. Review of the licensee's justification for revising the operating temperature of the AFW piping was based on the design output from several calculations referenced in the inspection report. The change to the AFW system operating temperature was a design change. The requirements for design changes are specified in 10 CFR 50, Appendix B, Criterion III, which in part requires that measures be established to assure that the design basis are correctly translated into design output documents. Criterion III further recuires that the design control measures shall provide for verifying the acequacy of the design, and that design changes be subject to design control measures commensurate with those applied to the original design. The procedures which implement the A'pendix a B, Criterion III requirements are the Engineering Directives Manual, w1ich are the Duke Power Company design control measures. The design change that the licensee implemented to the AFW system operating temperature is required to comply with 10 CFR 50, Appendix B, Criterion III. Therefore the requirements of EDM 101, as referenced in Finding 2 of Violation B were correct. 10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality be accomplished in accordance with prescribed instructions or procedures. Failure to implement the design verification and control measures as specified in EDM-101, and relying on unapproved calculations, to implement the design change, was a violation.

Furthermore, paragraph 203.9.2, Engineering Requirements, of procedure NSD 203 required operability evaluations to include preparation of a calculation in accordance with procedure EDH 101. The operability evaluation was required to be documented on a "203 Operability Notification Form".

NRC Conclusion For the above reasons, NRC concludes that a violation of 10 CFR 50.59, as stated in Violation B, did not occur. Therefore Violation B, issued on May L , 1996 is withdrawn. However, a violation of NRC requirements did occur as stated in the enclosed Notice of Violation.

Enclocure 1