05000321/LER-2005-001, Regarding Failure to Use an Appendix J, Option B, Type C Tested Component to Isolate Primary Containment Penetration
| ML052370438 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/24/2005 |
| From: | Sumner H Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-05-1460 LER 05-001-00 | |
| Download: ML052370438 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3212005001R00 - NRC Website | |
text
H. 1. Sumner, Jr.
Vice President Hatch Project August 24, 2005 Docket Nos.:
50-321 Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham. Alabama 35201 COMPANY Energy to Serve Your Worldw U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 1-2005-001 Failure to Use an A ~ ~ e n d i x J. Option B. Twe C Tested Component to Isolate Primarv Containment Ladies and Gentlemen:
Enclosed you will find a Hatch Unit 1 Licensee Event Report concerning the failure to use an Appendix J tested valve to isolate a primary containment penetration during repair activities.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, H. L. Sumner, Jr.
Enclosure: LER 1-2005-001 cc:
Southern Nuclear ODeratinp Comvanv Mr. J. T. Gasser, Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RTYPE: ~ ~ ~ 0 2. 0 0 4 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch
I 0 NRC FORM 366 LICENSEE EVENT REPORT (LER)
(See reverse for reauired number of
, h e NRC may
' digitslcharacters +or each block) conduct or spons&, and a is not required to respond to, the rmation collection.
Edwin I. Hatch Nuclear Plant - Unit 1 05000-32 1 (1
1 O F 4 Specify in Abstract below
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME ELEPHONE NUMBER (Indud8 Area Code)
Edwin I. Hatch / Kathy A. Underwood, Performance Analysis Supervisor (912) 537-5931 1 CAUSE I SYSTEM I COMPONENT I MANUFACTURER (
I I
I I
I REPORTABLE On 710 112005 at 1545 ET, Unit 1 was in the Run Mode at an approximate power level of 2793 CMWT (1 00 percent rated thermal power). At that time, it was determined that on 6/29/2005 a valve used to isolate primary containment during work on component 1P70-NO03 (a barrier for primary containment penetration
- 22) was not tested in accordance with Technical Specification 3.6.1.3 Condition C. This resulted in a condition prohibited by the plant's Technical Specifications. Component 1P70-NO03 associated with penetration 22 functions as a single barrier for the penetration and when it is INOPERABLE it is governed by Technical Specification 3.6.1.3 Condition C. The work associated with 1 P70-NO03 was completed and the clearance associated with the work restored on 613012005 and verified to be leak tight the next day.
The controls contained in two procedures associated with this event were determined to be inadequate in that they did not ensure the appropriate personnel reviewed the work package and they did not clearly identify that the safety related containment function for this work activity would be affected. These procedures will be revised to strengthen the controls that were identified as being inadequate during the investigation of this 1
event.
NRC FORM 368 (6-2004)
LICENSEE EVENT REPORT (LER)
Edwin I. Hatch Nuclear Plant - Unit 1
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On 4/29/2005, a problem was identified with instrument 1P70-N003 following a routine calibration. This instrument provides annunciation to the Main Control Room if the nitrogen supply pressure to the drywell pneumatics is out of specifications. A Work Order (WO) was generated to replace 1P70-N003-TV-1. The WO was scheduled to be worked June 28,2005. The WO was planned, reviewed, and packaged and the associated tag out for this work was drafted, reviewed and approved. Valve 1P70-N003-TV-1 was actually removed June 29,2005 for replacement.
On 7/01/2005 at 1545 ET, Unit 1 was in the Run Mode at an approximate power level of 2793 CMWT (100 percent rated thermal power). At that time, it was determined that the valve used to isolate primary containment (EIIS Code JM) on 6/29/2005 during work on component 1P70-NO03 (a barrier for primary containment penetration 22) was not in accordance with Technical Specification (TS) 3.6.1.3 Condition C.
This resulted in a condition prohibited by the plant's Technical Specifications. Component 1P70-N003 associated with penetration 22 functions as a single barrier for the penetration and is therefore governed by Technical Specification 3.6.1.3 Condition C. The Bases for TS 3.6.1.3 Condition C states in part that if the inoperable valve is required to be Type C tested per 10 CFR 50, Appendix J, Option B, the device chosen to isolate the penetration must also be subjected to Appendix J, Option B, Type C testing. The valve used to isolate the work activities associated with 1P70-NO03 was valve 1P70-F133 and this valve is not subjected to Appendix J, Option B, Type C testing. The work on component 1P70-NO03 was completed and the clearance restored on 6/30/2005. The duration of the work and inspection activities exceeded the Required Action Statement times associated with TS 3.6.1.3 Condition C.
The WO for this activity was stamped "LLRT Review Required." The LLRT review was performed without identifying any concerns for performing this activity in Modes 1,2, or 3. The LLRT review process does not require an Operations evaluation if the WO is scheduled to be performed during Modes 1,2 or 3.
The clearance review for operation impact was performed by a System Operator, which is a non-licensed position. There were two questions required to be answered by this review that were relevant to this event.
One question was does the Technical Specifications allow the equipment to be out of service with other activities scheduled during this time? The second question was, can the Clearance be performed in the current or planned Plant Status? These questions were intended to establish barriers that would prevent the plant from operating in a condition prohibited by the Technical Specifications. This form was completed by a System Operator that was not trained in Licensing or Technical Specifications. It should be noted however, that the Tagout Coordinator, a licensed SRO, did review the tagout without recognizing that the isolation valve being used by this clearance (1 P70-F133) was not an Appendix J, Option B, Type C tested component.
LICENSEE EVENT REPORT (LER)
Edwin I. Hatch Nuclear Plant - Unit 1 The Nitrogen Supply to the Drywell Pneumatics was still in service and pressurized to approximately 95 psig with no apparent leakage past isolation valve 1P70-F133. The piping downstream of the lP70-F133 was open after the removal of 1P70-003-TV-1 and the workers performing the work did not observe any nitrogen leaking past 1P70-F133. As a result, primary containment was never actually open to the environment during this evolution.
CAUSE OF EVENT
The evaluations for the impact of performing this evolution on line were inadequate.
The requirement for the WO to have an LLRT review and to be worked while the plant was operating in Mode 1 did not specifically trigger an operations review. The planners and schedulers did not realize that performing WO 105 1076502 in Mode 1 would impact Primary Containment.
The Operations Department Instruction required evaluations that were beyond the knowledge level or training of the System Operator performing the evaluation.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is required by 10 CFR 50.73(a)(2)(i)(B) because the plant entered a condition that was prohibited by the plant's Technical Specifications. Specifically, Technical Specification Bases for 3.6.1.3 Condition C requires in part that;
"... with one PCIV inoperable, except due to leakage not within limits, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. The device must be subjected to leakage testing requirements equivalent to the inoperable valve, except for inoperable valves in the Core Spray and Low Pressure Coolant Injection (LPCI) systems. For example: 1) if the inoperable valve is required to be Type C tested per 10 CFR 50, Appendix J, Option B, the device chosen to isolate the penetration must also be subjected to Appendix J, Option B, Type C testing."
The valve used to isolate the work activities associated with 1P70-N003 was valve lP70-F133 and this valve was not subjected to Appendix J, Option B, Type C testing. The work on component 1P70-N003 was completed and the clearance restored on 613012005. The duration of the work activity exceeded the Required Action Statement times associated with TS 3.6.1.3 Condition C resulting in the plant being in a condition prohibited by the plant's Technical Specifications. Although the work activity was actually performed within the Required Action Times associated with TS 3.6.1.3, the work was not inspected and tested until after the Required Action Times had been exceeded. Nevertheless, the inspections confirmed the work had been done satisfactorily and thus the valve was leak tight.
During this event primary containment was maintained by the closure of valve 1P70-F133. The leak tightness of this valve was assured in that the Nitrogen Supply to the Drywell Pneumatics was still in service and pressurized to approximately 95 psig with no apparent leakage past isolation valve 1P70-F133.
Therefore, it is concluded this event had no adverse impact on nuclear safety. This analysis is applicable to
CORRECTIVE ACTIONS
The repair work was verified to be leak tight on July 1,2005 in accordance with the provisions of 10 CFR 50 Appendix J Option B.
The appropriate personnel associated with the approval of this work activity were counseled concerning this A requirement has been established for Operations to review work orders performed during Modes 1,2 or 3 on LLRTIILRT components that may affect containment.
The Operations Department Instruction has been revised to require a currently licensed person to evaluate and complete any checklist item that addresses licensing issues if safety related equipment is involved with the work or tagout.
ADDITIONAL INFORMATION
Other Systems Affected: No systems other than those already mentioned in this report were affected by this
Failed Components Information
No failed components directly caused or resulted from this event.
Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events
There has not been any previous similar event in the past two years in which a primary containment isolation valve repair was not isolated with a Tech Spec compliant device.