ML20216C730

Notice of Violation from Insp on 970706-0816.Violation Noted:Operating Instruction SO23-5-1.7,Rev 10,was Not Implemented

ML20216C730
Person / Time
Site:	San Onofre
Issue date:	09/04/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20216C709	List: IR 05000361/1997017 ML20216C703 ML20216C730
References
50-361-97-17, 50-362-97-17, NUDOCS 9709090129
Download: ML20216C730 (4)

Text

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liNGLOSUSE_1 NOTICE OF VIOLATION Southern Cahfornia Edison Co. Docket Nos.: 50-361 San Onofro Nuclear Generating Station 50 362 License Nos.: NPF-10 NPF-15 During an NRC inspection conducted on July 6 through August 16,1997, four violations of NRC requirernents were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG 1600, the violations are listed below:

A. Unit 2 Technical Specification 5.5.1.1.a requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Appendix A, recommends general plant operating procedures for changing load.

Licensee Operating instruction SO23 5-1.7, Revision 10 " Power Operations,"

Section 6.3, " Power Ascension," in Step 6.3.1, directs operators to "throughout the power ascension, follow the guidelines of Section 6.2." Section 6.2, " Guidelines During Power Ascension," Step 6.2.5.3, requires that operators "do NOT EXCEED the applicable Maximum Core Power Escalation Rate of Attachment 1."

Attachment 1 states that "when fuel cladding leaks are known to exist," the Maximum Core Power Escalation Rata is 5 percent per hour, and directs operators to refer to the Operations Physics Summary for currcat status of fuel failure poi Reactor Engineering Transmittal. The Reactor Engineeiing Transmittal, dated July 8, 1997, states that there were indications of cladding defects, and that a ramp rate restriction of 5 percent per hour was in effect for the power ascension at or above 20 percent power.

Contrary to the above, on July 15,1997, Operating instruction SO23-51.7 Revision 10, was not implemented in that the operators increased Unit 2 reactor power from approximately 20 percent to 45 percent between 11:15 a.m. and 1:30 p.m. (approximately 11 percent per hour), with a maximum 1-hour change of approximately 14 percent per hour.

This is a Severity Level IV violation (Supplement I) applicable to Unit 2 (Violation 361/97017-02).

B. Unit 2 Technical Specification 5.5.1.1.a requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Appendix A, recommends procedures for operation of the chemical and volume control system, including the letdown / purification system.

9709090129 970904 PDR ADOCK 05000361 G PDR 1

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2 Operating Instruction SO23 3 3.21, Revision 15, "CVCS Charging and Letdown,"

Step 6.3.9.1, states that "if desired, due to LV-0227A leak by, then Perform the following:

Close S2(311901MU924, LV 0227A to Radwaste Block

• Place LV-0227A handswitch in Manual l

• Document closure by placing a Caution Tag on 2(311901MUO924 handwheel and LV 0227A handswitch

• Initiate an Action Request for LV-0227A leak by."

Contrary to the above, on July 29,1997, Operating instruction SO23 3-3.21 Revision 15, was not implemented in that, due to Valve LV-0227A leak by, the licenseo closed Valve S21901MUO924, documented closure by placing the required caution tags, initiated an action request for Valvo LV-0227A leak by, but did not place the Valve LV-0227A handswitch in manual.

This is a Severity Level IV violation (Supplement 1) applicable to Unit 2 (Violation 361/9717-03).

C. 10 CFR Part 50, Appendix B, Criterion V, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Procedure SO2313.34, Temporary Change Notice 2 2, " Reactor Remote Four Finger CEA [ control element assembly) Removal and Reinstallation," Step 6.1.29, directed " Lower the extension shaft and CEA guide ensuring that the guide seats properly on the CEA hub and that the extension shaft lowers correctly into the CEA hub. Check the position of the extension shaft with an underwater camera."

In Procedure SO231-3.46, Revision 1, " Reactor - Remote Five Finger CEA Extension Shaft Coupling," the Caution on page 4 directed that "The 4-element CEA (CEAs 88,89,90, and 91) extension shafts are not normally coupled. Refer to SO2313.34, Installation and Removal of Four Finger Control Element Assemblies, if 6 this is required."

Vendor Technical Manual Reactor VesselInternals instruction Manual," dated Jaly 6,1984, Appendix D, " Installation, Coupling, and Uncoupling of 4 Element CEA Extension Shafts," Step 7.1, recommended " position the 4 element CEA extension shaft alignment tool over a 4 element CEA location . . ."

Contrary to the above, on May 25,1997, Maintenance Order 9704212000, used to couple four fingered Control Element Assembly 91, was not appropriate to the  !

circumstances. Maintenance Order 9704212000 required that four-fingered Control j Element Assembly 91 be coupled to an extension shaf t, while in the core, with the  !

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t 3-I upper guide structure installed, and without the use of an extension shaft alignment tool or an underwater camera to verify proper coupling, using the remote five finger CEA extension shaf t coupling Procedure SO23 I 3.46. This resulted in the reactor being reassembled and placed in Mode 3 without Control Element Assembly 91 coupled to an extension shaft, b

, This is a Level IV violation (Supplement 1), applicable to Unit 3

(Violation 362/97017 04).

D. 10 CFR Part 50, Appendix B, Criterion IX, requires, in part, that measures shall be established to assure that special processes, including welding and nondestructive testing, are controlled and accomplished using qualified procedures in accordance with applicable codes and specifications.

, In a letter from the NRC Office of Nuclear Reactor Regulation to the Senior Vice i President, Southern Califomia Edison, dated January 11,1995, and pursuant to 10 CFR 50.55ala)(3)(ii), the use of ASME Code Case N-4161 was authorized.

4 ASME Code Case N-4161 required that nondestructive examination be performed in accordance with the acceptance criteria of the applicable subsection of the 1992 Edition of Section Ill.

4 Subsection 2546.3, Section Ill,1992 Edition, required that, for wall thicknesses

] less than 5/8 inch, rounded indications, observed while performing liquid dye penetrant examination, of greater than 1/8 inch on the pipe, were unacceptable.

Contrary to the above, on July 8,1997, the inspectors identified that licensee personnel performed liquid dye penetrant examination of a completed weld and adjacent pipe on Unit 2 Reactor Coolant Loop 1 A Injection Check Valve 2MUO19, using the wrong acceptance standard. Weld Record WR2 97-374 stipulated the use of Subsection NB 5350, Section ill,1992 Edition, as the acceptance standard for this examination. Subsection 5350, Section lil,1992 Edition, stipulated that rounded indications of up to 3/16-inch were acceptable.

This is a Severity Level IV violation (Supplement 1) applicable to Unit 2 (Violation 361/97017-05).

Pursuant to the provisions of 10 CFR Part 2.201, Southern California Edison Co. is hereby

. required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for Violations A, B, and D: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the i-

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1 results achieved, (3) the corrective steps thet will be taken to avoid further violations, and I (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response, if an adequete reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should i not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

l The NRC has concluded that information regarding the reason for Violation C, the corrective actions taken and planned to correct Violation C and prevent recurrence and the l date when full compliance was achieved are already adequately addressed in Enclosure 2.

However, you are required to submit a written statement or explanation pursuant to 10 CFR 2.2.01 if the description therein does not accurately reflect your corrective actions or your position. In that case, or if you choose to respond, clearly mark your response as a

" Reply to a Notice of Violation," and send it to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional' Administrator, Region IV, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice 1 of Violation (Notice).

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. However, if you find it i necessary to include such information, you should clearly indicate the specific information  !

that_you desire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information from the public.

Dated at Arlington, Texas this 4th day of September 1997 i