ML20217F703
| ML20217F703 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/06/1997 |
| From: | Nunn D SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-361-97-17, 50-362-97-17, NUDOCS 9710090031 | |
| Download: ML20217F703 (6) | |
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An MMON lYp RWWW Cunm October 6,1997 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Reply to a Notice of Violation San Onofre Nuclear Generating Station, Units 2 and 3
Reference:
Letter, Mr. Thomas P. Gwynn (USNRC) to Mr. Harold B. Rey (SCE),
dated September 4,1997 l
The referenced letter transmitted the results of NRC Inspection Report No.
50-361/97-17 and 50-362/97-17, conducted July 6 through August 16,1997, at Southern California Edison's (SCE), San Onofre Nuclear Generating Station, Units 2 and 3. The enclosure to the referenced letter contained a Notice of Violation requiring l
a response to three instances (361/97017-02,361/97017-03, and 361; 362/97017-05) of falling to follow procedures as written. Because the NRC had completed its review of the corrective actions for the third violation (362/9700-04), no response to that violation is required. The enclosure to this letter provides SCE's reply to the other three violations, if you have any further questions, please contact me.
Sincerely, I
b 9710090031 971006
'[M PDR ADOCK 05000361
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Enclosure:
As stated cc:
E. W, Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident inspector, San Onofre Units 2 and 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 l'. O. Ikn 128
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4 ENCLOSURE REPLY TO A NOTICE OF VIOLATION The enclosure to Mr. T. P. Gwynn's letter dated September 4,1997 states, in part:
VIOLATION A "A.
Unit 2 Technical Specification 5.5.1.1.s requires that written procedures be established, implemented, and slalotained 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 51.7, 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, Muldelines During Power Ascension,'
Step 6.2.5.3, requires tho' operators 'do NOT EXCEED the applicable Maximum Core Power ' escalation Rate of Attachment 1/ Attachment i states that 'when fuel cladding leaks are known to exist / tne Maximum Core Power Escalation Rate is 5 percent pur hour, and directs operators to refer to the Operations Physics Summary for current status of fuel failure per Reactor Engineering Transmittal. The Reactor Engineering 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, Operation 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 11:45 a.m. and 1:30 p.m. (approximately 11 percent per hour), with a maximum 1-hour change of approximately 14 perc6nt per hour.
"This is a Severity Level IV violation (Supplement I) applicable to Unit 2 (361/9717 02)."
1.
Reason for the Violation As noted in the inspection report, the reasons for the violation were:
(1) miscommunication between the Operations Manager and the Control Room Operators; end (2) inadequate management oversight. Because the Control Room Operators misinterpreted the Operations Manager's directions, and assumed that there were no fuel cladding leaks despite the Reactor Engineering Data Transmittal letter, Operators thought the procedural restriction no longer applied. Also, senior
ENCLOSURE management authorized the use of a higher power ramp rate than that previously specified without ensuring that actual procedural requirements allowed this change.
2.
Corrective Actions Taken and the Results Achieved l
As corrective actions: (1) the Operations Manager and operating crews were coached and counseled on problem solving and communication of issues; (2) senior management have reviewed this response; and (3) as noted in the inspection report, SCE radiochemistry analyses performed after equilibrium conditions were achieved in the RCS indicted the fuel cladding leak had not been adversely affected by l
the power ascension; and, (4) procedure SO23 51.7, Attachme:
1, is being reviewed I
to ensure that sufficient flexibility exists for management to modify ramp rate i
restrbtions in specific cases so long as they stay within vendor requirements.
J.
Corrective Actions That Will be Taken No further corrective actions are planned.
4.
Date When Full Compliance Will be Achieved Full compliance was achieved on July 15,1997, when the power ramp rate was restored to the 5% per hour limit.
VIOLATION B The enclosure to Mr. T. P. Gwynn's letter dated September 4,1997 states, in part:
"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.
" Operating instruction SO23-3-3.21 (sic), Revision 15, 'CVCS Charging and Letdown,' Step 6.3.9.1, states that 'if desired, due to LV-0227A leak-by, ilma Perform the following:
- Close S2(3)1901MU924, LV-0227A to Radwaste Block
- Place LV-0227A handswitch in Manual _
4 ENCLOSURE
- Document closure by placing a Caution Tag on 2(3) 1901MUO924 handwheel and LV-0227A handswitch
. Initiate an Action Request for LV-0227A leak-by.'
{
" Contrary to the above, on July 29,1997, Operating Instruction S023-3-3.21 (sic), Revision 15, was not implemented in that, due to Valve LV-0227A leak by, the licensee closed Valve S21901MUO924, documented i
closure by placing the required caution tags, initiated an action request for Valve 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 (361/9717-03),"
1 1,
Reason for the Violation The reason for the violation was individual personal error. As noted in the inspection i
report, the operator had not reviewed procedure S023-3-2,1 when he closed Valve 2MUO924.
l 2.
Corrective Actions Taken and the Results Achieved As discussed in the inspection report, the following corrective actions were taken:
(1) the individuals involved were counseled; (2) operating crews were briefed on the event; and (3) procedures SO23-3-2.1 and SO23-3-3.37 (a companion procedure on water inventory balance) were modified to remove the requirement for the diversion valve to be in manual when the radwaste isolation valve is closed (note: these changes will also make the procedures consistent with each other, which they previously weren't (SO23-3-3.37 did not require placing the control switch in manual, although the procedure was not in use at the time)).
3.
Corrective Actions That Will Be Taken No further corrective actions are planned.
4.
Date When Full Compliance Will be Achieved Full compliance was achieved on July 29,1997, when the letdown diver. ion valve handswitch was placed in the manual position.
o ENCLOSURE VIOLATION C No response was required for this violation.
VIOLATION D The enclosure to Mr. T. P. Gwynn's letter dated September 4,1997 states, in part:
"D.
10 CFR Part fo, Appendix B, Criterion IX, requires, in part, that measures shall be established to assure that special processes, including welding and nondestructive testing, are control'ed and accomplished using qualified procedures in accordance with applica >le codes and specifications.
"In a letter from the NRC Office of Nuclear Reactor Regulation to the Senior Vice President, Southern California Edison, dated January 11,1995, and pursuant to 10 CFR 50.55a(a)(3)(li), the use of ASME Code Case N-416-1 was authorized.
"ASME Code Case N-416-1 required that nondestructive examination be performed in accordance with the acceptance criteria of the applicable subsection of the 1992 Edition of Section Ill.
t
" 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.
Subsectior4 9350, Section Ill,1992 Edition, stipulated that rounded indications of up to 3/16 inch were acceptable.
"This is a Seve,ity Level IV violation (Supplement I) applicable to Unit 2 (361/97017-05)(sic)."
4
o..
ENCLOSURE 1.
Reason for the Violation As noted in the inspection report, the reason for the.3lation was an individual personal error in not properly identifying the ASME Code requirements.
2.
Corrective Actions Taken and the Results Achieved I
As noted in the inspection report, SCE reviewod applicable records of postweld nondestructive examination (NDE) performed that had used ASME Code Case N-416-1 for the cited application and similar applications. It was determined that this error did not result in any unacceptable conditions in the plant. Appropriate personnel have been briefed on this event.
3.
Corrective Actions That Will Be Taken No further corrective actions are planned.
4.
Date When Full Compliance Will be Achieved Full compliance was achieved by July 25,1997, when SCE verified no relevant indications (> 1/16 inch) were detected during the NDE.
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