05000333/LER-2015-001, Regarding COLR Thermal Limits Exceeded with Elevated Fuel Support Piece
| ML15180A207 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/29/2015 |
| From: | Brian Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-15-0081 LER 15-001-00 | |
| Download: ML15180A207 (5) | |
| 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 |
| 3332015001R00 - NRC Website | |
text
- =~ Entergy JAFP-15-0081 June 29, 2015 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP P.0.Box 110 Lycoming, NY 13093 Tel 315-349-6024 Fax 315-349-6480 Brian R. Sullivan Site Vice President - JAF
Subject:
LER: 2015-001, COLR Thermal Limits Exceeded With Elevated Fuel Support Piece
Dear Sir or Madam:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 This report is submitted in accordance with 1 O CFR 50.73(a)(2)(i)(B).
There are no commitments contained in this report.
Questions concerning this report may be addressed to Mr. Chris M. Adner, Regulatory Assurance Manager, at (315) 349-6766.
~-&
Jt J3rian R. Sullivan Site Vice President BRS/CMA/mh Enclosure(s): JAF LER 2015-001, COLR Thermal Limits Exceeded With Elevated Fuel Support Piece cc:
USNRC, Region 1 USNRC, Project Directorate USNRC, Resident Inspector INPO Records Center (ICES)
NRC FORM 366 (02-2014)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (01-2014)
LICENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME James A. FitzPatrick Nuclear Power Plant
- 2. DOCKET NUMBER 05000333
- 3. PAGE 1 OF 4
- 4. TITLE COLR Thermal Limits Exceeded With Elevated Fuel Support Piece
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED
MONTH
DAY
YEAR
YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME N/A DOCKET NUMBER N/A 4
30 2015 2015 - 001 - 00 6
29 2015 FACILITY NAME N/A DOCKET NUMBER N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 100 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in
=
Background===
A fuel cell [EIIS component identifier: AC] is a fundamental unit of core design. It consists of 4 square fuel assemblies supported by a fuel support piece (FSP) [AC] with a cruciform shape control rod blade (CRB) [AA]
that moves between the fuel assemblies. The fuel assemblies in the fuel cell provide the physical channel in which the CRB travels. During refueling outages, when fuel assemblies are removed from a fuel cell, a double blade guide (DBG) [CF] is used as the physical channel. A single blade guide is a long flexible frame; essentially a shell of a fuel assembly. Two of these blade guides connected together via a bail handle forms a DBG. They are inserted into diagonal fuel bundle locations in a fuel cell. DBGs prevent the control rod blade from tipping or leaning within the cell when fuel assemblies are removed.
The James A. FitzPatrick Nuclear Power Plant (JAF) refueling outage 21 (RO21) took place from late August through October 2014. This Outage included a full core off-load in order to accomplish a variety of in-vessel work.
Event Description
- 1. On September 2, 2014, a DBG is installed at fuel cell 38-39.
In-vessel work involved needing access to the empty space not occupied by the DBG or a fuel assembly.
When the work activities transitioned from one corner to another, the DBG needed to be rotated. The process involves retracting the CRB; lifting, rotating, and reseating the DBG; then inserting the CRB.
- 2. On September 6, 2014, at 04:00, rotated the DBG at core location 38-39.
The cruciform shape of the control rod, along with the fact that often the perpendicular diagonal cell locations are normally occupied by fuel, help ensure that the DBG is properly inserted without crossing of the legs. When the control rod is not inserted into the cell, an underwater camera is utilized to ensure proper seating of the DBG within the cell. In this core alteration, the legs of the DBG became crossed (i.e. the legs become located in face adjacent vs. diagonally adjacent bundle locations within a cell) and this condition was not identified by the crew.
- 3. On September 6, 2014, at 04:12, when control rod 38-39 was being inserted the refuel floor Operator saw the DBG start to lift off the top guide and contacted the control room operator to stop. An underwater camera verified that the DBG was not seated properly and that the legs were crossed.
The DBG was wedged when control rod 38-39 was inserted and it forced the legs of the DBG to separate and press into the FSP. As the control rod lifted it raised and twisted both the DBG and FSP. The 38-39 FSP was elevated by approximately 1.5 inches and rotated counterclockwise by approximately 6 degrees. U.S. NUCLEAR REGULATORY COMMISSIONN (01-2014)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE James A. FitzPatrick Nuclear Power Plant 05000333 YEAR SEQUENTIAL NUMBER REV N0.
3 of 4 2015 -
001 00 4. On September 15, 2014, core verification was conducted.
Core verification uses underwater cameras to verify the fuel assembly location, orientation, and cleanliness and whether they were properly seated. Proper seating of the fuel assemblies is completed by lowering a camera at the edge of the shroud and rotating the camera to obtain a horizontal view across the top of the core. Core verification was completed without identifying the elevated FSP.
- 5. On October 10, 2014, during power ascension after RO21, a high temperature alarm occurred for control rod 38-39 when it was fully withdrawn to notch 48.
The site addressed the problem by evaluating the alarm setpoint and moving the control rod to notch 46 which resulted in minimizing elevated temperatures.
- 6. On February 17, 2015, Troubleshooting identified that the elevated temperatures are indicative of an elevated FSP. Reviewing the core verification video confirmed that the fuel support piece was elevated.
The coolant flow into each fuel assembly in cell 38-39 is partially obstructed by the elevated and offset FSP. A Thermal-Hydraulic evaluation by General Electric Hitach Nuclear Energy (GEH) for this condition estimated a flow reduction of 23.63%. The flow reduction changes the fuel assembly heat transfer characteristics and as an interim compensatory action, a correction factor is applied to the calculations conducted by plant operators for thermal limits (Minimum Critical Power Ratio (MCPR)) to ensure Technical Specification (TS) compliance for the four assemblies in cell 38-39.
Event Analysis
On April 30, 2015, GEH and site personnel analyzed beginning of cycle for exposure for fuel cell 38-39 and identified several instances when the thermal limit for Operating Limit MCPR (OLMCPR) was exceeded.
These instances involved times when reactor power and flow were lowered. After RO21 in October 2014, there were four instances when one of the two recirculation pumps tripped causing the plant to reduce power and flow. Since these events happened before the elevated FSP was identified, no correction factor was applied to the four fuel assemblies in cell 38-39.
TS 3.2.2 limiting condition for operation (LCO) states that all MCPRs shall be greater than or equal to the MCPR operating limits specified in the Core Operating Limits Report (COLR). If one of the MCPR values are exceeded than either the limit is to be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 25%
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The amount of time which OLMCPR was exceeded during October 2014 was greater than the maximum allowed restoration time of six hours. This deficiency is reportable per 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
Cause
The direct cause was refuel bridge operator human error when they did not identify the crossed DBG legs in fuel cell 38-39 following rotation using an underwater camera.
The apparent cause was procedure OSP-66.001, Operations Support Procedure for fuel moves and core alterations, does not provide adequate guidance when rotating DBGs. There is no guidance on what actions to take or what verification is required if the DBG is found improperly seated and the control rod is inserted.
Two contributing causes to this event are an inadequate core verification process for detecting small height anomalies prior to plant start-up and no condition report was entered into the corrective action process when the DBG at 38-39 was improperly seated so no anomaly was expected during core verification. U.S. NUCLEAR REGULATORY COMMISSIONN (01-2014)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE James A. FitzPatrick Nuclear Power Plant 05000333 YEAR SEQUENTIAL NUMBER REV N0.
4 of 4 2015 -
001
00 Similar Events
A review was performed of the past three years of Licensee Event Reports (LERs) for events JAF has not previously experienced any other elevated FSP during plant operation above 25% power.
General Electric published a service information notice, SIL 376, for Misaligned FSP in November 1982.
Several Nuclear Plants had reported instances of mis-aligned and raised FSP. The SIL recommends exercising caution when inserting DBG into empty fuel cells (no fuel assemblies or control rods) to prevent placing the legs in the wrong cells. Second, if abnormalities are detected during control rod insertion such as it stops, inserts slowly, or the DBG lifts than DBG positioning and FSP orientation must be verified.
In October 24, 2002, Pilgrim Nuclear Generating Station identified an elevated FSP of 7/8 inch during plant operation. Pilgrim experienced similar high temperature control rod readings at that location. This operating experience aided the identification of an elevated FSP at JAF.
Corrective Actions
Completed Actions Correction factor of 0.22 is applied to calculated MCPR values at fuel cell location 38-39.
Future Actions Revise OSP-66.001 to include guidance from SIL 376 Rev 1 if a double blade guide is seated with the legs crossed.
Present this event to fuel movers and refuel floor supervisors during a pre-job brief.
During RFO22 inspect the control rod, alignment pin, top fit of CRD guide tube, and fuel support piece at 38-39 for damage and replace if required.
Revise Fleet Procedure EN-RE-210 (BWR Core Verification) to add enhancements to serve as additional barrier that could detect an elevated fuel support piece or unseated fuel support piece. Add a pre-requisite that all fuel support castings have been verified seated.
Safety Significance
Operating the plant within the thermal safety limit of Minimum Critical Power Ratio (MCPR) protects the fuel cladding integrity. GEH re-analyzed the MCPR margin during the beginning of Cycle 22 with the elevated FSP modeled and factored in potential transient scenarios. GEH concluded that the initial core MCPR values remained within the Safety Limit MCPR (SLMCPR) established by TS 2.1.1.2 plus the factor contributed by potential transients. At no point did JAF ever exceed SLMCPR or had the potential to exceed SLMCPR; therefore, fuel cladding integrity was never challenged.
This Licensee Event Report (LER) pertains to the more conservative Operating Limit MCPR (OLMCPR) established by the Technical Specifications 3.2.2 and contained within the Core Operating Limits Report (COLR). When the elevated FSP on cell 38-39 was identified the station enacted compensatory measures to maintain MCPR values within OLMCPR. The OLMCPR was only exceeded in October 2014 during power changes and reduced reactor flow.
Control rod operability: The safety related function of the control rod drive system is to achieve rapid insertion of the control rods on demand to shut down the reactor. The position of the FSP and fuel assemblies cannot interfere with control rod insertion. Control rod 38-39 was scram timed tested successfully at the beginning of the cycle. Notch times of control rod drive 38-39 are normal and do not indicate that there is friction between the blade and the fuel channels. Therefore, the control rod remains Operable.
References Condition Report: CR-JAF-2015-02075, Exceeded Operating Limit MCPR Condition Report: CR-JAF-2015-00789, Apparent Cause