05000255/LER-2003-001, Re Inoperable Steam Generator Low-Level Channels
| ML030770442 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/10/2003 |
| From: | Cooper D Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 03-001-00 | |
| Download: ML030770442 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2552003001R00 - NRC Website | |
text
N oEl Committed to Nuclear Excellen 5
t' Palisades Nuclear Plant Operated by Nuclear Management Company, LLC March 10, 2003 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 PALISADES NUCLEAR PLANT DOCKET 50-255 LICENSE NO. DPR-20 LICENSEE EVENT REPORT 03-001, INOPERABLE STEAM GENERATOR LOW-LEVEL CHANNELS Licensee Event Report (LER)03-001 is attached. The LER describes the discovery that all four steam generator reactor protection system low-level trip setpoints in each steam generator were set such that the trip could occur below the allowable value specified in Technical Specification 3.3.1. This occurrence is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
SUMMARY OF COMMITMENTS This letter contains no new commitments and no revisions to existing commitments.
Douglas E. Cooper Site Vice-President, Palisades CC Regional Administrator, USNRC, Region IlIl Project Manager, USNRC, NRR NRC Resident Inspector, Palisades Attachment
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- - D
27780 Blue Star Memorial Highway
- Covert, Ml 49043 Telephone: 616.764.2000
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150.0104 EXPIRES 7 c31-2004
, the NRC may not conduct or sponsor, and a digits/characters for each block) person is not required to respond to, the information collection
- 3. PAGE PALISADES NUCLEAR PLANT 05000255 1 OF 3
- 4. TITLE INOPERABLE STEAM GENERATOR LOW-LEVEL CHANNELS 5 EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED
_I FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR NUMBER NO MO DAY YEAR FACILITY NAME DOCKET NUMBER 01 15 2003 2003
- - 001 00 03 10 2003
- 9. OPERATING
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
- (Check all that apply)
MODE 1
20 2201 (b) 20 2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50 73(a)(2)(ix)(A)
- 10. POWER 20 2201 (d) 20 2203(a)(4) 50.73(a)(2)(iii) 50 73(a)(2)(x)
LEVEL 100 20 2203(a)(1) 50 36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73 71(a)(4)
__ 20 2203(a)(2)(i)
_ 50 36(c)(1)(i)(A) 50 73(a)(2)(v)(A)
_ 73 71(a)(5)
_ 20 2203(a)(2)(ii) 50 36(c)(2)
_ 50 73(a)(2)(v)(B)
Specify in Abstract below or in 20 2203(a)(2)(iii) 50 46(a)(3)(u) 50 73(a)(2)(v)(C)
(If more space is required, use additional copies of (If more space is required, use additonal copies of NRC Form 366A)
The need for further training is being evaluated for appropriate system and design engineers in the procedures and calculation of the static pressure shift correction factor term for Rosemount differential pressure transmitters.
Rigorous application of engineering principles was reiterated.
The instrument setpoint methodology design guide is being revised to incorporate the lessons learned from this event.
The SG level transmitter calibration procedure is being revised and the transmitters will be restored to the correct values during the 2003 refueling outage.
SAFETY SIGNIFICANCE
The allowable value for the SG low-level setpoints contained in TS 3.3.1 is >25.9%. The allowable value was chosen to assure that Auxiliary Feedwater (AFW) flow would be initiated while the SG could still act as a heat sink and steam source, and to assure that a reactor trip would not occur on low level without the actuation of AFW.
NMC determined that the configuration of the low level trip setpoints resulted in a worst-case setpoint of approximately 24.71%. Although less than the TS allowable value, this setpoint is greater than the analytical value (1 8.14%) contained in the plant safety analysis, after including total loop uncertainties.
Therefore, since analytical limits were maintained, this event had no safety significance.
PREVIOUS SIMILAR EVENTS
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