05000293/LER-2011-002, Technical Specification Required Shutdown - RBCCW B Declared Inoperable and LER 2011-002-00, Reactor Scram During a Planned Reactor Cool-Down with All Control Rods Fully Inserted
| ML11112A129 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/21/2011 |
| From: | Rich Smith Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.11.029 LER 2011-001-00, LER 2011-002-00 | |
| Download: ML11112A129 (15) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2932011002R00 - NRC Website | |
text
mEntergy Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Robert G. Smith, P.E.
Site Vice President April 21, 2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
SUBJECT:
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station Docket No.: 50-293 License No.: DPR-35 Licensee Event Report 2011-001-00, Technical Specification (TS) Required Shutdown - RBCCW "B" Declared Inoperable Licensee Event Report 2011-002-00, Reactor Scram During A Planned Reactor Cool-Down with All Control Rods Fully Inserted LETTER NUMBER: 2.11.029
Dear Sir or Madam:
The enclosed Licensee Event Reports (LERs) 2011-001-00, "Technical Specification (TS) Required Shutdown - RBCCW "B" Declared Inoperable" and 2011-002-00 "Reactor Scram During A Planned Reactor Cool-Down with All Control Rods Fully Inserted" are submitted in accordance with 10 CFR 50.73.
This letter contains no commitments.
Please do not hesitate to contact Mr. Joseph R. Lynch, (508) 830-8403, if there are any questions regarding this submittal.
Robert G. Smith RMB/rmb Attachments: 1. Licensee Event Report 2011-001-00, Technical Specification (TS) Required Shutdown -
RBCCW "B" Declared Inoperable (6 Pages)
- 2. Licensee Event Report 2011-002-00, Reactor Scram During A Planned Reactor Cool-Down with All Control Rods Fully Inserted (5 Pages)
PNPS Letter 2.11.029 Page 2 of 2 cc:
Mr. William M. Dean Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 INPO Records 700 Galleria Parkway Atlanta, GA 30399-5957 Mr. Richard V. Guzman, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 USNRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter Number 2.10.029 Licensee Event Report 2011-001-00, Technical Specification (TS) Required Shutdown - RBCCW "B" Declared Inoperable (6 pages)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
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 Service Branch (T-5 F53), U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to LICENSEE EVENT REPORT (LER) 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.
- 3. PAGE Pilgrim Nuclear Power Station 05000293 1 OF6
- 4. TITLE Technical Specification (TS) Required Shutdown - RBCCW 'B' Declared inoperable
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MNH DYYA YEA SEQUNIALIREV OT DYYA FACILITY NAME DOCKET NUMBER M
D NUMBER NO MONTH DAY YEAR N/A 05000 FACILITY NAME DOCKET NUMBER 02 20 2011 2011 001 00 04 20 2011 N/A 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)
Run 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) 20.2203(a)(2)(ii) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 1 R20.2203(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) 100 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
EVENT DESCRIPTION
At 0055 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> on Sunday, February 20, 2011, the Pilgrim Nuclear Power Station (PNPS) commenced a controlled shutdown of the reactor due to the 'B' train of Reactor Building Closed Cooling Water (RBCCW) being declared inoperable and expected to exceed its 72-hour Limiting Condition for Operability (LCO) as required by TS prior to return to operable status.
With the plant operating at 100% power, leakage of Salt Service Water (SSW) was detected in the RBCCW system due to high chloride levels and increased inventory in the system. An investigation into the event determined that the source of the SSW was isolated to the 'B' RBCCW heat exchanger which is designed to cool RBCCW under normal and post-accident conditions. The quantity of the leakage was determined to exceed the design limits established to ensure post-accident operation of the system and the 'B' train of RBCCW was subsequently declared inoperable.
The leak detection and repair activities identified a single tube leak resulting from an improperly modified tube sleeve (shortened and incorrect bevel). The modified sleeve was installed in a 2005 maintenance outage and over time accelerated wear on the parent tube.
BACKGROUND:
The Reactor Building Closed Cooling Water (RBCCW) System provides cooling to the Core Standby Cooling System (CSCS) components and provides a heat sink for the Residual Heat Removal (RHR) System heat exchangers. The system also provides required cooling to the equipment located in the Reactor Building during normal planned station operations, and to provide a barrier between the primary system and the Salt Service Water (SSW) System.
The RBCCW System consists of two independent closed loops for redundancy during accident conditions.
Each loop has three centrifugal pumps and takes suction from the associated RBCCW heat exchanger. A 500 gallon head tank for each loop is located at the highest point in the system and accommodates system volume changes, maintains static pressure in the loop, detects gross leaks in the system, and provides a means for adding makeup water. The two loops can be cross-tied through two 12-inch cross-tie headers using four valves.
The cross-tie valves are normally closed.
During plant operations, the RBCCW system also functions as an intermediate barrier between system equipment and the SSW system. The RBCCW loop pressure is normally higher than the salt service water system pressure preventing salt water contamination of the RBCCW system. Detectors in the RBCCW system continuously monitor radioactivity levels.
The RBCCW heat exchangers were placed in operation in approximately 1971. As a result of the station's eddy current testing program tube sleeves (also called inserts, shields or ferrules) were installed in the mid 1980's in both RBCCW and TBCCW heat exchangers. These sleeves were made of the same material as the tubes, 90-10 Copper Nickel.
EVENT DESCRIPTION
On 2/20/2011, a planned reactor shutdown/ cooldown was being performed in accordance with PNPS Procedures 2.1.5, Controlled Shutdown from Power and 2.1.7, Vessel Heatup and Cooldown to address leakage within the Reactor Building Closed Cooling Water (RBCCW) Loop 'B' heat exchanger (Reference LER 2011-001-00). During cool-down, with the startup feedwater regulating valve and reactor water cleanup (RWCU) letdown in service, Pilgrim experienced a reactor scram signal, Group II isolation, Group VI isolation and RBIS initiation on low reactor water level (reactor water level reached +10.4 inches - scram set-point is
+12 inches). All control rods were fully inserted at the time of the scram. The low reactor water level was the result of reactor water level control difficulties experienced while performing a reactor cool-down using the Mechanical Pressure Regulator (MPR). This method was selected following successful performance in the simulator during Just-In-Time (JIT) training. The use of the MPR was one of two procedurally allowed options for plant cool-down, as the second method being the use of the Bypass Valve Opening Jack (BVOJ).
BACKGROUND:
During steady state and dynamic plant conditions, reactor pressure is maintained by the Mechanical Hydraulic Control (MHC) System. During plant cooldown, one of two mechanisms can be utilized to adjust main steam line, and therefore reactor pressure limiting the rate of temperature reduction to that allowed by Technical Specifications. The Mechanical Pressure Regulator (MPR) adjusts the control bypass valve (BPV) positions to control reactor pressure at an established setpoint. The BVOJ is a motor actuated linkage that can be used to directly open the bypass valves.
EVENT ANALYSIS
During the event, the MPR was controlling reactor pressure by opening and closing the #1 BPV; the BPV movements caused reactor water level oscillations. This was due to the relatively coarse control nature of this cool-down method. Review following the event with several experienced personnel identified that the BVOJ provides a more fluid depressurization producing vessel level shrink of a much smaller magnitude than the MPR. In this case, the on-off control of the MPR resulted in the "at the controls" (ATC) operator taking action to significantly reduce the feedwater flow rate to prevent a high water level condition.
The vessel level shrink coupled with reduced feedwater flow to account for the reduced vessel inventory resulted in reactor water level lowering below the RPS actuation/RBIS and PCIS isolation setpoint of +12 inches reactor water level. The cool-down operation was stopped and operations management performed a stand-down with the operating crew.
Just In Time (JIT) training had been conducted the previous day. In attendance were the Shift Manager (SM),
Control Room Supervisor (CRS), the Administrative Control Room Supervisor, Assistant Control Room Supervisor (ACRS) and two Reactor Operators (ROs). The training included review of PNPS Procedure 2.1.7, Reactor Pressure Vessel Cooldown Rate Schedule and dynamic implementation in the simulator. The procedure directs, 'The cooldown will be accomplished using the MHC System (adjusting the MPR down to 150 psig then the Bypass Valve Opening Jack (BVOJ) for the last 150 psig or using the BVOJ from initial pressure all the way down)." No method was prescribed as preferred by the procedure, training, orU.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER NO.
3 OF 5 2011 -
002 00 collective experience of the participants, so the SM directed that both methods be evaluated by the team.
During the JIT training, incremental cooldown steps were performed using both mechanisms (MPR and BVOJ) starting at rated reactor pressure, 1030 psig. Adequate control of depressurization rate and reactor water level was experienced in both cases. A water level swell of approximately 4" occurred when the BPV was opened with a corresponding shrink of the same magnitude when the BPV closed. No adjustment to reactor water level makeup or reject was needed to maintain level within a narrow band. The decision was made by the SM to use the MPR method at the plant.
The cooldown was directed to be executed by the CRS and was initiated by the ATC operator both of who had participated in the JIT training. One of the additional ROs, who had not attended the JIT training, was assigned as the peer check. Because of the effect of normal steam loads, the cooldown was commenced from a lower reactor pressure of approximately 700 psig. The ATC operator began to lower the MPR setpoint in a continuous fashion as directed by the procedure to establish a target pressure to 585 psig. A prompt reactor water level swell of approximately 14" occurred causing the ATC to stop lowering the MPR setpoint and then take action to reduce feed water flow by closing down on the startup regulating valve. When water level began to lower, the operator recommenced lowering pressure using the MPR, and achieved the 585 psig MPR target setpoint. As observed on plant computer traces, level and BPV position cycled a total of 5 times with varying magnitude. The MPR operated by closing the BPV when pressure dropped to the established setpoint causing the cycling of BPV and indicated reactor level. When the BPV closed, vessel shrink combined with a very low feed water flow rate resulted in water level lowering to +12" in less than thirty seconds to a minimum of about 10.5".
A water level of +12" produced the expected plant response (reactor scram signal, primary containment system Group II, VI and reactor building isolations). Following the actuations and initiations, the control room crew verified that all automatic actions had appropriately occurred and took action to re-establish RWCU letdown flow path, reset the scram signal and restore normal ventilation. During the error review meeting, it was clear that the crew understood the fundamental concepts of reactor vessel level swell and shrink during depressurization and stabilization. The impact of nearly securing feedwater flow on indicated level when the MPR setpoint was reached was not fully appreciated by the RO or CRS.
An 8-hour Non-Emergency 10 CFR 50.72 notification was made to the USNRC.
CAUSE OF EVENT
The Root Cause of the event was a failed opportunity to capture and up-date the reactor cooldown procedure with relevant historical Pilgrim Operating Experience regarding previously attempted cooldown evolutions using the MPR.
Contributing Causes
The crew did not apply sufficient questioning attitude and stop when unsure, when the magnitude of the initial reactor water level swell exceeded that experienced during Just-In-Time (JIT) training.
The Simulator did n~t adequately model the plant's reactor level response during MPR'cool-down at these low flow, high temperature conditions. This condition contributed to a high sense of confidence in performing the cooldown with the MPR evolution.
EXTENT OF CONDITION:U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER NO.
4 OF 5 2011 -
002 00 A review of station errors and events for the past five years failed to identify any occurrences attributed to deficiencies in any of the following areas: failure to capture pertinent operating experience in procedure or shortfalls in simulator modeling. Accuracy of simulator modeling was additionally evaluated by review of the plant simulator index and simulator design review board meeting minutes. No significant discrepancies were found and existing deficiencies were appropriately prioritized. A review of crew and plant performance where JIT training had been utilized determined that the training was effective in supporting successful performance.
Gaps in management oversight were identified in at least two of the occurrences (CR-PNP-2009-0499 and CR-PNP-2009-4036). Management engagement and correction of at risk behaviors is a significant corporate initiative through the use of Entergy Nuclear Platform 3: Set and Continuously Enforce High Standards and Fleet Procedure EN-FAP-OM-001: Leadership Forums for Continuous Improvement. These initiatives are considered sufficient to address extent of problem / condition.
FAILED COMPONENT IDENTIFICATION:
Not applicable.
CORRECTIVE ACTIONS
Immediate corrective actions taken were to temporarily halt the cool-down operation while operations conducted a stand-down. Plant cool-down was subsequently performed successfully utilizing the BVOJ.
Corrective actions taken included the revision of the reactor heat-up / cool-down procedure to incorporate lessons learned to identify the Bypass Valve Opening Jack (BVOJ) as the preferred method for executing a reactor pressure vessel cool-down.
Corrective actions planned include the performing of an analysis of MPR/RPV and level response during plant cool-down at the plant simulator and evaluate results for disposition.
The corrective actions are being tracked in the Pilgrim Station Corrective Action Program via CR-PNP-201 1-00733.
ASSESSMENT OF SAFETY CONSEQUENCES
The event posed no threat to public health and safety.
A low reactor water level signal (+12") with all control rods fully inserted resulted in a scram signal, PCIS Group II, Group VI and Reactor Building Isolation signals. Following verification that all automatic actions had occurred as expected, the reactor scram and isolation signals were reset, restoring system configurations to their pre-initiation status. The plant remained within the established shutdown risk evaluation of the five key safety functions (Inventory Control, Decay Heat Removal, :Power Availability, Reactivity Control and Containment). At its lowest point of +10.5", reactor water level was maintained greater than 10' above top of active fuel. There was no radiological or industrial safety impact. Based on the fact that there was no challenge to nuclear, radiological or industrial safety, the impact on safety was not significant.
SIMILAR EVENTS
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER No.
5 OF 5 2011 -
002
- - 00 Pilgrim 07/10/2007 Pressure control was oscillating between the EPR, MPR and Bypass Valve Opening Jack. Pressure "Control" and "Not in Control" lights on the EPR, MPR and Bypass Valve Opening Jack were all oscillating.
During a thermal backwash on 7/10/07 while the reactor was at 50% power, the control room received a turbine trip and reactor scram on a low vacuum trip. The reactor was running at roughly 945 psig at the time of the scram. The EPIC traces show the 3 bypass valves open initially, relieve the initial post scram pressure spike and then close as the expected response. Reactor pressure dropped to roughly 840 psig. Within 10 minutes, decay heat was causing the pressure to rise. At this point, reactor pressure only increased to 928 psig when the MPR took control and the # 1 bypass valve began to oscillate. Steam supply pressure only recovered to 922 psig when the bypass valves began to oscillate. The Apparent Cause of the reactor pressure oscillation was a burr on the MPR pilot valve which most likely was caused by a piece of debris within the turbine lube oil system or age related wear to the pilot valve.
REFERENCES:
CR-PNP-2011-0773 PNPS Procedure 2.1.5 Controlled Shutdown from Power PNPS Procedure 2.1.7 Vessel Heatup and Cooldown