05000354/LER-2004-006, Regarding High Pressure Coolant Injection Design System Requirements Not Demonstrated
| ML042660119 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/13/2004 |
| From: | Hutton J Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N04-0391 LER 04-006-00 | |
| Download: ML042660119 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 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)(i)(B), Prohibited by Technical Specifications |
| 3542004006R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 SEP 13 2004 0
LR-N04-0391 PSEG NuclearLLC U.S. Regulatory Commission Document Control Desk Washington, DC 20555 LER 354/04-006-00 HOPE CREEK GENERATING STATION - UNIT I FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 This Licensee Event Report entitled, High Pressure Coolant Injection Design System Requirements Not Demonstrated, submitted pursuant to the requirements of 1 OCFR50.73(a)(2)(v)(D).
This event was reported to the NRC (Event Report 40876). Subsequent engineering review determined that the diesel was not inoperable, therefore the notification made to report the Plant Shutdown in accordance with technical specifications 50.72(b)(2)(i) is hereby retracted.
Wincerely,-.
ames Hutton Plant Manager - Hope Creek Attachment RFY C
Distribution LER File 3.7 95-2168 REV. 7199
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/3012007 l6-2004)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person Is not required to respond to, the
- 3. PAGE Hope Creek Generating Station 05000 354 1 OF 3
- 4. TITLE High Pressure Coolant Injection Design System Requirements Not Demonstrated
- 5. EVENT DATE
- 6. LER NUMBER l
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONT SEQUENTIAL REV MONTh YEAR FACILITY NAME DOCKET NUMBER H
DAY YEAR YEAR NUMBER NO.
MOT AY YA Z
FACILITY NAME DOCKET NUMBER 7
16 2004 2004 - 006 -
00 9
13 2004
- 9. OPERATING MODE
- 1. THIS REPORT IS SUBMITTED PURSUANTTO THEREQUIREMENTS OFI 1CFR§: (Check allthatapply) 0 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 1 0 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)
F1 20.2203(a)(1) 0 20.2203(a)(4)
E 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B)
[O 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
] 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) o 20.2203(a)(2)(iii)
El 50.36(c)(2) 0 50.73(a)(2)(v)(A)
El 73.71 (a)(4) 0 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B)
El 73.71 (a)(5) 95 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER 0 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify in Abstract below nr in NRC. Fnrm nRAA
- 12. UCENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Incudre Area Code)
R. Yewdall, Licensing Engineer 856-339-2469
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANLI-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE a ~
T MFACTUJRER j
TO EPIX USjCM NE FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED 115. EXPECTED MONTH DAY YEAR SUBMISSION 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Umit to 1400 spaces, I.e., approximately 15 single-spaced typewritten lines)
On July 16, 2004 the Hope Creek High Pressure Coolant Injection System (HPCI) was declared inoperable due to the inability to demonstrate via calculation or test data that the system could perform at the level assumed in the design basis accident analyses. Upon discovery, the station entered Technical Specification (TS) 3.5.1.c, a 14 day Limiting Condition of Operation (LCO) for HPCI inoperability. At this time the D Emergency Diesel Generator (EDG) was considered inoperable due to surveillance testing being conducted. As a result, the plant entered TS 3.0.3 and initiated a plant shutdown. This event was reported to the NRC (Event Report 40876). Subsequent engineering review determined that the diesel was not inoperable, therefore the notification made to report the Plant Shutdown required by technical specifications 50.72(b)(2)(i) is hereby retracted.
An original design document that recommended that the 2 HPCI injection line restrictor orifices be rebored to a larger size had not been acted upon. The actual HPCI flow rate was determined by calculation and surveillance test results to be less than the design requirements of 5600 gpm at 1156 psia. This is a long-standing design issue.
Corrective actions include: reboring the orifices, performing testing to document adequate flow, revising design calculations, performing design document reviews of other systems and initiating an analysis to determine if HPCI was capable of meeting functional requirements in the as-found conditions.
This event is being reported in accordance with 10CFR50.73 (a) (2) (v) (D), Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.
NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YER SEQUENTIAL IREVISION R
NUMBER NUMBER Hope Creek Generating Station 05000354 l
2 OF 3 l__ __
2004 006 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
IDENTIFICATION OF OCCURRENCE Event Date: July 16, 2004 Discovery Date: July 16, 2004 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was in Operating Condition I (Power Operation), at the time of discovery. Concurrent with entering Technical Specification 3.5.1 to declare HPCI inoperable the "0D Emergency Diesel Generator (EDG) was in operation for the conduct of surveillance testing. No other required structures, systems or components were inoperable at the start of this event that contributed to the event.
DESCRIPTION OF OCCURRENCE On July 16, 2004 the High Pressure Coolant Injection System (HPCI) was declared inoperable due to the inability of the HPCI system to deliver 5600 gpm of flow to the reactor at the pressure of the lowest Safety Relief Valve (SRV) lift point as assumed in the design basis accident analyses. Upon discovery, Technical Specification (TS) action statement 3.5.1.c was entered at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, which states, With the HPCI system inoperable, restore HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to
- 200 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.! At the time of discovery, the D EDG was being tested. The diesel was considered to be inoperable because improper instrument test leads were being utilized. With the diesel considered inoperable and HPCI inoperable station Operations determined that it would be appropriate to enter TS 3.0.3 and take actions to initiate steps to shutdown the unit in accordance with the TS Action Statement. The diesel testing was secured and TS 3.0.3 was exited. This event was reported to the NRC in accordance with IOCFR50.72(b)(2)(i)/50.72(b)(3)(v)(d)
(Event Report 40876). Subsequent engineering review determined that the use of the test leads did not cause the diesel to be inoperable, therefore the notification made to report the Plant Shutdown required by technical specifications 10CFR50.72(b)(2)(i) is hereby retracted.
Hope Creek HPCI system design requirements included the ability to supply 5600 gpm of coolant to the reactor at 1135 psig. A license change to increase the setpoint tolerance for the Safety Relief Valves SRV increased the system design pressure to 1156 psig. This change was not updated in the pump in-service test. The HPCI flow rate was determined by calculation and surveillance test results to be less than 5600 gpm at 1156 psig.
The HPCI system provides coolant to the reactor through one of the core spray spargers and one of the feedwater spargers. Each supply line contains a restricting orifice designed to control the required coolant flow from each pathway.
Rebored orifices were replaced in each supply line with approved flow tests confirming adequate flow rates. The HPCI system was returned to operable at 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br /> on July 28, 2004.
This event is being reported in accordance with 1 OCFR50.73 (a) (2) (v) (D), Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accidentU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION lYEAR NUMBER lNUMBER Hope Creek Generating Station 05000354 l
3 OF 3 II 2004 006 00 CAUSE OF OCCURRENCE The cause of this occurrence was the lack of design and configuration control. An original design calculation, calculation, BJ-0018, was identified, which recommended that the existing (original) orifices be rebored to larger sizes, however, the orifices were not rebored. This caused undersized orifices to remain in place since plant startup. The system design flow requirement of the original 5600 gpm at a design pressure of 1135 psia, was subsequently changed to 1156 psia as a result of a 1996 License Change Request (LCR) increasing the tolerance of the (SRV) from 1% to 3%.
In addition, design calculation BJ-0023 used incorrect design input pump speed of 4500 rpm which was greater than the 4150 rpm speed limitation of the turbine.
The specific problems are the following:
The restricting orifices in the core spray and feedwater supply lines were undersized; An original design change recommending resizing the orifices was not acted upon; A 1996 LCR that changed SRV tolerance from 1 % to 3% did not result in appropriate changes to design basis documents; The In-Service Test for HPCI did not confirm design basis; Design calculation BJ-0023 used incorrect design input pump speed of 4500 rpm vs. 4150 rpm because speed limitations imposed by test and surveillance procedures were not used in the calculation.
PREVIOUS OCCURRENCES
A review of LERs for the two prior years at Hope Creek and Salem did not identify any similar occurrences of this type.
No similar occurrences of this type involving the HPCI system were identified.
SAFETY CONSEQUENCES AND IMPLICATIONS
There were no safety consequences associated with this event since other independent reactor protection systems were available at the time of discovery. An analysis to determine that HPCI was capable of meeting functional requirements at the as found condition is being performed.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02 has occurred because the inability to demonstrate via calculation or test data that the system could perform at the level assumed in the design basis accident analyses.
CORRECTIVE ACTION
The corrective actions to address the identified problem are as follows:
Replaced flow restriction orifices with orifices of the correct size to provide design flow at design pressure; Performed a functional test to confirm and document correct HPCI flow; Corrected HPCI surveillance and functional test procedures to test HPCI at correct design flow and pressure; Revised design calculation BJ-0023 to include design changes; Performing design document reviews of other systems to ensure that the extent of the condition is fully evaluated.
COMMITMENTS
This LER contains no Commitments.