05000445/LER-2011-001, For Comanche Peak, Units 1 & 2, Regarding Potential for Steam Voiding Causing Residual Heat Removal System Inoperability
| ML11145A112 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/18/2011 |
| From: | Lucas M Luminant Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CP-201100678, TXX-11061 LER 11-001-00 | |
| Download: ML11145A112 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 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)(i)(B), Prohibited by Technical Specifications |
| 4452011001R00 - NRC Website | |
text
Luminant Rafael Flores Senior Vice President
& Chief Nuclear Officer rafael.flores@Luminant.com Luminant Power P 0 Box 1002 6322 North FM 56 Glen Rose, TX 76043 T 254 897 5550 C 817 559 0403 F 254 897 6652 CP-201100678 TXX -11061 Ref. #
10CFR50.73(a)(2)(v)(D) 10CFR50.73(a)(2) (vii)
May 18, 2011 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT (CPNPP)
DOCKET NO. 50-445 and 50-446 POTENTIAL FOR STEAM VOIDING CAUSING RESIDUAL HEAT REMOVAL SYSTEM INOPERABILITY LICENSEE EVENT REPORT 445/11-001-00
Dear Sir or Madam:
Enclosed is Licensee Event Report (LER) 445/11-001-00, "Potential For Steam Voiding Causing Residual Heat Removal System Inoperability," for Comanche Peak Nuclear Power Plant (CPNPP) Units 1 and 2.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this submittal, please contact Mr. Tim Hope, Manager, Nuclear Licensing, at (254) 897-6370.
Sincerely, Luminant Generation Company LLC Rafael By:
Flores 42'1 Mitch L. Lucas Site Vice President Enclosure c -
E. E. Collins, Region IV B. K. Singal, NRR Resident Inspectors, Comanche Peak
~J~6~9 A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway ' Comanche Peak
- Diablo Canyon
- Palo Verde
- San Onofre
- South Texas Project
- Wolf Creek
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 8/31/2010 (9-2007)
, the NRC digits/characters for each block) may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Comanche Peak Nuclear Power Plant Unit 1 05000 445 1 OF 4
- 4. TITLE Potential For Steam Voiding Causing Residual Heat Removal System Inoperability
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED
_NFACILITY NAME DOCUMENTMNUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR CPNPP Unit 2 05000446 I
NUMBER NO.FCINAM DOCUMENT NME 03 22 2011 2011 001 00 05 18 2011 FACILITY NAME DOCUMENTNUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
[l 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[]
50.73(a)(2)(vii)
[1l 20.2201(d)
[E 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[1 20.2203(a)(4)
H 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
H 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[:
20.2203(a)(2)(ii)
[D 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[E 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100 El 20.2203(a)(2)(iv)
[E 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
[J 50.73(a)(2)(v)(D)
VOLUNTARY LER
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Timothy A. Hope, Manager, Nuclear Licensing (254)897-6370CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX I
FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION L
YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
LJNO DATE
______________________________________________DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
In November 2009, Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 09-8,"Presence of Vapor in Emergency Core Cooling System/Residual Heat Removal System in Modes 3/4 Loss-of-Coolant Accident Conditions." This NSAL clarified the guidance provided in a previous document, and was issued to ensure consideration of the significantly reduced elevation head present when the residual heat removal (RH) system supply is transferred from the refueling water storage tank (RWST) to the emergency core cooling system (ECCS) recirculation sump.
An evaluation confirmed that the temperature limit currently applied at Comanche Peak to the RH system for alignment for shutdown cooling could result in flashing of liquids in the hot leg suction lines when the RH system suction is transferred to the RWST or the ECCS recirculation sump. The evaluation concluded that the RH temperature must be reduced to eliminate the potential for flashing of hot water within the isolated hot leg suction piping during transfer to the RWST or ECCS recirculation sump.
On March 22, 2010, a review identified three occurrences (one on Unit 1 and two on Unit 2) in the past three years where both RH trains were placed into operation prior to reaching Mode 5 (</= 200 degrees F). The cause of this event was the failure to recognize the RHR system limitations necessary to support all modes of RHR operation (shutdown cooling, ECCS injection, and ECCS recirculation) without adversely impacting each other and the consequent failure to take steps necessary to preclude voiding in the RHR system under all postulated system operating conditions. Corrective actions included revising station operating procedures to prohibit both RHR pumps from being aligned in the Shutdown Cooling Mode with RCS temperature > 200°F.
There were no actual safety consequences impacting plant or public safety as a result of the event.
All times in this report are approximate and Central Standard Time unless noted otherwise.
NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPER (If more space is required, use additional copies of SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES On March 22, 2011, as a result of a review of Information Notice (IN) 2010-11 and NSAL-09-8, it was determined that, while it appeared that controls had been established to adequately address the new concerns presented under NSAL-09-8, the basis for the 250 degree F temperature limitation previously imposed by the procedures could not be readily ascertained.
Therefore, an engineering evaluation was initiated to confirm the temperature limits previously implemented were acceptable, and to confirm the conclusion that the limits established in response to NSAL-93-004 appropriately addressed the NSAL-09-8 concerns. The engineering evaluation could not confirm that the 250 degree F limit currently applied to the RH [EIIS: (BP)]
system for alignment for ECCS injection was sufficient to prevent flashing/voiding in RH system suction piping when aligned to the RWST [EIIS: (BE)(TK)] or to the ECCS recirculation sump but could result in flashing of liquids in the hot leg suction lines. This potential exists due to the elevation differences between the RH lines at the containment penetrations and the expected containment sump level combined with the postulated containment pressure at the established 250 degree F limit. The evaluation concluded that the RH system temperature must be reduced to 210 degrees F in order to eliminate the potential for flashing of hot water within the isolated hot leg suction piping during transfer to the RWST or the ECCS recirculation sump. This information affects the manner in which the RH system will be required to be operated in Mode 3 (>/= 350 degrees F) and Mode 4 (350 degrees F > T-average > 200 degrees F).
E.
THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL PERSONNEL ERROR This condition was identified as a result of an engineering evaluation performed to confirm that temperature limits previously implemented in response to NSAL-93-004 were acceptable and appropriately addressed the concerns reported in IN-2010-11 and NSAL-09-8.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17)
The potential exists for flashing/steam voiding of an RH system hot leg suction line if the RH system is aligned for ECCS injection or recirculation, following termination of shutdown cooling with an RH temperature that exceeds 210 degrees F, and the suction of the RH system is transferred to the RWST or the ECCS recirculation sump during a LOCA. Therefore, this is considered a safety system functional failure.
In the event that the RH system became inoperable, abnormal and emergency procedures exist that provide guidance to immediately secure any RH pumps aligned for shutdown cooling to prevent pump damage, to restore core cooling through alignment of a high head safety injection pump in injection mode, and restoration of the intermediate head safety injection pumps if necessary. Existing procedures also include steps to vent and refill the RH loops if necessary. In Modes 3 or 4, at least one charging pump is available and would be aligned to the RWST. Additionally, the steam generators would be available with auxiliary feedwater providing a heat sink to aid in decay heat removal.
When this issue was first identified by Westinghouse in NSAL-93-004, Westinghouse performed an Assessment of Safety Significance. Referencing prior work documented in WCAP-1 2476, "Evaluation of LOCA during Mode 3 and Mode 4 Operation for Westinghouse NSSS," which looked at the probabilities of a LOCA in Modes 3 or 4 and available guidance for actions to cope with a shutdown LOCA, Westinghouse concluded that this issue was not risk significant in regard to large LOCAs in Mode 3 and the relative risk was not much different whether or not flashing occurs in Mode 4. In the more recently issued NSAL-09-8, Westinghouse states that the conclusion of the previous safety significance assessments for this issue was based on the low risk of this event. This conclusion remains valid since no new information changes this condition. The consequences of RH system failure due to suction flashing in Modes 3 or 4 remained bounded by the core damage consequences of the Mode 1 LOCA events. This is because of the reduced pipe break probability due to the relatively low temperature and pressure that exists for the majority of time the plant is in these modes. It is also reflective of the time the plant is in these modes, which is very short relative to the time it is in Mode 1. Therefore, the risk significance of this event is considered to be low.
IV. CAUSE OF THE EVENT
The CPNPP organization failed to recognize the RH system limitations necessary to support all modes of RH operation (shutdown cooling, ECCS injection, and ECCS recirculation) without adversely impacting each other and consequently did not take steps necessary to preclude voiding in the RH system under all postulated system operating conditions.
V.
CORRECTIVE ACTIONS
Station operating procedures were revised to prohibit both RH pumps from being aligned in the Shutdown Cooling Mode with RCS temperature > 200'F. In addition, in accordance with the CPNPP Corrective Action Program, the RH system design basis document (DBD-ME-260) will be revised to address system limitations during cooldown.
VI.
PREVIOUS SIMILAR EVENTS
There have been no previous similar reportable events at CPNPP in the last three years.PRINTED ON RECYCLED PAPER