05000366/LER-2010-001, Regarding Failure to Recognize PCIV as Inoperable Results in a Condition Prohibited by the Technical Specification
| ML101200272 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/30/2010 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-10-0812 LER 10-001-00 | |
| Download: ML101200272 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3662010001R00 - NRC Website | |
text
Dennis R. Madison Southern Nuclear Vice President I-latch Operating Company, Inc.
Piant Edwin I. Hatch 1102e Hatch Parkway Nortr, Baxley, G90rgia 3'i S1~~
To! 912.537.5859 Fax 912.368.20TI SOUTHERNA COMPANY April 30, 2010 Docket No.:
50-366 NL-10-0812 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Failure to Recognize PCIV as Inoperable Results in a Condition Prohibited By the Technical Specification Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning the failure to recognize a PCIV as inoperable which resulted in a condition prohibited by the technical specification.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, D. R. Madison Vice President - Hatch DRM/MJKI Enclosure: LER 2-2010-001 cc:
Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch
03 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 U.S. NUCLEAR REGULATORY COMMISSION 9-2007)
, the NRC may not conduct or sponsor, and a person is not reqUired to respond to, the information collection.
- 2. DOCKET NUMBER PAGE
- 1. FACILITY NAME 05000366 1 OF 4 Edwin I. Hatch Nuclear Plant Unit 2 r
f4. TITLE Failure to Recognize PCIV as Inoperable Results in a Condition Prohibited By the Technical Specification
- 8. OTHER FACILITIES INVOLVED FACILITY NAME
- 7. REPORT DATE
- 6. LER NUMBER
- 5. EVENT DATE DOCKET NUMBER REV SEQUENTIAL YEAR MONTH DAY YEAR DAY YEAR MONTH 05000 NO.
NUMBER DOCKET NUMBER FACILITY NAME 04 2010 30 2010 0
05000 2010
- - 001 10
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
- 19. OPERATING MODE o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) 1 o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x)
- 10. POWER LEVEL o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71 (a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71 (a)(5) 99.8 o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) 181 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME I~ELEPHONE NUMBER (Include Area Code)
Edwin I. HatCh I Steve Tipps, Principal Licensing Engineer 912-537-5880 MANU REPORTABLE MANU REPORTABLE SYSTEM
CAUSE
COMPONENT
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX X
81\\1 R344 Yes SHV
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED SUBMISSION MONTH DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO DATE ABSTRACT (Limitto 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On March 10,2010 at 20:00 EST, Unit 2 was at approximately 2799 CMWTh, 99.8 percent power. Earlier that day Operations personnel were performing the Reactor Core Isolation Cooling (RCIC) operability procedure. During that evolution an annunciator indicated that the RCIC barometric condenser pressure was high. Subsequent investigation determined that the vacuum pump discharge check valve was stuck closed. A review of the system operability was performed and since the barometric condenser is not required for RCIC to fulfill its' design function for its' mission time it was determined that RCIC was operable. This valve performs a second function of primary containment isolation valve. The cause of the valve sticking was unknown and thus the valve must be considered inoperable and the appropriate Technical Specification should be entered for the inoperable valve. This was not identified until after the action time for the applicable Technical Specification had expired; therefore, a condition prohibited by the Technical Specification existed.
The cause of this event was failure to review operability from both a system and a component level.
Corrective actions are to incorporate this example into the Operations training program and the failed valve has been replaced and tested with an acceptable PCIV.
PRINTED ON RECYCLED PAPER NRC FORM 366 (9-2007)
(If more space is required, use additional copies of NRC Form 366A)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).
DESCRIPTION OF EVENT
On March 10,2010 at 20:00 EST, Unit 2 was at approximately 2799 CMWTh, 99.8 percent power.
Earlier that day Operations personnel were performing the Reactor Core Isolation Cooling (RCIC, EllS Code BN) operability procedure. During that evolution an annunciator indicated that the RCIC barometric condenser pressure was high. Subsequent investigation determined that the vacuum pump discharge check valve, 2E51-F028, was stuck closed. A review of the system operability was performed and since the barometric condenser is not required for RCIC to fulfill the design function for the system mission time it was determined that RCIC was operable. This valve performs a second function of primary containment isolation valve (PCIV, EllS Code NH). The cause of the valve sticking was unknown and thus the valve must be considered inoperable and the appropriate Technical Specification should be entered for the inoperable Primary Containment Isolation Valve. This was not identified until after the action time for the applicable Technical Specification had expired; therefore, a condition prohibited by the Technical Specification existed.
CAUSE OF EVENT
The cause of this event was the failure to review the PCIV operability from both a system and a component level. This resulted in the failure to identify the need to enter the plant Technical Specifications for the inoperable PCIV.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable under the provisions of 10 CFR 50.73(a)(2)(i)(B) Any operation or condition which was prohibited by the plant's Technical Specifications. Specifically, the PCIV was inoperable and the compensatory actions were not taken within the allowed timeframe.
The function of the Primary Containment (EllS Code NH) is to isolate and contain fission products released from the reactor primary system (EllS Code AD) following a design basis accident (DBA) and to confine the postulated release of radioactive material. The Primary Containment consists of a steel vessel which surrounds the reactor primary system and provides a barrier against the uncontrolled release of radioactive material to the environment. Some leakage from the Primary Containment is assumed to occur, although the majority of the leakage is assumed to be released into the Secondary Containment (EllS Code NG). The total allowable leakage rate for the Primary Containment is designated "L sub a", and is equal to 1.2 percent by weight of the containment air volume per day. The leakage that occurs within the secondary containment is treated by the Standby Gas Treatment System (EllS Code BH) before being released at an elevated point through the Main Stack (EllS Code VL). PRINTED ON RECYCLED PAPER (9*2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 1
SEQUENTIAL IREVISION NUMBER NUMBER YEAR 3
OF 4
2010 001 o
The failed valve, the RCIC vacuum pump discharge check valve, is within the secondary containment boundary. In addition the valve discharges into the top of the Torus (EllS Code NH) and terminates below the water line. The Torus is postulated to remain water filled post accident. Therefore this valve does not communicate with the gas atmosphere within the Torus.
Primary Containment leakage criteria were established using conservative licensing basis evaluation methods in accordance with NRC Regulatory Guide 1.3. These methods conservatively assume that the postulated accident results in fuel damage with 100 percent of the core noble gas activity and 50 percent of the iodine activity released.
The Final Safety Analysis Report (FSAR) for Plant Hatch Unit 2 designates the DBA as the break of a Reactor Recirculation System (EllS Code AD) pipe which results in the rapid depressurization of the reactor vessel to the Primary Containment. However, the FSAR analysis shows that, for such an accident, resulting peak fuel cladding temperatures would be less than those required to produce damage to the fuel. The plant-specific SAFERIGESTR analysis for this accident scenario shows that no damage to the fuel cladding would occur even if additional failures are postulated, such as failures of certain power supplies and certain low pressure emergency core cooling systems. Therefore, by this analysis, the only radioactive materials present in the released coolant would be those already present due to normal operation and the small additional amount of contaminated or activated crud released from vessel internals and primary system piping during the initial stages of the transient. In addition since this valve communicates with Primary Containment through a pipe that is submerged in the Torus communication with a gaseous release is not postulated. Realistically, therefore, the 10 CFR 100 off site dose limits would likely not have been exceeded had an actual event occurred.
Based on this analysis contained in the FSAR, it is concluded that the RCIC valve failure being reported did not result in any adverse impact on nuclear safety. This analysis applies to all operating conditions.
CORRECTIVE ACTIONS
Corrective action is to incorporate this example into the Operation training program.
The failed val ve has been replaced and tested with an acceptable PCIV.
ADDITIONAL INFORMATION
Other Systems Affected: None
Failed Components Information
Master Parts List Number: 2E51-F028B Manufacturer: Rockwell International Model Number: 3674T Type: Valve, Shutoff Manufacturer Code: R344 EllS System Code: BN Reportable to EPIX: Yes Root Cause Code: X EllS Component Code: SHV PRINTED ON RECYCLED PAPER u.s. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
(9-2007)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin I. Hatch Nuclear Plant Unit 2 05000366 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 4
OF 4
2010 001 0
Commitment Information:
This report does not create any new permanent licensing commitments.
Previous Similar Events
LER 2-2009-005 is an event where a manual action could have been taken instead of entering the Technical Specification. In that event the manual action, realignment of a suction source, was not taken and the Technical Specification also was not entered. It is similar to this event in that proper entry into the Technical Specification as not made. The corrective action for this event focused on revision of a procedure to correct the specific event by requiring the manual action to be taken.
Therefore the corrective action would not have prevented the current event.
PRINTED ON RECYCLED PAPER