05000354/LER-2006-004, Re Main Steam Line Radiation Monitor Set Points
| ML063040563 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/23/2006 |
| From: | Massaro M Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N06-0425 LER 06-004-00 | |
| Download: ML063040563 (6) | |
| 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 |
| 3542006004R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC OCT S 3 2006 1 OCFR50.73 LR-N06-0425 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 Facility Operating License No. NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 06-004-00 In accordance with 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC, is submitting Licensee Event Report Number 06-004-00, Docket No. 50-354.
Should you have any questions concerning this letter, please contact Mr. Frederick Berg at (856) 339-3108.
Sincerely, Michael J. Massaro Plant Manager Hope Creek Generating Station
Attachment:
Licensee Event Report
'ý- E2 95-2168 REV. 7/99
Page2 OCT S32O0 cc:
Mr. S. Collins, Administrator - Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. S. Bailey, Licensing Project Manager - Hope Creek U.S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Resident Inspector office - Hope Creek (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC may sfor each block)inot conduct or sponsor, and a person is not required to respond to, the digits/characters frecblk)information collection.
- 3. PAGE Hope Creek Generating Station 05000354 1 OF 4
- 4. TITLE Main Steam Line Radiation Monitor Set Points
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV FNTH DAY YEAR N/A MONH AY YEA YAR NUMBER NO.
M FACILITY NAME DOCKET NUMBER 08 23 2006 2006 - 004 -
00 10 23 2006 N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 0l 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C) 17 50.73(a)(2)(vii)
CE 20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1) 0l 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[I 20.2203(a)(2)(i) 0l 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[] 20.2203(a)(2)(ii)
El 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)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii) 0l 50.73(a)(2)(v)(B)
El 73,71(a)(5) 100-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)
[D 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Frederick W. Berg, Compliance Engineer 856-339-1160CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
Latent errors in the Technical Specification Daily Surveillance Log rendered all Main Steam Line Radiation Monitors (MSLRMs) inoperable-due to non-conservative monitor set points.
Subsequent crew errors in implementing thetrequired actions of the Technical Specifications led to Technical Specification (T/S) non-compliance.
On August 23, 2006, PSEG determined that the set points for the MSLRMs were above the maximum allowable T/S Table 3.3.2-2 value of 3.6 times full power background. The required T/S Action to place an inoperable channel in both trip systems in the tripped condition within one hour was not completed.
Causes include:
Guidance contained in the Technical Specification Daily Surveillance Log was incorrect.
o Operability Screening did not initially recognize that all four MSLRMs were inoperable.
Crew independent review was not effective; the reviewer did not ensure that correct actions for inoperability were initiated.
NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 6. LER NUMBER YEAR SEQUENTIAL REVISION NUMBER NUMBER 2006 004 00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Main Steam - EIIS Identifier {SB}*
- Energy Industry Identification System {EIIS} codes and component function identifier codes appear as
{SS/ccC}
IDENTIFICATION OF EVENT Event Date: August 9, 2006 Discovery Date/Time: August 23, 2006 - 1610
CONDITIONS PRIOR TO EVENT
Hope Creek was in Operational Condition 1 with reactor power at approximately 100% prior to the discovery of the incorrect set points. No structures, components, or systems were inoperable at the time of discovery that contributed to the event.
DESCRIPTION OF EVENT
PSEG procedure, HC.OP-DL.ZZ-0026, Technical Specification Daily Surveillance Log, was misleading and did not stipulate that the monitor was inoperable if the thresholds were exceeded.
The log requires Engineering to evaluate the 3 times full power background set point when the ratio of actual to baseline average full power background reaches 1.0 plus or minus 20%: Based on the calibration method, the ratio should have been 0.833 to 1.2.
On 8/15/06 a notification was generated for System Engineering to reevaluate MSLRM background set points, due to the HC.OP-DL.ZZ-0026 (Q) Main Steam Line (MSL) Average Full Power Background Check approaching the 80% limit in the procedure. The lowering background radiation levels were a result of application of noble metals during the refueling outage that ended on May 6, 2006. Action to reevaluate MSLRM background set points was delayed based on 1) the 80% value had not been exceeded and 2) the log did not stipulate that the monitor was inoperable if the thresholds were exceeded.
On 8/23/06 the MSL Average Full Power Background Check for "C" MSLRM was determined to be 0.79, below the threshold of 0.80 and a notification was created. The Control Room Supervisor (CRS) directed System Engineering to review the notification and commence evaluation of the MSLRM set points for re-adjustment using HC.SE-GP.SP-0001, Main Steam Line Rad Monitor Setpoint Determination.
While performing the operability screening on this notification, the CRS reviewed HC.SE-GP.SP-0001 and T/S 3.3.2. After evaluation, the CRS determined that using the current full power background MSLRM levels and.the existing trip set points, the set points were above the maximum allowable value of 3.6 times full power background for three of the four MSLRMs.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PA YEAR I SEQUENTIAL IREVISION Hope Creek Generating Station 05000354I NUMBER NUMBER.
3 OF 2006 004 00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
After a brief discussion and review of the data, the Shift Manager (SM) and the CRS determined that the "A","B",& "C" MSLRMs were inoperable. A discussion with the SM, CRS, and Assistant Operations Manager (AOM) led to the conclusion that a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> grace period, similar to that allowed for MSLRM setpoint adjustments following Hydrogen Water Chemistry Injection system trip or removal from service, could not be applied and the T/S Action must be entered.
At -1610 the CRS entered the T/S Action Statement (TSAS) for three inoperable monitors, "A", "B",
& "C". The T/S requirement for 3 inoperable channels is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place "A" and "C" MSLRMs in the tripped condition and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to place the "B" MSLRM in the tripped condition. While the paper work was being prepared, the AOM noted that the "D" channel was also inoperable for exceeding the 3.6 times value, and the paperwork was adjusted for all four channels. T/S 3.3.2 Action c. was entered and documented in TSAS06-309. Maintenance was contacted to place "A" and "C" channels in the trip condition.
At -1654 MSLRM channel "A" was placed in the tripped condition.
At -1657 MSLRM channel "C" was placed in the tripped condition.
At -1805 the on-coming SM noted that the T/S 3.3.2 Action c. also required tripping of the "B" & "D" channels or isolating the reactor sample valves within the one hour if there are no operable channels.
At -1814 the reactor sample valves were closed to meet the T/S required Action. This was not accomplished within the required one hour from declaration of inoperability.
A review of the Technical Specification Daily Surveillance Log determined that at least one of the MSLRMs had been below the operability threshold of.833 since 8/9/2006. Due to the error of designating the threshold for readjustment at 0.80 in the daily log, the appropriate T/S Actions were not taken at the time. This condition is reportable per 10CFR50.73(a)(2)(i)(B), "operation or condition prohibited by Technical Specifications".
CAUSE OF EVENT
Causes of the event include:
- Guidance contained in HC.OP-DL.ZZ-0026 (Q) Technical Specification Daily Surveillance Log was incorrect. This latent error was not previously detected since normal MSLRM background level response is to increase over the cycle causing the fixed setpoint to become more conservative before adjustments are necessary.,
Operability Screening did not initially recognize that all four MSLRMs were inoperable.
- Crew independent review was not effective; the reviewer did not ensure correct actions for inoperability were initiated.
EXTENT OF CONDITION
- 1.
A review of the remaining HC.OP-DL.ZZ-0026 (Q) readings found no similar circumstances in that all other required T/S readings provided correct guidance for action when exceeding threshold values. Calculations, when required, were determined to be clear and concise.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR
ý SEQUENTIAL REVISION Hope Creek Generating Station 05000354 NUMBER NUMBER 4 OF 4 2006 004 00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
PREVIOUS OCCURRENCES
A review of events for the two prior years at Hope Creek identified a delayed operability assessment on Control Room Emergency Filtration system (LER 354/05-005-00). However, corrective actions for this LER were specific to the event and would not have prevented the delayed operability assessment of the MSLRMs.
SAFETY CONSEQUENCES
The safety significance of this event is minimal.
Amendment 53 to the Technical Specification removed the automatic MSIV and Steam Line Drain isolation function of the MSLRMs in lieu of operator action. Amendment 143 added Technical Specification requirements for the trip and isolation of the Mechanical Vacuum Pumps on a MSLRM isolation signal. However, this function was not required as the vacuum pumps were not in service during the period of inoperability.
The only remaining automatic function associated with the MSLRM isolation signal is isolation of the 3/4" reactor sample line. This isolation is not credited in the safety analysis for release mitigation (Eng Calc H-1-CG-MDC-1795). Additionally, the Main Steam Line Background Radiation levels were lowering slowly due to post application response to Noble Metal Chemical Injection, and although the setpoint was not correspondingly lowered, the automatic sample line isolation function would have occurred on.rising radiation levels.
Based on a review of this event and the definition of Safety System Functional Failure (SSFF) in Nuclear Energy Institute (NEI) 99-02, no SSFF occurred.
CORRECTIVE ACTIONS
- 1.
The daily log was changed to correct the inaccurate acceptance criteria.
- 2.
Instituted interim use of a third SRO to verify correct technical specification actions are taken for emergent LCO entries.
- 3.
Issued Temporary Standing Order to ensure all licensed individuals were aware of the issue.
COMMITMENTS
This LER contains no commitments.