05000321/LER-2009-001, Regarding Pump Suction Swap for HPCI and RCIC Non-Conservative with Respect to Technical Specification Requirements
| ML091240249 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/04/2009 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-09-0689 LER 09-001-00 | |
| Download: ML091240249 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 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 |
| 3212009001R00 - NRC Website | |
text
Dennis R. Madison Southern Nuclear Vice President - Hatch Operating Company, Inc.
Plant Edwin I Hatch 11028 Hatch Parkway North Baxley. Georgia 31513 Tel 912.5375859 May 4,2009 Fax 912.3662077 SOUTHERN A COMPANY Docket No.:
50-321 NL-09-0689 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Pump Suction Swap for HPCI and RCIC Non-Conservative with Respect to Technical Specification Requirements 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 pump suction swap for HPCI and RCIC non-conservative with respect to Technical Specification requirements.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
//}~~
/ yY)aeL.'
D~~
Vice President - Hatch DRM/MJKJdaj Enclosure: LER 1-2009-001 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. J. A. Hickey, Senior Resident Inspector - Hatch
,.RC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0813112010 902007)
, the NRC may not conduct or sponsor, and a person is not required to respond 10. the information collection.
- 13. PAGE Edwin r. Hatch Nuclear Plant Unit 1 05000 321 1 OF 4
- 14. TITLE Pump Suction Swap for HPCI and RCIC Non-Conservative With Respect To Technical Specification Requirements
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 03 09 2009 2009 - 001 -
0 05 04 2009 05000
~. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) o 20.2201 (b) o 2O.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vil) 1 o 2O.2201(d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 2O,2203(a)(1) o 20.2203(a)(4) o 5O.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 5O.73(a)(2)(iii)
[] 50. 73(a)(2)(ix)(A)
- 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50,36(c)(1 )(ii)(A) o 5O.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iil) o 5O.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) 99.8 o 2O.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) o 2O.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) 1:81 50.73(a)(2)(i)(B) o 5O.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER ACILITY NAME I~ELEPHONE NUMBER (Indude Area Code)
[Edwin I. Hatch I Kathy Underwood, Performance Improvement Supervisor 912-537-5931 CAUSE SYSTEM COMPONENT MANU REPORTABLE
CAUSE
SYSTEM COMPONENT MANU*
REPORTABLE FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED SUBMISSION MONTH DAY YEAR o YES (Ifyes, complete 15. EXPECTED SUBMISSION DA TE) 1:81 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On March 9 at 1200 EDT, Unit 1 was in Mode 1 at 2798 CMWT, 99.8 percent power. Based on instrument elevation survey data taken on March 6, 2009, and subsequent evaluation on March 9. 2009, a determination was made that a level switch for the High Pressure Coolant Injection (HPCI) and a level switch for the Reactor Core Isolation Cooling (RCIC) suction swap from the Condensate Storage Tank (CST) to the suppression pool had an actual setpoint that was less conservative than that specified in the Technical Specifications. During further evaluation on March 17, 2009 it was determined that the current setpoint for the CST level suction swap to the suppression pool does not take into account the necessary reduced voltage assumption for certain DC motor operated valves. This has the effect of requiring a higher level switch setpoint to prevent vortexing at the suction from the CST. Finally, on March 30, 2009 it was determined that the level switch for the suppression pool high level suction swap for RCIC pump suction had an actual setpoint that was less conservative than that specified in the Technical Specifications. The above items are related to HPCI I RCIC pump suction swap setpoints.
This event was caused by calculation, design and configuration control weaknesses related to HPCI I RCIC CST and Suppression Pool level switches and inadequate verification of elevation during installation.
A design change has been implemented to correct the errors in the instrumentation system designs and resulting setpoint errors, and a work order has been implemented to adiust the height of the remaining instrument.
PRINTED ON RECYCLED PAPER NRC FORM 386 (ll-2007) (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 1 05000321 YEAR I
SEQUENTIAL (REVISION NUMBER NUMBER 2
OF 4
2009 001 0
NARRAllVE (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 9 at 1200 EDT. Unit 1 was in Mode 1 at 2798 CMWT, 99.8 percent power. Based on instrument elevation survey data taken on March 6, 2009, and the evaluation of that data on March 9, 2009, a determination was made that the level switch (lE41-N003) for High Pressure Coolant Injection (HPCI) (EllS Code BJ) suction swap from the Condensate Storage Tank (CST) (EllS Code KA) to the suppression pool had an actual setpoint that was less conservative than that specified in the Technical Specifications. Loss of this instrument requires HPCI to be declared inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of loss of HPCI initiation capability. Operations personnel entered the applicable Technical Specifications Required Action Statement (RAS) for an inoperable HPCI system and realigned HPCI to the suppression pool in accordance with Tech Specs. At that point, HPCI was returned to an operable status. A similar review was performed for instrumentation associated with the Reactor Core Isolation Cooling (RCIC) system (EllS Code BN). That evaluation determined that level switch (IE51-N060) for RCIC suction swap from the CST to the suppression pool also had an actual setpoint that was less conservative than that specified in the Technical Specifications. Loss of this instrument requires RCIC to be declared inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of loss of RCIC initiation capability. Operations personnel entered the applicable Technical Specifications RAS for an inoperable RCIC system and realigned RCIC to the suppression pool in accordance with Tech Specs.
At that point RCIC was returned to an operable status. The redundant instruments lE41-N002 and IE51-N061 remained operable and both HPCI and RCIC would have continued to perform their respective functions but with the automatic suction swap capability degraded due to a reduction in redundancy.
During continued review of the setpoint criteria on March 17,2009 it was determined that the setpoint for the CST level suction swap to the suppression pool was based on a switchover time of 164 seconds (82 seconds each for valves lE41-F041 and lE41-F042) to open and for the CST suction valve (lE41 FOOl) to close. Current valve stroke times are assessed at post-accident reduced battery (DC) voltages associated with post-accident assumptions that the battery chargers are not available. Taking into account the reduced voltage assumption, and other conservatisms for valve friction, effectively increases the assumed switchover time, based on the limiting valve stroke times in the current torque switch setting guide. This additional time requires the CST low level setpoint for suction swap to the torus to be higher in order to minimize vortexing at the suction from the CST. Based on HPCI I RCIC flow after throttling HPCI to 3,000 gpm. including flow instrument uncertainty, this would add about 3" to the existing setpoint. This would make the current minimum setpoint requirement non conservative based on current assumptions of instrument uncertainty. Initially the only HPCI switch setpoint that was assumed to be affected was lE41-N003. Based on the additional discovery regarding reduced battery voltages, the redundant switch i E41-N002 was also affected. However, since Unit 1 HPCI remained aligned to the suppression pool at the time of this discovery, HPCI would have continued to perform its function.
PRINTED ON RECYCLED PAPER (&-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 1 05000321 YEAR I
SEQUENTIAL IREVISION NUMBER NUMBER 3
OF 4
2009 001 0
Based on instrument elevation survey data taken as part of an "extent of condition" review from the issue identified with the RCIC suction swap from the CST to the suppression pool, a determination was made on March 3D, 2009 that the level switch for the suppression pool high level suction swap for RCIC pump suction has an actual setpoint that was less conservative than that specified in the Technical Specifications. Loss of this instrument (lE51-N062B) requires RCIC to be declared inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of loss of RCIC initiation capability. The corresponding Tech Spec Bases further defines this time to be from discovery of loss of automatic component initiation capability for the RCIC system. At the time of discovery, RCIC suction was aligned to the suppression pool in accordance with the Tech Specs as a result of the previous instrument setpoint condition. The redundant instrument, (lE51-N062A), remained operable. Additionally, RCIC would continue to perform its function but with the automatic suction swap capability degraded due to a reduction in redundancy.
CAUSE OF EVENT
This event was caused by calculation, design and configuration control weaknesses related to HPCI I RCIC CST and Suppression Pool level switches and inadequate verification of elevation during installation.
REPORTABlLITY ANALYSIS AND SAFETY ASSESSMENT The loss of the above instruments would not prevent HPCI I RCIC from performing their respective functions. However, the conditions described above do result in non-conservative setpoints, which is a condition prohibited by the technical specifications as recognized by 10CFR50.73(a)(2)(i)(B).
The HPCI system consists of a steam turbine driven pump and the necessary piping and valves to transfer water from the suppression pool or the condensate storage tank to the reactor vessel. The system is designed to inject water to the reactor vessel over a range of reactor pressures from approximately 160 psig through full-rated pressure. The HPCI system starts and injects automatically whenever low reactor water level or high drywell pressure indicates the possibility of an abnormal loss of coolant inventory. The HPCI system is designed to replace lost reactor coolant inventory in cases where a small line break occurs which does not result in full depressurization of the reactor vessel.
The Reactor Core Isolation Cooling system is a steam turbine driven system similar to the HPCI system.
RCIC is designed to provide core cooling to the reactor pressure vessel upon a loss of the feedwater/condensate supply. The RCIC system is not a credited accident mitigation system and is a much lower volume system (approximately 400 gpm vs. 4000 gpm for HPCI). However in the event of a loss of feedwater event, both the HPCI and RCIC systems are designed to automatically initiate.
The backup for the HPCI I RCIC systems is the Automatic Depressurization System (ADS, EllS Code JE) together with two low pressure injection systems: The Low Pressure Coolant Injection (LPCI, EnS Code BO) system and the Core Spray (CS, EllS Code BM) system. The CS system is composed of two independent, redundant, 100 percent capacity subsystems. Each subsystem consists of a motor-driven pump, its own dedicated spray sparger located above the core, and piping and valves to transfer water from the suppression pool to the sparger. Upon receipt of an initiation signal, the CS pumps in both subsystems start. Once ADS has reduced reactor pressure sufficiently, CS system flow begins.
PRINTED ON RECVCLED PAPER (9-2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET u.s. NUCLEAR REGULATORY COMMISSION t. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000321 YEAR I
SEOUENTIAL IREVISION NUMBER NUMBER 4
OF 4
2009 001 0
LPCI is an operating mode of the Residual Heat Removal (EllS Code BO) system. There are two independent, redundant, 100 percent capacity LPCI subsystems, each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the reactor vessel. Upon receipt of an initiation signal, all four LPCI pumps automatically start. Once ADS has reduced reactor pressure sufficiently, the LPCI flow to the reactor vessel begins.
The ADS system consists of 7 of the 11 Safety Relief Valves (SRV). It is designed to provide depressurization of the Reactor Coolant System during a small break Loss of Coolant Accident (LOCA), if HPCI I RCIC fails or is unable to maintain required water level in the Reactor Pressure Vessel (RPV). ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure Emergency Core Cooling System (ECCS) subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety. This analysis is applicable to all power levels and operating modes in which a LOCA is postulated to occur.
CORRECTIVE ACTIONS
A Design Change has been implemented to raise the elevation and change the setpoint of 1E4I-NOO2, IE4I-NOO3, and 1E51-N060.
A Work Order has been implemented to lower the installation height of IE5I-N062B to its design setpoint.
ADDITIONAL INFORMAnON Other Systems Affected: None
Failed Components Information
None Commitment Information: This report does not create any new permanent licensing commitments.
Previous Similar Events
There are no similar events within the past two years in which an instrument design and installation resulted in a less conservative setpoint than what is specified in the Technical Specifications.
PRINTED ON RECYCLED PAPER