05000321/LER-2008-003, Regarding Sensed Low EHC Pressure Causes Turbine Trip Resulting in a Reactor Scram
| ML082380876 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/25/2008 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-08-1279 LER 08-003-00 | |
| Download: ML082380876 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3212008003R00 - NRC Website | |
text
D. R**dison (Dennial Southern Nucleer Vice President Hatch Operating Campen" Inc.
Plant Edwin I. Hatch 11028 Hatch Parkway, North Baxley, Georgia 31513 Tel 912.537.5859 Fax 912.366.2077 SOUTHERN.\\
August 25, 2008 COMPANY E,,~rgy ID S~rt1~ YDur WorM""
Docket No.:
50-321 NL-08-1279 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 1 Licensee Event Report Sensed Low EHC Pressure Causes Turbine Trip Resulting in a Reactor Scram Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a)(2)(iv)(A), Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning an event that resulted in an automatic reactor scram This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
~~
Vice President - Hatch DRMlMJKldaj Enclosure: LER 1-2008-003 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatorv Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPER NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME Edwin I. Hatch Nuclear Plant Unit 1
- 2. DOCKET NUMBER 05000 321
- 3. PAGE 1 OF 3
- 4. TITLE Sensed Low EHC Pressure Causes Turbine Trip Resulting in a Reactor Scram
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 07 04 2008 2008 003 0
08 25 2008 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE 1
- 10. POWER LEVEL 99.7
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME Edwin I. Hatch / Kathy Underwood, Performance Analysis Supervisor TELEPHONE NUMBER (Include Area Code) 912-537-5931 CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
CAUSE
SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO
- 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On July 4, 2008 at 08:39 EDT, Unit 1 was in the Run mode at a power level of approximately 2797 CMWT, 99.7 percent rated thermal power. At that time testing of the 'Electro-hydraulic Control (EHC) pump auto start was in progress. During the testing, a Low EHC Pressure was sensed which resulted in a turbine trip.
The reactor tripped on turbine control valve fast closure. Following the reactor scram, reactor pressure peaked at approximately 1120 psig, resulting in four of the eleven safety relief valves (SRVs) opening as designed to reduce pressure. Water level decreased due to void collapse resulting in the closure of the Group 2 primary containment isolation valves, and a Reactor Feed Pump Turbine speed increase, both per design.
The feedwater level control system controlled reactor water level with a minimum water level of approximately 2.5 inches above instrument zero (about 160 inches above the top of active fuel). All control rods fully inserted. Pressure did not reach the nominal actuation set points for the remaining seven SRVs.
This event was caused by a sensed low pressure in the EHC system tripping the main turbine which results in an automatic reactor scram.
Performance of the weekly EHC pump auto start procedure has been suspended pending additional corrective actions. Installation of a time delay circuit, relocation of the pressure sensor tap to its previous location, along with additional corrective actions will be considered and tracked in the corrective action program.
(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 (EIIS Code XX).
DESCRIPTION OF EVENT
On July 4, 2008 at 08:39 EDT, Unit 1 was in the Run mode at a power level of approximately 2797 CMWT, 99.7 percent rated thermal power. At that time testing of the 'Electro-hydraulic Control (EHC) pump auto start (EIIS Code TG) was in progress. During the testing, a Low EHC Pressure was sensed which resulted in a turbine trip. The reactor tripped on turbine control valve fast closure. Following the reactor scram, reactor pressure peaked at approximately 1120 psig, resulting in four of the eleven safety relief valves (SRV, EIIS Code SB) opening as designed to reduce pressure. Water level decreased due to void collapse from the rapid reactor pressure increase. The decrease in water level resulted in the closure of the Group 2 primary containment isolation valves, and a Reactor Feed Pump Turbine (RFPT) speed increase (EIIS Code SJ), both per design. As reactor pressure decreased water level increased to a maximum value of approximately 33 inches due to swell after both Recirculation Pumps (EIIS Code AD) tripped automatically on an End of Cycle-Recirculation Pump Trip, as per design. The feedwater level control system controlled reactor water level with a minimum water level of approximately 2.5 inches above instrument zero (about 160 inches above the top of active fuel). All control rods (EIIS Code AA) fully inserted. Pressure did not reach the nominal actuation set points for the remaining seven SRVs.
CAUSE OF EVENT
This event was caused by a sensed low pressure in the EHC system tripping the turbine which resulted in an automatic reactor scram. The turbine trip was the result of the combination of the following modifications. A new Mark VI turbine control system was installed in Spring 2006. During that modification the point at which the pressure sensors tap off of the EHC line was changed from the manifold where accumulators are attached to the smaller tubing which also feeds the auto start solenoid valve, this resulted in a larger sensed pressure drop during testing. The digital pressure transmitters which were installed with the original Mark VI modification were changed to analog transmitters during the Spring 2008 refueling outage. The analog transmitters have a significantly faster response time of 100 ms compared to the 225 ms response time of the digital transmitters.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is required by 10 CFR 50.73 (a)(2)(iv)(A) because of the unplanned actuation of reportable system. Specifically, the reactor protection system actuated on turbine control valve fast closure when the main turbine tripped following the turbine control valve fast closure. Group 2 primary containment isolation valves closed and four of the SRVs opened on high vessel pressure. Fast closure of the turbine control valves initiates a reactor scram. The valves close as rapidly as possible to prevent over-speed of the turbine-generator rotor. Valve closing causes a sudden reduction in steam flow that, in turn, results in a reactor vessel pressure increase. If the pressure increases to the pressure relief setpoints, some or all of the SRVs will briefly discharge steam to the suppression pool (EIIS Code BL). All SRVs functioned properly (9-2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL NUMBER REVISION NUMBER Edwin I. Hatch Nuclear Plant Unit 1 05000321 2008 003 0
3 OF 3
PRINTED ON RECYCLED PAPER during the scram. Per design, SRVs A, C, G, H lifted since reactor pressure reached approximately 1120 psig. The Reactor scram initiation by turbine control valve fast closure prevents the core from exceeding thermal hydraulic safety limits following a main turbine trip. Closure of the valves results in the loss of the normal heat sink (main condenser, EIIS Code SG) thereby increasing reactor pressure, neutron flux, and heat flux transients that must be limited. A reactor scram is initiated on the valve fast closures in anticipation of these transients. The reactor trip ensures that the minimum critical power ratio safety limit is not exceeded. In this event, the main turbine tripped when low pressure in the EHC system was sensed.
The turbine trip actuated the reactor protection system and resulted in a reactor scram. All required safety systems functioned as expected given the water level and pressure transients caused by the turbine, and reactor trips. Vessel water level was maintained well above the top of the active fue1 throughout the transient.
Based upon the preceding analysis, it is concluded this event had no adverse impact on nuclear safety. The analysis is applicable to all power levels.
CORRECTIVE ACTIONS
Performance of the weekly EHC pump auto start procedure has been suspended pending additional
corrective actions
Installation of a time delay circuit, relocation of the pressure sensor tap to its previous location, along with additional corrective actions will be considered and tracked in the corrective action program.
ADDITIONAL INFORMATION
Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.
Failed Components Information
None Commitment Information: This report does not create any permanent licensing commitments.
A previous similar event in the last two years in which the reactor scrammed automatically due to a main turbine trip was reported in the following Licensee Event Report:
2-2006-002, Unit Scram On Turbine Control Valve Fast Closure. A power-load unbalance was sensed resulting in a turbine control valve fast closure. The sensed power-load unbalance was a false indication introduced by performance of a calibration procedure. This event was the result of performing a surveillance procedure on-line that should not have been performed at that time. The corrective action for this event would not have prevented the current event which resulted from a plant modification.