05000263/LER-2005-006, Re Unrecognized Plant Configuration Change
| ML053550194 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/19/2005 |
| From: | Conway J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-05-101 LER 05-006-00 | |
| Download: ML053550194 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2) 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)(viii)(A) |
| 2632005006R00 - NRC Website | |
text
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Monticello Nuclear Generating Plant Operated by Nuclear Management Company, LLC December 19, 2005 L-MT-05-101 10 CFR Part 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket No. 50-263 License No. DPR-22 LER 2005-006. 'Unrecoanized Plant Confluration Chanoe" A Licensee Event Report for this occurrence is attached.
This letter makes no new commitments or changes any existing commitments.
John T. Conway Site Vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC 2807 West County Road 75
- Monticello, Minnesota 55362-9637 Telephone: 763-295-5151
- Fax: 763-295-1454
O-E
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2007 (6-2004)
COMMISSION
, the NRC may not conduct or sponsor, and a digits/characters for each block) person Is not required to respond to, the infation collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Monticello Nuclear Generating Plant 05000263 1 of 4 TITLE (4) Unrecognized Plant Configuration Change EVENT DATE 5)
LER NUMBER (6)
REPORT DATE
- 17)
OTHER FACILITIES INVOLVED (8)
MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKETNUMBER MNUMBER NO Y05000 FACILITY NAME DOCKET NUMBER 10 18 2005 2005 006 00 12 119 2005 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR : (Checkailthatapply) (11)
MODE (9) 20.2201(b)
=
20.2203(a)(3)(1i)
=
50.73(a)(2)(ii)(B) 50.73(a)(2Xix)(A)
POWER 100 20.2201(d) 20.2203(a)(4) 50.73(aX2)(i1i) 50.73(a)(2)(x)
LEVEL (10) 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(v)(A) 73.71 (a)(4) 20.2203(aX2)(i) 50.36(c)(1)(ii)(A) 50.73(aX2Xv)(A) 73.71 (a)(5) 20.2203(aX2)(1i) 50.36(cX2) 50.73(a)(2Xv)(B)
OTHER Specify In Abstract below or 20.2203(a)(2)(iii) 50.46(a)(3)(li) 50.73(aX2)(v)(C)
In NRC Form 366A 20.2203(a)(2XIv) 50.73(a)(2)(i)(A) 50.73(a)(2)(vXD)
=
20.2203(a)(2)(v)
X 50.73(a)(2)(i)(B) 50.73(a)(2xvii) 20.2203(a)(2)(vi) 50.73(a)(2)(i)C) 50.73(a)(2)(viii)(A)
=
20.2203(a)(3)(i) 50.73(a)(2Xii)(A) 50.73(a)(2)(viil)(B)
LICENSEE CONTACT FOR THIS IER (12)
NAME TELEPHONE NUMBER (Include Area Code)
Ron Baumer [763-295-1357 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
S M
COMMANU REPORTABLE
CAUSE
SYSTEM COMPONENT FACTU RER l
£I E
FACTURER TO EPIX SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE).
lX NO DATE U
15)BMI ABSTRACT On October 17, 2005, during a review of a planned Emergency Filtration Train (EFT) maintenance work order, a shift manager recognized that the isolation used for this activity affected the Emergency Core Cooling Systems (ECCS) due to impacts on the associated Emergency Service Water (ESW) pump start logic. This maintenance work had already been approved for implementation; however, the compensatory measures planned for the EFT maintenance were determined to be inadequate and the work was put on hold prior to implementation. On October 18, 2005 during the extent of condition review for the issue, it was determined that a similar condition had previously existed. The impact on ECCS during the previous isolation was not recognized or evaluated in advance and resulted in a plant configuration control error.
The root cause evaluation for the event determined that management and supervision did not provide the necessary direction and oversight of isolation activities to ensure expectations were clear, appropriate resources were applied, and roles and responsibilities in the isolation preparation/approval, work order impact, and work order approval processes were clear.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER 6 PAGE (3)
SEQUENTIAL REVSION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 2 of 4 2005 006 00 TEXT (If more space Is required, use additional copies of NRC Fonn 366A) (17)
Description
In April of 2004, a non-compliance with 10 CFR 50 Appendix R cable [EIIS Component Code CBL]
separation requirements was identified as an operable but non-conforming condition. A subsequent July 2004 evaluation resulted in the initiation of a modification to correct the condition. Engineering performed an installation impact assessment and determined that only the Emergency Filtration Train (EFT) [EIIS System Code BH] and Control Room Ventilation (CRV) [EIIS System Code VI] system Technical Specifications (TS) would be impacted during installation of the modification. During a February 2005 management challenge board meeting, a decision was made to implement part of this modification during the 2005 Refueling Outage (RFO) and the remainder online. The decision to complete part of this work online was based on 2005 RFO resource availability and the premise that the installation only impacted the EFT and CRV systems. Isolation, installation, and preoperational testing procedures were prepared by design engineering and approved by operations/system engineering in September of 2005. The modification was installed, pre-operational testing completed, and the systems restored in October 2005.
On October 17, 2005, during a review of an unassociated planned EFT maintenance work order, a shift manager recognized that the isolation used for this activity affected the Emergency Core Cooling Systems (ECCS) [EIIS System Code BO] due to impacts on the associated Emergency Service Water (ESW) [EIIS System Code BI] pump [EIIS Component Code P] starting logic. This work had already been approved for implementation; however, the compensatory measures planned for the EFT maintenance were determined to be inadequate and the work was put on hold prior to implementation.
On October 18, 2005 during the extent of condition review for this issue, it was determined that a similar condition had previously existed. The impact on ECCS during the previous isolation to support modification work was not recognized or evaluated in advance and resulted in a plant configuration control error. The opening of the breaker [EIIS Component Code BKR] for support of the modification work resulted in a loss of the auto-start feature of the #13 ESW pump. The ESW pump was required to support the operability of the ECCS room cooler and required for the operability of the division 1 Core Spray (CS) [EIIS System Code BM] and Residual Heat Removal (RHR) [EIIS System Code BO] pumps.
Although the automatic start function was unavailable, the pump could have been manually started from the control room with the isolation in place. Having both the division 1 CS and RHR pumps inoperable at the same time placed the plant in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown action statement. A review of the event determined that the breaker was open for 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> on October 3 and for 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> on October 10, 2005, resulting in exceeding a Technical Specification action statement.
Event Analysis
Since the above event was a condition prohibited by the plant's technical specification, it is reportable under 10 CFR 50. 73(a)(2)(i)(B), "Operation or Condition Prohibited by Technical Specifications." There was no corresponding 10 CFR 50.72 notification required for this event.
The event is not classified as a safety system functional failure.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME 1)
DOCKET 2 LER NUMBER 6 PAGE (3)
FSEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 3 of 4 2005 006 00 TEXr (If more space is required, use additonal copies of NRC Form 366A) (17)
Safety Significance
The ESW system supports component cooling requirements for two basic functions. First, each train of the ESW system provides cooling to the corresponding train of the CRV system. Second, the ESW system supports various ECCS pumps by providing room cooling for the High Pressure Coolant Injection (HPCI) [EIIS System Code BJ], CS and RHR system pumps, as well as motor bearing cooling to both CS pumps and two of the four RHR pumps. The risk analysis model does not consider the ECCS room and motor cooling dependencies on ESW to be necessary for the CS, HPCI, and RHR systems to be successful in performing their safety functions. Since the ECCS pumps are assumed to be capable of performing their safety functions without the need for either motor or room cooling, the risk significance of failure to automatically start the ESW pump is negligible. Failure of the CRV cooling function has no impact on the likelihood of core damage.
The Probabilistic Risk Assessment (PRA) group performed an evaluation for significance. The risk impact incurred by defeating the automatic start circuitry for #13 ESW pump was of low significance (<
1.0 E-06/yr difference in Core Damage Frequency).
In addition to the above PRA analysis, the second division of ESW/ECCS was fully operable in accordance with technical specifications.
Cause
The root causes for the event were:
- 1. Management and supervision did not provide the necessary direction and oversight of complex activities and work management processes to ensure expectations were clear and that appropriate resources were applied.
- 2. Roles and responsibilities in the isolation preparation/approval, work order impact, and work order approval processes are unclear.
Corrective Action
The following interim actions were taken by the station after the event:
- 1. Operations issued a memorandum to communicate management expectations for development of isolations by operations personnel.
- 2. A senior experienced Senior Reactor Operator has been assigned as the Work Control Manager, specifically to enhance the isolation development and approval process.
- 3. Only Operations Department personnel are authorized to prepare isolations.
- 4. The station added a requirement in the tagging program to review the isolation during preparation for the need to conduct a technical review.
NRC FORM 36A (1-2001)
4 IU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER M6 PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 4 of 4
. 2005 006 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
The following corrective actions will be completed and tracked in the station corrective action program:
- 1. The station will revise the isolation, work impact, and work approval processes to consolidate requirements, clarify expectations, and eliminate redundancy.
- 2. The station will strengthen the training/qualification of isolation preparers and approvers.
- 3. The station will improve management oversight and tracking of isolation related issues.
- 4. The station will revise the roles and responsibilities of individuals involved with isolations, including interactions when technical reviewers are necessary.
Failed Component Identification N/A
Previous Similar Events
A review found previous events related to isolation errors:
- 1. Condition Report 02009465 - Adverse trend with respect to identifying proper technical specification limiting conditions of operation entry and exit requirements for work activities.
One of the causal factors for CR 02009465 was behavioral in nature. The causal factor in the root cause report states, uManagement and staff are demonstrating inappropriate behaviors by reviewing and approving work activities without fully understanding the technical aspects of the situation and not sufficiently challenging the information."
- 2. LER 2005 Loss of shutdown cooling. The LER identified one of the contributing causes as Operations instructions not requiring impact statements on all work packages. The implication is that reviews of certain work packages were not rigorous enough to correctly identify all of the plant impacts for a given activity. One of the corrective actions for this root cause was to strengthen procedural requirements for the use of impact statements for work orders.
- 3. LER 2005 Unexpected trip of # 16 Bus. The root cause was determined to be site directives which do not contain detailed responsibilities and actions that must be performed to accomplish the task of preparing and reviewing a Post Maintenance Test. Part of the corrective actions resulted in strengthening administrative processes to deal with these types of complex evolutions.
In the three previous Monticello events, the common outcome was strengthening requirements. For example, in the case of LER 2005-03, a corrective action strengthened the requirements for impact statements for work orders.